NUREG/CR-6726 

SAND2001-1147P

Aging Management and 

Performance of Stainless 

Steel Bellows in Nuclear 

Power Plants 

Sandia National Laboratories

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001

0%9*R 0 

D)2(E U 

*
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NUREG/CR-6726 

SAND2001-1147P 

Aging Management and 

Performance of Stainless 

Steel Bellows in Nuclear 

Power Plants 

Manuscript Completed: April 2001 

Date Published: May 2001 

Prepared by 

J.L. Cherry, J.M. Clauss, M. Pilch, J.J. Gregory, J.A. Smith 

Sandia National Laboratories 

Albuquerque, NM 87185-0744 

H.L. Graves III, NRC Project Manager 

Division of Engineering Technology 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

NRC Job Code W6690
Abstract

In commercial nuclear power plants, many containments use stainless steel bellows at piping penetrations, in dry

well-to-wetwell vent lines, or as a part of the fuel transfer tube. Bellows are an integral part of the primary contain

ment pressure boundary in nuclear power plants. These bellows allow for thermal expansion and contraction of the 

pipe, vent line, or fuel transfer tube, and accommodate relative motion between the containment and other structures, 

such as a shield wall. 

Ensuring that bellows in operational plants remain leak tight is an important safety issue. A number of bellows have 

been replaced when transgranular stress corrosion cracking caused them to leak. Determining how to evaluate bel

lows for age-related damage is a critical issue that must be addressed for plants operating under their original license, 

as well as for plants applying for license extension. 

The types of bellows, locations in a containment, and required bellows leak tests are described for major categories 

of boiling-water reactor (BWR) and pressurized-water reactor (PWR) power plants. Compilation of instances of 

bellows cracking, excessive leakage, or structural integrity questions, as well as testing issues and other degradation 

problems, are reviewed and summarized. The aging mechanisms that caused the degradation are discussed. Inspec

tion methods currently in use, and a typical industry approach to managing bellows aging are described. Degradation 

that resulted from corrosion, misalignment, and other damage was considered. The risk and regulatory significance 

of containment bellows when subjected to Current License Basis (CLB), degraded, and severe accident (SA) condi

tions are also addressed.

iii
Contents 

Abstract ........................................................................................................................................................................ iii 

Acknowledgm ents ........................................................................................................................................................ vii 

Acronym s and Initialisms .............................................................................................................................................. ix 

I INTRODUCTION ............................................................................................................................................ 1-1 

1.1 Purpose and Objective ............................................................................................................................. 1-1 

1.2 Background .............................................................................................................................................. 1-1 

1.3 Scope ........................................................................................................................................................ 1-2 

2. BELLOW S DESCRIPTION ............................................................................................................................. 2-1 

2.1 Containment Structure ............................................................................................................................. 2-1 

2.2 Containment Bellows ............................................................................................................................... 2-1 

3. OPERATIONAL EXPERIENCE ..................................................................................................................... 3-1 

3.1 Operating and Service History ................................................................................................................. 3-1 

3.1.1 Operating Experience from Nuclear Regulatory Commission Generic Communications ........... 3-1 

3.1.2 Operating Experience from Industry Publications ...................................................................... 3-1 

3.1.3 Operating Experience from Nuclear Plant Reliability Data System Data .................................... 3-2 

3.1.4 Operating Experience from License Event Report Data ............................................................. 3-2 

3.1.5 Nuclear Plant Reliability Data System and License Event Report Comparison .......................... 3-4 

3.2 Operational Experience Summ ary ............................................................................................................ 3-4 

4. AGING M ECHANISM S AN D EFFECTS ON BELLOW S ......................................................................... 4-1 

4.1 Electrochem ical Aging M echanism s and Effects ................................................................................ 4-1 

4.2 M icrobiologically Influenced Corrosion ............................................................................................. 4-1 

4.3 Stresses in Bellows ................................................................................................................................... 4-2 

4.3.1 Static Loads ................................................................................................................................. 4-2 

4.3.2 Cyclic Loads ................................................................................................................................ 4-2 

4.3.3 Differential M ovement ................................................................................................................ 4-3 

4.4 Degradation Related to Installation, M aintenance, or Operation ............................................................. 4-4 

4.5 Radiation Effects .................................................................................................................................... 4-4 

5. IN SPECTION AN D EVALUATION OF BELLOW S ................................................................................ 5-I 

5.1 Requirem ents for Continued Operation during Current License Period .................................................. 5-1 

5.1.1 Surveillance Program s ................................................................................................................. 5-1 

5.1.2 Plant W alkdowns ......................................................................................................................... 5-2 

5.1.3 Leakage Testing, 10 CFR 50, Appendix J .................................................................................. 5-2 

5.1.4 Settlement Trending Program ...................................................................................................... 5-3 

5.1.5 Detailed Fatigue Analysis Using Current Condition .................................................................... 5-3 

5.1.6 Cleaning ....................................................................................................................................... 5-3 

5.1.7 Repair .......................................................................................................................................... 5-4 

5.2 License Renewal Requirements ............................................................................................................... 5-4 

5.2.1 Aging M anagement Review ........................................................................................................ 5-4 

5.2.2 Special Considerations for Stainless Steel Bellows ..................................................................... 5-4

v
6. REGULATORY AND RISK SIGNIFICANCE OF DEGRADED BELLOWS .......................... 6-I 

6.1 Representative Industry Logic for Bellows Replacement ........................................................................ 6-1 

