Materials and Manufacturing, Opportunities and … · Opportunities and Constraints, in New Nuclear Build ... Intermediate upper shelf toughness (e.g. MMA, Sub.Arc) Transition toughness

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Materials and Manufacturing,

Opportunities and Constraints,

in New Nuclear Build

J.B. Borradaile, R.M. Mitchell, H.R. Dugdale (Rolls-Royce)

10 April 2014

Sustainable Nuclear Energy Conference

9-11 April 2014, Manchester

Agenda

10 April 2014

• Introduction to Rolls-Royce Capability

• Design, Structural Integrity and Reliability

• Safety Classification of Components

• HIP in the Nuclear Industry: A Case Study

• Development of HIP Nickel Based Alloys Capability

• Conclusions

Rolls-Royce Proprietary Information

Rolls-Royce has supplied nuclear

PWR plant and nuclear services for

over 50 years supporting civil and

naval applications

Trusted to Deliver Excellence

Rolls-Royce is a global business

providing integrated power systems for

use on land, at sea and in the air.

3


10 April 2014

Nuclear Sector business locations


Design Process

In-Service Modification

Configuration Management

Maintenance of Design Intent

Design Definition (creation) Design Definition (verification)

STAGE 1 Preliminary Concept Definition

STAGE 2 Full Concept definition

STAGE 3 Product Realisation

STAGE 4 Production & In-service Support

STAGE 5 Continuing In-Service Support

STAGE 6 End of life disposal

STAGE 0 Innovation & Opportunity Selection

Component Design - GQP C.4

Product Change Control - GQP C.1.4

Customer requirements and key drivers + Research and Capability requirements/ investment

Statement of Requirements + Preliminary Concept Design Scheme

Full Concept Design Scheme + Draft Design Substantiation Report + Definition of Material Requirements

Final Reference Design Scheme + Design Substantiation Report + Manufacturing Drawings + Manufacturing, installation, testing and commissioning procedures

Verification of manufacturing, Installation, testing and commissioning procedures + DSR Review

Revalidation and Inspection + Maintenance and Upkeep + DSR Review

Lay-up + Surveillance

Product Introduction and Lifecycle Management - GQP C.1.8


Design Intent

Requirements

No unacceptable

defects introduced

during welding

No environmentally

assisted cracking in-

service

No defect initiation

Stage in Life

Cycle

Design

Weld location and

geometry (ease of

welding / inspection)

Material selection Geometry and

surface finish

specification

Manufacture

Weld procedures and

welder qualification

Heat treatment

control/stress relief

Process controls

and inspection

Commission

-

Plant fill procedure

-

In-service

Control of

maintenance

requiring welding

Environmental

controls (operational

and maintenance)

Operation within

design envelope

Design Intent

Maintenance of

Design Intent

Environment

Stress Material


Nuclear Safety Principles

10 April 2014

7

•Proven Engineering Practices:

Nuclear power technology is to be based on engineering

practises which are proven by testing and experience

•Equipment Qualification:

Safety components and systems shall be chosen which

are qualified for the environmental conditions

•Continuous Improvement:

Operating organisations and designers shall seek to

improve safety standards and safety performance in

present and future plant. Techniques such as

maintaining excellent material condition and component

performance shall be employed


• ASME recognises different levels of importance of each component.

• It requires provision of a level structural reliability relative to the safety importance of the individual component (Class 1, 2 or 3).

• ASME does not provide guidance on the selection of a specific classification to assign a component.

It is the owner’s responsibility through provision of the Design Specification, to provide an appropriate classification for components.

ASME Classes


• Compliance with design codes such as ASME allows a structural reliability of ~10-5/year to be claimed, based on failure statistics from non-nuclear pressure vessels.

• For those components with intolerable consequences of failure (uncontained release of fission products to the public) it needs to be demonstrated that failure is incredible, which in the UK is defined as a failure rate <10-7/year.

• Consequently, a higher safety classification and demonstration of reliability is required for a catastrophic failure mode of the Reactor Pressure Vessel, compared with other Class 1 primary circuit components, with less severe consequences of failure.

• This introduces the concept of Incredibility of Failure or IoF.