6.2 Sandia National Laboratories Testing of Like-New and Degraded Bellows ....................................... 6-1 

6.3 Severe Accident Risk Perspective Regarding Bellows Leakage .............................................................. 6-7 

6.4 Future Research Recommendations .................................................................................................. 6-9 

7. SUMMARY AND CONCLUSIONS ................................................................................................................ 7-1 

7 .1 S u m m ary .................................................................................................................................................. 7 -1 

7 .2 C o nc lu sio n s .............................................................................................................................................. 7 -1 

7.3 A dditional R esearch N eeded .................................................................................................................... 7-2 

8 . R E F E RE N C E S ................................................................................................................................................. 8 -1 

Figures 

Figure 2-1. Typical Piping Penetration with Bellows ............................................................................................... 2-4 

Figure 2-2. Typical Piping Penetration with Bellows - Steel Containment ............................................................... 2-4 

Figure 2-3. Typical Piping Penetration with Bellows - Concrete Containment ........................................................ 2-5 

Figure 2-4. Typical Fuel Transfer Tube - PW R ....................................................................................................... 2-5 

Figure 6-1. LLRT Flow Rate vs. Time (from Brown and Tice, 1993) .................................... 6-2 

Figure 6-2. Representative of Industry Logic for Replacement of Bellows .............................................................. 6-3 

Tables 

Table 2-1. BW R Primary Containment List ............................................................................................................ 2-2 

Table 2-2. PWR Primary Containment List ............................................................................................................. 2-2 

Table 2-3. Typical BW R Mark I Bellows (Monticello) .......................................................................................... 2-6 

Table 2-4. Typical BW R Mark II (Nine Mile Point Unit 2) .................................................................................... 2-6 

Table 2-5. Typical BW R Mark III Bellows (Grand Gulf) .................................................................................... 2-6 

Table 2-6. Typical PWR Ice Condenser Bellows (Watts Bar) ................................................................................ 2-7 

Table 2-7. Typical PWR Reinforced Concrete Containment Bellows (Palisades) .................................................. 2-7 

Table 2-8. PW R Steel Containment Bellows (Prairie Island) .................................................................................. 2-7 

Table 2-9. Typical PWR Sub-Atmospheric Bellows (Surry) ................................................................................... 2-7 

Table 5-1. Aging Management Programs for Bellows ............................................................................................ 5-2 

Table 6-1. Experiment Results from Like-New Containment Piping Bellows Subjected to Severe Accident 

C o n d itio n s .............................................................................................................................................. 6 -4 

Table 6-2. Experiment Results from Degraded Bellows Subjected to Severe Accident Conditions ....................... 6-5 

Table 6-3. Insight from Sandia National Laboratories' Testing of Like-New and Degraded Containment 

B e llo w s ................................................................................................................................................... 6 -6 

Table 6-4. Consequences for Selected Source Terms from NUREG/CR-l 150 (NRC, 1990) ................................ 6-9

vi
Acknowledgments

The US Nuclear Regulatory Commission (NRC) sponsored this research program. Herman L. Graves is the NRC 

program manager. 

Information about penetration failures was gathered by Sandia National Laboratories (SNL) for the "Aging Man

agement Guideline for Commercial Nuclear Power Plants - Containment Penetrations," which is an unpublished 

work prepared for the US Department of Energy's (DOE's) Commercial Light-Water Reactor program. The DOE 

program, with Dennis Harrison as the DOE Nuclear Energy program manager, will publish their report through the 

Electric Power Research Institute. 

This report uses reference and database information gathered for the DOE program. The DOE report is broad and 

considers all penetrations, while this report specifically addresses bellows issues. This project's additional research 

led to valuable insights that apply specifically to the bellows in nuclear power plants.

vii
Acronyms and Initialisms

AMR aging management review 

ASME American Society of Mechanical Engineers 

B&PV Boiler and Pressure Vessel 

BWR boiling-water reactor 

CFR Code of Federal Regulations 

CLB Current License Basis 

CLWR commercial light-water reactor 

DBA design basis accident 

DOE Department of Energy 

ECL early containment leak 

EJMA Expansion Joint Manufacturers Association 

EPRI Electric Power Research Institute 

FR Federal Register 

FSAR Final Safety Analysis Report 

ILRT integrated leak rate test 

INPO Institute for Nuclear Power Operations 

IPA Integrated Plant Assessment 

ISI in-service inspection 

LCL late containment leak 

LER License Event Report 

LLRT local leak rate test 

LOCA loss of coolant accident 

MELCOR system level code 

MIC Microbiologically Influenced Corrosion 

MPFF maintenance preventable function failure 

NCF no containment failure 

NEI Nuclear Energy Institute 

NPRDS Nuclear Plant Reliability Data System 

NRC Nuclear Regulatory Commission 

ORNL Oak Ridge National Laboratory 

PB Peach Bottom 

PRA probabilistic risk assessment 

PT Penetrant Test 

PWR pressurized-water reactor 

RC reactor coolant 

RCIC reactor core isolation cooling 

RHR residual heat removal 

RPV reactor pressure vessel 

RWCU reactor water cleanup

ix
SA severe accident 

SCC stress corrosion cracking 

SCFH standard cubic feet per hour 

SEQ Sequoyah 

SG steam generator 

SNL Sandia National Laboratories 

SSCs systems, structures, and components 

SSE safe shutdown earthquake 

TF trichlorotrifluoroethane 

TGSCC transgranular stress corrosion cracking 

TIP traversing in-core probe 

TLAA time-limited aging analyses 

TR B test return, train B

x
1. INTRODUCTION

1.1 Purpose and Objective 

Most steel and a few concrete containments are con

structed with stainless steel bellows at piping penetra

tions that are subject to large movements. Bellows al

low relative movement between the steel containment 

shell and the concrete shield wall, while minimizing the 

load imposed on piping and the wall. Bellows are also 

used in other locations, such as boiling-water reactor 

(BWR) Mark I containment drywell-to-wetwell vent 

lines. Pressurized-water reactor (PWR) and BWR 

Mark III containments may include fuel transfer tube 

bellows, which allow relative movement between the 

containment and the transfer tube to the spent fuel pool. 

Bellows are an integral part of the primary containment 

pressure boundary in nuclear power plants. If an acci

dent occurs, bellows may be subjected to pressure and 

temperature conditions that cause considerable axial 

and lateral deflections. 

To ensure public safety, it is necessary to understand 

bellows aging issues in operating nuclear power plants. 

Age-related degradation of bellows observed in operat

ing nuclear power plants affects the overall performance 

of the containment system. Preventive maintenance and 

bellows replacement may be necessary during the cur

rent license period to assure that there is no loss of in

tended function(s) and no unacceptable reduction in 

safety margins. In addition, continued operation of 

commercial plants for periods that extend beyond the 

original 40-year license period may be a desirable op

tion for many US utilities. 

US utilities must meet certain Nuclear Regulatory 

Commission (NRC) requirements before the NRC will 

extend a facility's operating license. One requirement 

is to show that the aging of systems, structures, and 

components (SSCs) within the scope of license renewal 

is managed so that these items will not degrade to the 

point where they no longer perform their intended func

tions. The License Renewal Rule (10 Code of Federal 

Regulations [CFR] 54, 1996) specifically requires an 

aging management review (AMR) of containment 

penetrations. 

This research assesses the performance of containment 

bellows when subjected to Current License Basis 

(CLB), degraded, and severe accident (SA) conditions. 

Degradation that results from corrosion, misalignment, 

and other damage has been considered. The objective 

of the work is to provide recommendations that can be 

used for plant in-service inspection, license renewal,

and for determining the impact degraded bellows may 

have on containment capacity. 

1.2 Background 

Bellows are used primarily in steel containments and 

are of two main types: (1) used in both BWR and PWR 

containments, process piping bellows vary in size from 

15 to 150 cm in diameter. Similar in design, fuel trans

fer tube bellows typically are about 105 cm in diameter. 

(2) Vent line bellows, which are used in BWR Mark I 

containments, range in size from 165 to 315 cm in di

ameter. Process piping and fuel transfer tube bellows 

are normally constructed of two plies of SA240, type 

304, stainless steel which are separated by a thin wire 

mesh (-0.25-mm wire diameter). The redundant outer 

ply was intended to provide a way to check for leakage 

of the bellows by pressurizing the space between the 

plies and noting any drop in pressure. In contrast with 

process piping bellows, the majority of vent line bel

lows are single ply; approximately 10% are two ply. 

Single-ply vent line bellows are tested during the inte

grated leak rate test (ILRT) (10 CFR 50, Appendix J, 

1995), because they cannot be tested locally. 

The two types of leak rate tests discussed in this report 

are defined in 10 CFR 50, Appendix J. The ILRT de

termines the overall leakage rate for the containment 

though all possible leakage paths, including leakage 

though containment welds, valves, fittings, and all 

penetrations of the containment. The ILRT is con

ducted after the containment has been completed and is 

ready for operation and periodically thereafter. The 

other is the local leak rate test (LLRT), which tests for 

local leak rates across containment penetrations, in

cluding containment bellows. 

Bellows are designed to accommodate thermal expan

sion of the pipes and differential movements that occur 

"* between the containment shell and the concrete 

shield building in most steel containments, 

"* between the drywell and wetwell in Mark I BWRs, 

and 

"* between the containment building and the transfer 

tube in Mark Ills and PWRs. 

The design bases for these movements are obtained by 

summing appropriate combinations of the maximum

1-1
deformations associated with normal operation, safe 

shutdown earthquake (SSE), loss of coolant accident 

(LOCA), settlement, and other loading conditions that 

could cause differential movement. 

The design standards are provided by the Expansion Joint Manufacturers Association (EJMA, 1998). Nor

mally, bellows are conservatively designed to withstand 

more than 150 cycles of design-basis loadings, with 

some estimates that 5000 cycles of design-basis loading 

can safely be withstood. Bellows typically experience 

only a few cycles from thermal loads at startup and 

shutdown of the reactor. 

In the unlikely event of a SA, the bellows could be 

subjected to pressure, temperature, and deflections well 

beyond the design basis. The axial displacement of the 

bellows would be relatively small until the pressure in 

the containment vessel caused plastic yielding in the 

containment wall; plastic yielding will not occur until 

the containment pressure is well above the design pres

sure. After the containment begins gross plastic defor

mation, small increases in pressure would result in large displacements that the bellows must accommodate. 

Typically, the bellows will not become fully com

pressed or elongated until the pressure is several times 

larger than the design pressure of the containment. In 

most cases, the bellows would be compressed, but in a 

few cases, the bellows would be elongated due to the 

bellows being installed on the inside of the containment. 

In addition to the bellows being elongated or com

pressed, they are also deflected transversely due to up ward movement of the containment vessel walls caused 

by the temperature and pressure loadings. 