Safety Classification


IoF Concept Multi-Layered

Major principle in Nuclear Safety is Defence in Depth, provision of multiple layers of protection

Some components it is not possible to provide this defence by physical means

For these Incredibility of Failure (IoF) needs to be demonstrated, retaining the principle of Defence in Depth, through application of appropriate experience, testing, analysis and monitoring

Conceptual Defence in Depth – based on leg element structure UK Technical Advisory Group on Structural Integrity of High Integrity Plant

TAGSI

GOOD DESIGN & MANF

TESTING FAILURE ASSESSMENT

FORWARNING of FAILURE

DEFENCE IN DEPTH

SEGREGATION DIVERSITY

REDUNDANCY


NSRP Safety Classifications

System,

Structure,

Component,

Classification

Consequences of Failure

Approximate

Failure rate /

annum

ASME III Code

Classification

Consequence of

Failure

IoF

Failure leads inevitably to

fuel failure and uncontained

fission product release <10-7 Class 1 Catastrophic

High Integrity Failure would inevitably lead

to fuel failure 10-6 Class 1 Major

Safety

Critical

Failure would lead to a

demand for a safety system

to operate to prevent fuel

failure

10-5 Class1 Serious

Safety

Related

In combination with other

failures (including operator

error ) failure would lead to

the demand for a safety

system

10-4

Class 2 Minor

Non-Safety Failure would only lead to

reduced plant availability 10-3 Class 3 Negligible

Procedure is DETERMINISTIC – probabilistic studies may be used to support the deterministic calculations

ASME III


Stages in the Procedure

Assess Damage Tolerance Step 2

Determine Risk Category Step 3

Identify Structural Integrity requirements

Step 4

Define Component Safety Classification

Step 1


1 2 3 High upper shelf toughness

High tearing resistance

(e.g. TIG)

Intermediate upper

shelf toughness

(e.g. MMA, Sub.Arc)

Transition toughness

Limited tearing resistance

Non-welded components Welds with simple

geometry and easy

access for welding

and NDE

Welds with complex geometry and

difficult access for welding and NDE

Within stress limits

Secondary stresses low

Stress relieved welds

Within stress limits

Non-stress relieved

ductile welds

Dissimilar ductile

welds

Structural

discontinuity

Gross structural discontinuity

Rapid temperature changes

Non-stress relieved non-ductile

welds

Dissimilar non-ductile welds

Known well understood

degradation mechanism

Judgements/uncertainties

resolved by

surveillance/monitoring

programmes

FUF<0.4

0.4<FUF<0.8

Moderate crack

growth

Degradation

mechanism that

results in reduction in

toughness but no

failure mode

FUF>0.8

High crack growth

Degradation mechanism that brings

about a change in failure mode

Material failure mode

Likelihood of significant defects

Loading/ Stress level

Degradation mechanism

Score

Assessment of Damage Tolerance


Ranking of Damage Tolerance

Sum the Damage Tolerance scores

Damage Tolerance Total Score

High (H) 4

Medium (M) 5 to 8

Low (L) 9 to 12

Use Damage Tolerance Ranking in Conjunction With Consequence of Failure Ranking

Matrix of Potential risk


A A B C C

A B C C C

B C C C C

IoF HI SC SR NS

Low (9-12)

Medium (5-8)

High (4)

Safety Classification (Consequences of Failure)

Damage Tolerance

• ASME Class 1

• 4 legged approach (Required)

• SMI (ASME V) + review of credible defects

• MAI to support a defect tolerance assessment

• R6 target reserve factors

• ASME Class 1

• 4 legged approach (required for cat A, recommended for cat B)

• SMI (ASME V)

• R6 sensitivity study

• ASME Class 1

• ASME Class 2 (SR), Class 3 (NS)

• 4 legged approach (useful)

• SMI (ASME V)

CATEGORY A CATEGORY B CATEGORY C

SMI (Standard Manufacturing Inspections) - Confirm quality

MAI (Manufacturing Acceptance Inspections) – Qualified inspections that target defects of structural significance and supports the fracture assessment

Matrix of Potential Risk and Structural Integrity Requirements


Examples

High Quality Butt Weld in Large Diameter Austenitic Stainless Steel Reactor Coolant Boundary Piping A postulated gross failure of this weld could result in fuel failure based on the assumption that a catastrophic failure would not be isolable or protectable. Consequence is Fuel Failure - Safety Classification is High Integrity Material failure mode = 1 Likelihood of defects = 2 Loading/stress level = 2 Degradation mechanism = 1 Total score` = 6 Damage Tolerance is Medium. The location is therefore in Risk Category B. Risk Category B welded location, has the following structural integrity requirements: Compliance with ASME III Code Class 1 Design and Fabrication requirements (including SMI) R6 defect tolerance sensitivity study TAGSI safety case structure

High Quality Butt Weld in Small Diameter Austenitic Stainless Steel Reactor Coolant Boundary Piping A postulated gross failure of this weld would not result in fuel failure as it could be protected by emergency core cooling system. Failure is protected - Safety Classification is Safety Critical Material failure mode = 1 Likelihood of defects = 2 Loading/stress level = 2 Degradation mechanism = 1 Total score` = 6 Damage Tolerance is Medium. The location is therefore in Risk Category C. Risk Category C welded location, has the following structural integrity requirements: Compliance with ASME III Code Class 1 Design and Fabrication requirements (including SMI)


Safety Classification Summary

• The traditional approach to safety classification was to designate all safety significant components as ASME III Class 1.