1.3 Scope 

In performing the tasks described in this report, the 

following references were reviewed: 

"* Electric Power Research Institute (EPRI) and other 

industry-generated reports; 

"* NRC bulletins, information notices, circulars, ge

neric letters, and reports; 

* CFR and Federal Register (FR); 

"* vendor manuals;

"* industry codes and standards (e.g., American Soci

ety of Mechanical Engineers [ASME], and EJMA); 

"* miscellaneous references and technical papers; 

" Institute for Nuclear Power Operations (INPO) 

Nuclear Plant Reliability Data System (NPRDS) 

and NRC/Oak Ridge National Laboratory (ORNL) 

License Event Report (LER) databases; and 

"* input from host utilities for the US Department of 

Energy (DOE) Containment Penetrations Aging 

Management Guideline. 

Stainless steel expansion joint bellows, used in both 

BWR and PWR containments, have been evaluated in 

this study. Applicable codes and standards for the de

sign, testing, and repair of bellows were considered 

during this review. Locations where bellows typically 

are used, bellows type, and material are described in 

this report. This information provides a baseline for the 

Chapter 6 assessment of risk significance of bellows 

leakage. 

Compilation of instances of bellows cracking, excessive 

leaking, or structural integrity considerations as well as testing issues and other degradation problems are re

viewed and summarized. Data on failure of bellows 

from INPO's NPRDS, and the NRC/ORNL LER data

bases and available literature are described. A review 

of NRC information notices, bulletins, and circulars was 

also conducted to identify age-related failures. Detailed 

evaluations of the reduction in pressure capacity caused 

by the damage were not made at this time; however, 

conditions observed have been described. 

An assessment is made about the procedures that have 

been used to qualify bellows for continued service after 

aging degradation has occurred, such as nonlinear finite 

element analysis, elastic-plastic fracture mechanics, and 

leak rate projection. Chapter 7 provides recommenda

tions for additional criteria that may need to be consid

ered to ensure the bellows meet their functional and 

structural design requirements for the remainder of their 

design life, as well as for additional periods of time if 

the operating license is extended. 

Finally, an assessment has been made about the 

strengths and weaknesses of the in-service inspection 

programs used.

1-2
2. BELLOWS DESCRIPTION

2.1 Containment Structure 

Containment designs for the 110 nuclear power plants 

operating in late 1996 are identified in Table 2-1 

(BWRs) and Table 2-2 (PWRs). Typically, concrete 

containments have an interior steel liner plate. Simi

larly, steel containments are surrounded by a concrete 

shield wall. Therefore, independent of the containment 

design, containment penetrations pass through a rela

tively thin steel wall and a relatively thick concrete wall. 

In a concrete containments, the concrete wall, liner, and 

penetration nozzle move together as primary contain

ment temperature and pressure vary. This is not the 

case for a steel containment where the containment ves

sel and penetration nozzle move independently of the 

fixed concrete shield wall. 

2.2 Containment Bellows 

Some piping penetrations for steel containments and 

some fuel transfer tube penetrations contain metallic 

expansion joints that are pressure boundary components 

for the primary containment. In addition, vent line bel

lows in BWR Mark I containments are also pressure 

boundary components for the primary containment. 

These stainless steel expansion joints contain single or 

double bellows that provide for differential movement 

at the penetrations. This is necessary to accommodate 

construction misalignment, potential settlement differ

entials, and movement caused by temperature, pressure, 

or other operational or accidental loadings. Most vent 

line bellows in BWR Mark I containments are single 

ply, and most process piping and fuel transfer tube bel

lows in BWR and PWR containments are double ply 

with four to eight convolutions. The thickness of a sin

gle ply is 0.76 to 1.52 mm. The double-ply bellows 

generally have a tight stainless mesh to separate the 

plies. Two current issues associated with these stainless 

steel bellows are 

" potential for degradation due to fatigue and trans

granular stress corrosion cracking (TGSCC) 

(Brown and Tice, 1993; Berg and Brown, 1995); 

and 

"* inability to leak test the double-ply bellows with air 

pressure between the plies (Brown and Tice, 1993; 

NRC, 1992).

Numerous types of bellows are installed in US commer

cial nuclear power plants. The common bellows manu

facturers identified during the preparation of this report 

are Chicago Bridge and Iron, Pathway, Parker Hanne

fin, and Tube Turns. 

Piping systems that penetrate the primary containment 

can be designed with the containment penetration as a 

pipe anchor location (Figure 2-1). The expansion joint 

shown in the pipe sleeve outside the containment 

(Figure 2-1) relieves thermal stresses that would de

velop from differential thermal movement of the pipe 

and sleeve. In addition, the expansion joint can ac

commodate some misalignment of the penetration cen

terline and reduce transverse loadings generated by 

piping outside containment. 

Typical piping penetrations with bellows for steel con

tainments are shown in Figure 2-1 (annulus between 

steel containment and shield wall) and Figure 2-2 (small 

gap or compressible material between steel containment 

and shield wall). In these situations, a significant 

amount of pipe movement must be accommodated at 

the penetration to avoid high loads at the connection to 

the containment. The sleeved opening in the concrete 

shield wall is relatively unimportant as long as it is large 

enough not to interfere with the penetration design. The 

double bellows stainless steel expansion joints in the 

penetration sleeve can absorb both axial and transverse 

movement. The guard pipe protects the bellows from 

damage in the case of a pipe break. The primary con

tainment atmosphere resides between the process pipe 

and the guard pipe and between the guard pipe and the 

sleeve. The multiple flued head (part of the piping), the 

sleeve, and the expansion joint all represent primary 

containment pressure boundary barriers. Figure 2-3 

shows typical piping penetrations with bellows in con

crete containments. 