• IAEA and UK safety assessment principles for nuclear plants require components to be classified based on their safety functions and then designed and constructed to achieve the required reliability level.

• A safety classification process has been developed which has five levels; the highest two levels, High Integrity and IoF require additional demonstration of reliability than can be gained from strict compliance with ASME III Class 1 rules.

• A multi-legged structural integrity case is adopted for those components that require high reliability demonstration ie High Integrity/IoF using the UK Technical Advisory Group on the Assessment of High Integrity Nuclear Plant (TAGSI) format.

• The two specific areas where the ASME III Class 1 requirements may need to be exceeded to achieve the additional reliability demonstration are: • Explicit demonstration of defect tolerance

• Validation of inspection techniques to demonstrate that tolerable size of defects (plus an appropriate margin) can be reliability detected and characterized.


• Initial step is to produce powder metal of desired composition. Molten metal is poured through ring of high pressure inert gas nozzles. This breaks up molten stream into fine droplets which rapidly solidify within the atomisation tower.

• Powder is then sieved to desired size distribution, which limits segregation and inclusion size.

• A low alloy steel can is filled with the powder, degassed and sealed. The filled can is then subjected to a high temperature (>1100 C) and pressure (>100 MPa) for a number of hours until powder fully consolidated, along with a shrinkage of ~ 30%.

• Can is removed by machining or pickling

Advanced Nuclear Manufacturing Case Study : Hot Isostatic Pressing (HIP) of powder metals


Why HIP?

Attractive manufacturing route for NSRP components, able to provide protection against manufacturing route obsolescence, improved mechanical properties, better control of defects and more reproducible results.

Microstructures are isotropic, equiaxed with a small grain size, properties not normally achieved in heavy section components. This helps facilitate ultrasonic NDE examination - Additionally, inclusions are small and more benign compared to forgings.

Turnaround times and costs can be reduced when compared to large forgings.

Complex shapes can be created, which enables weld removal from the design, simplifying construction and NDT requirements


Introduction: Advantages of HIPping

• Fine, equiaxed grain size

• Improved inspectability

• Material cleanliness

• Repeatability

• Geometric complexity (near-net shapes)

• Cost

• Batch sizes

• Lead time

10 April 2014 Rolls-Royce Proprietary Information

Introduction: HIP in the Nuclear Industry

Fine, equiaxed grain size

Improved inspectability

Material cleanliness

Repeatability

Geometric complexity (near-net shapes)

Cost

Batch sizes

Lead time

•10 April 2014

21


RR Nuclear HIP Strategy Background

To satisfy the Nuclear Safety Principles a gradual

introduction strategy for HIP NSRP components

evolved

This included proving the technology for specific

applications, and development of the

specification, procurement and justification

experience.


CAT A Components - Conceptual Strength in Depth

Multi-legged structure (TAGSI 4 leg approach)

LEG 1

Interpolation of

experience

(Design and

Manufacture)

LEG 2

Functional

Testing

LEG 3

Failure Analysis

LEG 4

Forewarning of

Failure

• Multifaceted, based on experience and sound

engineering practice

• Tolerant to defects and fault conditions.

• Strength of the case is judged by the strength and

independence of each leg

• For introduction of HIP, both Leg 1 and Leg 3 needed

to be stronger

• Leg 3 is required to be strong for IoF and HI components


Gradual Introduction Strategy

Demonstrate by mechanical testing to recognised standards that HIPped material is equivalent to wrought form

Obtain manufacturing and in-service experience of the technology through HIP of non-pressure boundary components

Further develop manufacturing and in-service experience pf the technology by applying it to leak limited pressure boundaries of isolable components

Further develop manufacturing and in-service experience of the technology by applying it to isolable pressure boundary components

Apply technology to un-isolable pressure boundary components





Further develop manufacturing and in-service experience of the technology by applying it to leak limited pressure boundaries of isolable components