A typical fuel transfer tube, as used in a PWR contain

ment, is shown in Figure 2-4. A fuel transfer tube links 

the primary containment refueling canal with the fuel 

transfer canal and spent fuel pit in the fuel handling 

building. The fuel transfer tube is a shielded underwa

ter pathway for moving spent fuel assemblies out of 

primary containment during refueling operations. Ex

pansion joints are required to accommodate both con

struction misalignment and differential movement be

tween the refueling canal, containment structure, and 

spent fuel pit.

2-1
Table 2-1. BWR Primary Containment List

Primary Containment Designation* 

Unit Mark 0 Mark I Mark II Mark III 

a b a b c a b 

Big Rock Point 1 

Browns Ferry 1, 2, and 3 3 

Brunswick I and 2 2 

Clinton I 

Cooper 1 

Dresden 2 and 3 2 

Duane Arnold I 

Edwin I. Hatch 1 and 2 2 

Fermi 2 1 

Grand Gulf 1 

Hope Creek 1 

James A. FitzPatrick 1 

LaSalle I and 2 2 

Limerick 1 and 2 2 

Millstone 1 1 

Monticello 1 

Nine Mile Point 1 1 

Nine Mile Point 2 1 

Oyster Creek I 

Peach Bottom 2 and 3 2 

Perry I 

Pilgrim I 

Quad Cities I and 2 2 

River Bend 1 

Susquehanna 1 and 2 2 

Vermont Yankee I 

WNP-2 I 

Total Units (37) 1 22 2 1 5 2 2 2 

*Notes: 

Mark 0: Large, dry, steel sphere. 

Mark I a: Steel drywell and wetwell. 

Mark I b: Reinforced concrete dry-,ell and steel wetwell. 

Mark II a: Steel drywell and wetwell. 

Mark II b: Reinforced concrete drywell and wetwell. 

Mark II c: Reinforced concrete drywell and post-tensioned concrete wetwell. 

Mark Ili a: Reinforced concrete drywell and steel wetwell. 

Mark IH b: Reinforced concrete drywell and wetwellTable 2-2. PWR Primary Containment List

Primary Containment Designation* 

Unit Ice Post-tensioned Reinforced Steel Sub

Concrete Concrete Atmosphere 

a b a b c d a b a b 

Arkansas Nuclear One I and 2 2 

Beaver Valley I and 2 2 

Braidwood I and 2 2 

Byron I and 2 2 

Callaway I I 

Calvert Cliffs I and 2 2 1 1
Table 2-2. PWR Primary Containment List (continued)

Primary Containment Designation* 

Post-tensioned Reinforced Sub

Concrete Concrete Atmosphere 

a b a b c d a b a b 

Catawba I and 2 2 

Comanche Peak I and 2 2 

Crystal River 3 1 

D.C. Cook I and 2 2 

Davis-Besse 1 1 

Diablo Canyon 1 and 2 2 

Fort Calhoun I 

Ginna I 

H.B. Robinson 2 1 

Haddam Neck 1 

Indian Point 2 and 3 2 

Joseph M. Farley 1 and 2 2 

Kewaunee 1 

Maine Yankee I 

McGuire 1 and 2 2 

Millstone 2 1 

Millstone 3 1 

North Anna l and 2 1 1 2 

Oconee 1,2, and 3 3 

Palisades I 

Palo Verde 1, 2, and 3 3 

Point Beach I and 2 2 

Prairie Island 1 and 2 2 

Salem l and 2 2 

San Onofre 2 and 3 2 

Seabrook 1 I 

Sequoyah 1 and 2 2 

Shearon Harris I 

South Texas Project 1 and 2 2 

St. Lucie l and 2 2 

Summer I 

Surry l and 2 2 

Three Mile Island 1 I 

Turkey Point 3 and 4 2 

Vogtle l and 2 2 

Waterford 3 1 

Watts Bar 1 I 

Wolf Creek 1 I 

Zion l and 2 2 

Total Units (73) 7 2 4 1 32 1 11 1 7 6 1 

*Notes: 

1. Ice condenser a: Steel cylinder with concrete shield building. 

Ice condenser b: Reinforced concrete cylinder with steel lining. 

2. Post-tensioned concrete a: Concrete cylinder with steel liner, I -D vertical tendons. 

Post-tensioned concrete b: Concrete cylinder with steel liner, diagonal tendons. 

Post-tensioned concrete c: Concrete cylinder with steel liner, 3-D tendons. 

Post-tensioned concrete d: Concrete cylinder with steel liner, 3-D tendons plus secondary containment. 

3. Reinforced concrete a: Concrete cylinder with steel liner. 

Reinforced concrete b: Concrete cylinder with steel liner plus secondary containment. 

4. Steel: A steel cylinder containment vessel with a concrete enclosure building. 

5. Sub-atmospheric a: Reinforced concrete cylinder with steel liner. 

Sub-atmospheric b: Reinforced concrete cylinder with steel liner plus secondary containment.

2-3
-Impingement Ring

kw sUldr Guard Pipe Insul atlon,, / 

Outside Shield Buildng WaI •Inside • Vessel 

Containment Shell 

Figure 2-1. Typical Piping Penetration with Bellows. 

Bellows 

Expansion Joint (Steel) 

Containment Bellows Cover Plate Shell -- / 

Centering Spacers q ,Multiple Flued Head Ftth 

In sulation -4-4 

Process •• 

Pipe 

Guard Pipe 

5 

Sleeve 

Concrete Sleeve 

Insidol 

Drywal I 

Figure 2-2. Typical Piping Penetration with Bellows - Steel Containment.

ngs

2-4
OUTS IDE 

GT4INMt='JT

Coofing 

air outlet

Figure 2-3. Typical Piping Penetration with Bellows - Concrete Containment.

Transfer Can 

Uner

Containment Unernab Air Test Conn.----

/ Slee•

.TOW Channel j "containment Wal 7 C Spent Fuel ""-..fAir Test •' Notson Stud •" • Exa - --- Conn . / ...Anchors (-Ste;el:'•' 

Gasket -mn 

Blind- Flange •! Fuel Transfer TUbe 

Fasteners L.-. . 7' . -. In, • i *,. ..,, WlSev 

'\ / . : ,• • ;;t Expansion ' 1 /...;'._ .. • :"~j'. . Joint ".

"- h (Eiastomr.o 

-Test Channel

Figure 2-4. Typical Fuel Transfer Tube - PWR.

Pit WY I"

sion Joint,.: 

. Weld Neck 

Flange

'I..C 

> ul ..

I K.' . , 

nt Fuel Pit Liner

2-5

ION Hc'O

ve-Is: :.."':" .. ' . .- - -•

1, I• I = .=
New, unaged stainless steel bellows are tested to char

acterize and predict their response under design basis 

accident (DBA) conditions. These studies have shown 

that the level of extension or compression necessary to 

cause bellows failure, and resulting loss of containment, 

is several times greater than that required to satisfy the 

primary containment design specifications. Analytical 

studies accompanying these tests, which also provided 

information regarding the fatigue resistance of these 

stainless steel bellows, showed that they could with

stand many more extension and compression cycles 

than expected during the operating life of the plant.

Bellows are typically made of austenitic stainless steel 

SA240. Type 304. However, other designations have 

been used in a few cases: A-167-63, Type 304 and 

SB443-alloy 625 Gr 1. 

The types of bellows and the numbers of bellows used 

as part of the primary containment boundary are shown 

for typical BWR Mark I (Table 2-3), Mark II (Table 

2-4), and Mark III (Table 2-5) plants. Similarly, this 

information is given for a typical PWR Ice Condenser 

(Table 2-6), Reinforced Concrete (Table 2-7), Steel 

Containment (Table 2-8), and Sub-Atmospheric plant 

(Table 2-9).

Table 2-3. Typical BWR Mark I Bellows (Monticello)

Table 2-4. Typical BWR Mark II (Nine Mile Point Unit 2)

Table 2-5. Typical BWR Mark III Bellows (Grand Gulf)

2-6

Description Test Type* Number of Bellows 

Drywell-to-Torus Vent Pipes A 8 

Main Steam A, B 4 

Main Steam Drain Line A, B I 

Feedwater A, B 2 

RCIC Steam Supply A, B I 

HPCI Steam Supply A, B I 

RHR Supply (from Reactor Recirc) A, B 1 

RHR Return and LPCI A, B 2 

RWCU Supply A, B 2 

Core Spray A, B 2 

RHR Head Spray (to RPV Head) A, B I 

Total 24

*Notes: 

1. This list was developed from the Monticello Final Safety Analysis Report (FSAR), Table 5.2-3a. Revision 10 

(NRC, 2000b). 

2. Type A: 10 CFR 50, Appendix J, ILRT 

3. Type B: 10 CFR 50, Appendix J, LLRT
Table 2-6. Typical PWR Ice Condenser Bellows (Watts Bar)

Description Test Type* Number of Bellows Fuel Transfer Tube A, B I 

Feedwater Bypass A, B 4 

Feedwater A, B 4 

Main Steam A, B 4 Steam Generator (SG) Blowdown A, B 4 

Chemical and Volume Control System A, B I 

Residual Heat Removal Hot Leg Injection A, B 1 

RHR Pump Discharge TR B A, B 2 

Safet Inection System A, B 2 

Pressure Relief Valve Discharge A, B I 

Safety Injection A, B 3 RC Drain Tank A, B 3 Glycol Supply and Return A, B 2 

RHR Supply to Tanks A, B I 

Maintenance Port A, B 2 Total 

35 

*Notes: 

I. This list was developed from the Watts Bar FSAR, Table 6.2.4-1, Amendment 85 (NRC, 20000. 2. Type A: 10 CFR 50, Appendix J, ILRT 

3. Type B: t0 CFR 50, Appendix J, LLRT

Table 2-7. Typical PWR Reinforced Concrete Containment Bellows (Palisades) 

Description Test Type Number of Bellows No bellows none 0 Total 0 

Table 2-8. PWR Steel Containment Bellows (Prairie Island)

Main Steam Header Num er of Belle A, B 2 Main Feedwater Headers A, B 2 SG Blowdown A, B 2 RHR Loop Out and In A, B 2 CVCS Letdown Line A, B I 

Total 9 

*Notes: 

I. This list was developed from the Prairie Island FSAR, Table 5.2- 1, [(Part A), Unit 1], Revision 13 (NRC, 2000d). 

2. Type A: 10 CFR 50, Appendix J, ILRT 

3. Type B: 10 CFR 50, Appendix J, LLRT 

Table 2-9. Typical PWR Sub-Atmospheric Bellows (Surry)

Test Type* Number of Bellows 

A,B 

___I

2-7

i
3. OPERATIONAL EXPERIENCE

All bellows used in penetrations are safety related be

cause they are required to prevent or mitigate the con

sequences of accidents that might result in offsite expo

sures covered by regulatory guidelines (10CFR 100, 

1962). As noted in Section III.f (page 22477) in the 

supplementary information portion of the amended li

cense renewal rule (10 CFR 54, 1996), all containment 

penetrations (i.e., bellows) must be included in the 

scope of license renewal as "passive" components. 