Gradual Introduction Strategy: Tensile properties of HIP and wrought 316L

10 April 2014

0

100

200

300

400

500

600

700

21.1 37.8 93.3 148.9 204.4 260 315.6 371.1 426.7 482.2 510

MP

a

Temperature °C

0.2% Proof HIP

0.2% Proof Wrought

UTS HIP

UTS Wrought









HIPing of hard-wearing Stellite 6 (Co-base) and Tristelle 5183 (Fe-

base) bars and HIP bonding of inserts to austenitic 304LE and

Monel 4070 small-bore globe valves since 1994

Rolls-Royce (nuclear) applications of HIP

HIPped Stellite/Tristelle

Seat machined from bar

Seat HIPped to valve body billet

Final machining

Replaced oxy-acetylene deposit of Stellite

Reduced non-conformance and removed bottleneck in route

HIPped seat provided better grain structure and in-service longevity

Rolls-Royce data – Proprietary & Confidential Information

28

Proceedings of PVP-2005, PVP2005-71711

HIPped back seat

HIPped main seat


Implementation of HIP

•10 April 2014

29

Manufacture and bonding of HIP Stellite 6 hard facings

onto valves

Oxy-acetylene Stellite 6 deposit on stainless steel

x100

HIPped Stellite 6 powder bonded onto stainless steel

x100










31

HIPped Stainless Steel 316L Omega Seals

Omega seals are welded to main

component assembly

No discernible difference between

wrought/HIPped welded material

HIP offers smaller defect sizes and an

ability to supply small quantities at

acceptable cost and lead-time

Over 500 omega seals manufactured,

with over 200 in-service. Proceedings of ICAPP 2008 Paper 8110



HIPped 316L stainless steel machined and welded omega seals.



Microstructure of Forged and HIPped 316L

Forging (ASTM No. 2) HIPped powder (ASTM No. 5)

Type 316L structures (x100)

Omega seal was first stainless steel HIP application, high level of material cleanliness required









Full size Tee piece in Type 316L ~ 2 tons

Destructively tested, isotropic mechanical properties confirmed

Production components were introduced onto nuclear plant

First use of HIP in a Primary Circuit pressure retaining application


Isolable PC boundary – Tee piece 2009


Isolable PC boundary – Tee piece 2009

Previously, forging of the Tee had

used a three ram press and closed

die process

Typical issues experienced included

large grain structures and surface-

breaking defects

HIP proved an attractive alternative

In addition to project savings, HIP

offered advantages in both

mechanical properties and improved

inspectability


• HIP pipework in 304LE austenitic

stainless steel enables elimination

of a number of large bore and

small bore stub connection welds.

• Development work produced a

stable and reproducible technique

• HIP pipework sections have also

been introduced onto plant

• Current requirement to machine

bore rules out elbows and

diameters where access cannot be

gained


Isolable PC boundary – HIP Pipework



Valve Body and Cylinder Demonstrators

Valve body and cylinder technology demonstrators manufactured from Type 304LE powder.



Pump Bowl Demonstrator

Thickest section component

produced to date

Traditionally sand cast

No inclusions above 15 μm reported

Grain size ASTM grade 5









ASME Code Case N-834

In November 2011, a code case submission for

the use of HIP Type 316L austenitic stainless

steel on nuclear plant was made to ASME boiler

and pressure vessel committee.

Code case was approved on October 22, 2013

It was the opinion of the committee that, ASTM

A988/A988M-11 UNS S31603 may be used for

Section III, division 1, subsection NB, Class 1

components in construction


Future Nuclear HIP Strategy

10 April 2014 Rolls-Royce Proprietary Information

The Development of HIPped Nickel Based Alloys

10 April 2014

43

• Interest in developing HIP Alloy 625 began in 1990

• Research programme examined both the potential of the HIP process and the opportunity to make use of new materials.

• Alloy 625 was identified as having the potential to offer benefits to a wide range of plant applications

• HIP process is especially relevant to alloys like Alloy 625, the same properties that make it appealing to a designer also made it difficult to fabricate and machine


Development of HIP NBA: Valve Production

10 April 2014

44

• Single experimental project to demonstrate the feasibility of producing an Alloy 625 valve using HIP

• Design offers a degree of complexity without being overambitious

• Opportunity to combine a number of production stages into one:

Body machining

Seat installation

Addition of stubs

Fitting of liner

• Three sections:

Optimisation of HIP process

Production of valve bodies

Rig testing of completed valve


Development of HIP Valve: Parameters

10 April 2014

45

Particle Size Distribution

0

5

10

15

20

25

30

<45 45-75 75-106 106-180 180-250 250-420 >420

Particle Size Range (μm)