PWR and BWR Mark IIIs include access penetrations 

termed fuel transfer tubes, which utilize a bellows to 

allow relative movement between the containment and 

the transfer tube. Most steel containments, and a few 

concrete containments, provide metallic bellows to ac

commodate piping thermal expansion and/or the differ

ential movement between the steel containment shell 

and concrete shield wall. The BWR Mark I contain

ments have vent line bellows between the drywell and 

the wetwell. 

3.1 Operating and Service History 

Four information sources were reviewed to gain an un

derstanding of operating and service history. Only 

documents that pertain to aging of bellows were consid

ered in this report: 

(1) A review of NRC generic communications (bulle

tins, circulars, generic letters, and information notices) 

was conducted to assess the regulatory perspective of 

US operating experience. (2) Industry studies and lit

erature were searched for documents relating to pene

tration operating history and failures. Containment 

penetration data derived from (3) INPO's NPRDS and 

(4) NRC's LER databases were also reviewed. Compo

nent failures described in these databases were analyzed 

to identify significant failure mechanisms and their 

relative likelihood of occurrence. 

3.1.1 Operating Experience from Nuclear 

Regulatory Commission Generic 

Communications 

NRC Information Notice 92-20 (1992) documented the 

inadequacy of performing local leak rate testing for 

piping penetrations by pressurizing between the plies in 

two-ply expansion joint bellows. Tests being performed 

in this manner were Type B LLRTs in accordance with 

10 CFR 50, Appendix J (1995). Tests at Quad Cities, 

Dresden, Perry, and Clinton indicated that these between-the-plies tests could underestimate the leakage 

across the bellows by a factor of more than 20. The 

conclusion was that these two-ply bellows were manu

factured using a method that inhibited the free flow of 

air between the bellows plies. The NRC provided ex

emptions to plants with this type of two-ply bellows. 

The exemptions specify that an alternative means for 

testing these bellows is required and test frequencies 

must be increased until the existing bellows are replaced 

with testable bellows. Alternative testing includes test

ing bellows where any level of leakage has been de

tected using helium leak detection equipment. 

3.1.2 Operating Experience from 

Industry Publications 

Sams and Trojovsky (1980) analyzed LERs of various 

containment penetration events submitted to the NRC 

between January 1, 1972, and December 31, 1978. 

Penetration components were defined as the penetration 

assembly and all hardware associated with the normal 

operation of the penetration. Manufacturer, failure 

mode, failure mechanism, type, and date of event were 

recorded. 

Sams and Trojovsky (1980) identified 431 potentially 

applicable LERs, for all penetrations. Of these 431 

reports, three contained instances of bellows failures. 

On December 13, 1974, Dresden Unit 2 reported that a 

penetration bellows showed excess leakage. The cause 

was identified as a cracked bellows with numerous pin

holes. On June 16, 1975, Zion Unit 2 reported that a 

penetration bellows was leaking in excess of allowable 

limits. On August 1, 1977, Dresden Unit I failed its 

ILRT: two bellows had cracks; five bellows had pin

hole leaks. Nine failed bellows were reported by Sams 

and Trojovsky (1980). 

Shackelford et al. (1985) presented design data and 

detailed drawings of piping and access penetrations for 

26 nuclear units. The primary reason for gathering the 

data was to further develop methodologies for analyzing 

and testing penetration responses under SA conditions. 

Relative stiffness ratios for large access penetrations 

were developed. These ratios were used to select those 

penetration models for which detailed analysis and 

testing should be performed. 

Kulak et al. (1985) discussed analyses of large hot pip

ing penetrations tested to pressures and temperatures 

beyond their design specification. The analysis also 

evaluated the fatigue resistance of piping penetration

3-1
expansion joints. Based on conservative elastic calcu

lations, the stainless steel bellows in these hot piping 

penetrations should be able to withstand more than 150 

major compression cycles. 

Greimann et al. (1991) discuss the nondestructive ex

amination of full-penetration welds in the guard pipe 

and sleeve of hot piping penetrations. Volumetric ex

amination of these welds, using ultrasonic techniques 

previously approved by the NRC and ASME, to identify 

flaws found the techniques to be inadequate for flaw 

detection. Subsequent radiographic examination of 

these welds identified flaws that exceeded allowable 

levels. These welds were made from one side with 

backing rings. It is apparent that this weld configura

tion did not allow appropriate interpretation of ultra

sonic examination results. Licensees were requested to 

ensure that similar welds in hot piping penetrations are 

nondestructively examined using radiographic tech

niques. 

Lambert and Parks (1994) performed a series of 13 tests 

on piping and vent line expansion bellows in a "like

new" condition to various levels and combinations of 

internal pressure, temperature, axial compression or 

elongation and lateral deformation. Some of the tests 

were conducted at elevated temperature (218'C). The 

test objective was to determine the behavior of the bel

lows in a SA, with conditions that were well beyond 

design conditions. Test results showed that bellows in 

the "like-new" condition are capable of remaining leak 

tight under extreme combinations of displacement, 

pressure, and elevated temperatures in both compres

sion and extension. 