% o

f P

art

icle

s

Element C Si Cr Ni Fe Mo Nb Mn P S Co Al Ta N O

Composition

% 0.02 0.37 21.3 59.3 4.36 9.3 3.53 0.33 0.15 0.007 0.04 0.003 1.05 0.108 0.021

Temperature (°C) Pressure (MPa) Time at Temperature

(°C)

1120 103 4

1160 103 3

1200 103 4


Development of HIP Valve: Optimisation of HIP

Parameters

• Effect of HIP temperature on mechanical properties


Development of HIP Valve: Valve Production

10 April 2014

47

• Three valve bodies produced

• Can design optimisation

• Rig testing considered successful

• Major departure from conventional method of producing component


Development of HIP Valve: Material

Characterisation

10 April 2014

48

• Tensile test results compared with ASME wrought data

0

100

200

300

400

500

600

700

800

900

1000

0 100 200 300 400 500 600

Temperature °C

Str

ess M

Pa Wrought 0.2% proof

Wrought UTS

HIP 0.2% proof

HIP UTS


Development of HIP Valve: Material Characterisation

10 April 2014

49

• Fracture toughness values were lower than expected


Development of HIP Valve: Material Characterisation

10 April 2014

50

• Results from rotating beam specimens are comparable to wrought material, whilst those from cantilever bend tests are superior


Development of HIP Valve: Project Conclusions

10 April 2014

51

• Project was determined to have successfully demonstrated the feasibility of producing a valve using the HIP process

• Significant reduction in the number of production stages

• Mechanical properties of Alloy 625 require more work

• Extensive development work required to transition valve from prototype to production part


Further Development of NBA

10 April 2014

52

• Interest in the HIPping of NBA has recently been ignited

• Development of Alloys 690 and 625

• Basic test programme aiming to characterise materials and optimise HIP parameters

• Material properties are compared to their wrought equivalents: regulatory requirements

• Aiming to manufacture HIP NBA components for plant applications


Further Development of NBA: Alloy 625

10 April 2014

53

• University of Birmingham programme

• Production of 15 kg HIP bars

• Small powder size: 45 ± 15 μm

• HIP Conditions: 1160°C, 103 MPa for 240 mins

• Heat Treatments are being studied


The Development of HIPped Alloy 625: Mechanical Properties

10 April 2014

54

0

100

200

300

400

500

600

700

800

900

1000

0 100 200 300 400 500 600

Temperature (°C)

UT

S (

MP

a) HIP UTS

Wrought ASME UTS

HIP 0.2% proof

Wrought ASME 0.2% proof

• Tensile tests showed favourable mechanical properties when compared with wrought ASME data


The Development of HIPped Alloy 625: Mechanical Properties

Sample Vickers

(H v0.3)

Brinell Rockwell

1 260 Equivalent

Brinell

hardness

10 mm C

ball 3000

kgf (HB)

Equivalent

Rockwell

hardness

150 kgf

(HRC)

2 275

3 281

4 265

5 266

6 260

7 281

8 267

9 276

10 271

Average 270 H v0.3 257 HB 25 HRC

Temperature L T

RT 84 106

RT 96 108

RT 92 102

RT 103 103

RT 105 105

RT 96 119

• Favourable hardness and Charpy impact test results


HIP Development of NBA: Alloy 625 MA

10 April 2014

56


HIP Case Study Summary

10 April 2014

57

HIP powder processing has become a valuable manufacturing

technique

The use of stainless steel HIPped components has been

extensively validated through both laboratory and prototype

component testing

Work is beginning to optimise the HIPping of Inconel alloys,

including Alloy 625

The methodology established for taking HIP stainless steel

components from design phase up to component safety case will

be key for the introduction of other HIPped materials onto plant


Sizewell B (Gen III size) Naval Propulsion Equivalent

RPV Steam

Generator

Pressuriser Reactor

Coolant

Pump

RPV

Steam

Generator Pressuriser RCP

Size Matters: Propulsion components are significantly smaller than land based plant


• HIP has been demonstrated for applications on Naval Propulsion

Plant .

• An ASME Code Case has been achieved for the use of HIP Type

316L austenitic stainless steel on Class 1 components.

• HIP advantages include inspectability of the component, control of

defects, batch sizes, lead times and are not only metallurgical.

• There is no reason that HIP should not be used on Civil Plant and

Small Modular Reactors with their requirement for increased

production rates may provide the impetus

Conclusions


59


Documents

Materials and Manufacturing, Opportunities and … · Opportunities and Constraints, in New Nuclear Build ... Intermediate upper shelf toughness (e.g. MMA, Sub.Arc) Transition toughness