Lambert and Parks (1995) also tested six piping and 

vent line bellows that had been artificially aged to pro

duce TGSCC in the bellows. The test objective was to 

determine the behavior of the corroded bellows in a SA. 

The bellows were subjected to elevated pressure, tem

perature, and lateral and axial deformation conditions 

similar to the "like-new" bellows (Lambert and Parks, 

1994). The goal of testing the corroded bellows was to 

produce corroded conditions evident via visual inspec

tion, but not severe enough to cause the bellows to leak. 

Lambert and Parks (1994) concluded, "Because of the 

limited number of corroded specimens and the problems 

experienced in producing and quantifying the precise 

amount of desired corrosion, it is impossible to draw 

any comprehensive conclusions regarding the leak-tight 

capabilities of bellows with varying degrees of corro

sion." However, as one might expect, tests showed that 

corroded bellows did not perform as well as "like-new" 

bellows.

3.1.3 Operating Experience from Nuclear 

Plant Reliability Data System Data 

Bellows failure data were reviewed. One source of fail

ure data is the INPO's NPRDS database, which can be 

queried to obtain information such as the type of com

ponent, date of discovery, cause category, and a brief 

description of the event. NPRDS data are not focused 

directly on component aging; NPRDS does not neces

sarily address root cause or mechanism of component 

degradation. Additionally, not all degradations ob

served during maintenance activities are identified in 

the database. Not all plants participate in the NPRDS 

database; those nuclear power plant personnel who have 

reported may not have provided data for their plant's 

entire period of operation. As a result of these limita

tions, the database cannot be used to provide probabil

istic information about the reliability of a bellows with 

respect to age-related degradation. However, the data 

can be used to corroborate bellows degradation infor

mation identified by other sources. 

The NPRDS database was searched using component 

category (i.e., containment penetrations) and subcate

gory (process piping). Very little degradation and fail

ure data were available in NPRDS for piping penetra

tions. Thirty-five piping penetration incidents were 

identified; however, only 12 of the incident reports at

tributed incidents to leaking stainless steel expansion 

joint bellows. 

In most cases, degradation was identified through leak

age tests (10 CFR 50, 1995). Maintenance action, spe

cifically bellows replacement, was required to resolve 

these degradation incidents. 

There were three incident reports that involved fuel 

transfer tube aging, where the principal degraded sub

component was a gasket. The root cause was not identi

fied, but the failures were not the metallic bellows. 

3.1.4 Operating Experience from 

License Event Report Data 

NRC LERs are another source of penetration failure and 

degradation data. LERs are issued by nuclear plant 

operators when component failures and plant operating 

events meet the reporting requirements specified in 10 

CFR 50.73 (1995). As with NPRDS data, LERs do not 

directly record data related to component aging. In 

addition, the criteria for issuance of an LER do not en

compass all component failures. Hence, LER data pro

vide only a partial picture of failure information and the 

data may or may not be representative of general com3-2
ponent failure behavior. LER data can be used to sup

port findings derived from other data sources (e.g., 

NPRDS, industry studies) and to corroborate postulated 

aging mechanisms. 

Sams and Trojovsky (1980) describe a search of LERs 

for penetration failures from 1972 through 1978 (see 

Section 3.1.2). A search of the LER database main

tained by ORNL identified 447 abstracts for the period 

1980 through 1998 that might potentially apply to pri

mary containment penetrations. Each report was re

viewed; in cases where applicability of a given report 

could not be reliably determined, the report was dis

carded. Five LERs were applicable to bellows failures 

in primary containment penetrations. The failures were 

identified during leak rate testing and surveillance. The 

causes of the failures were identified as TGSCC. A 

brief summary follows of the bellows failure informa

tion in the LERs for the 1980 through 1998 period. 

(Section 3.1.2 contains a brief summary of bellows fail

ures in the LERs for the 1972 to 1978 time period.) 

Dresden Nuclear Power Station Unit 2 

On April 2, 1991, while Unit 2 was shutdown for a 

short maintenance outage, six drywell bellows penetra

tions were inspected for cracks. (During the previous 

Unit 2 refuel outage, eight of the 24 bellows had a posi

tive leakage rate. The inspection lot of six was chosen 

from these eight based on accessibility and dose.) 

Through-wall cracks were found on the exterior bellows 

ply at penetration X-144. 

The initial indications on bellows X-144 were identified 

through a penetrant test (PT). Following the PT exami

nation, the bellows were pressurized and coated with a 

soap bubble solution to check for through leakage. This 

inspection revealed additional leak indications unde

tectable by PT. The bellows were also inspected with 

helium leak detection equipment. The results of this 

inspection revealed leakage past both the inboard and 

outboard bellows plies. The root cause of the indica

tions found on bellows X-144 was attributed to TGSCC. 

The largest crack was 7/16 inch long. 

On January 21, 1993 with Unit 2 in a refuel outage, 

bellows penetrations X- 113, X-125, X-149A, and X

149B were replaced with a new design which provided 

increased space between the plies. No reason was given 

for replacing the bellows. This change allowed the total 

surface of the bellows to be challenged during Type B 

Local Leak Rate Testing. During the 1993 refuel out

age, bellows penetration X-144 was sealed from within 

the drywell. This planned modification eliminated bellows X-144 as an Appendix J Type B testable volume 

(10 CFR 50, 1995). 

Dresden Nuclear Power Station Unit 3 

During the September 1991 refueling outage, 91 pene

trations were tested, including valves, electrical pene

trations, bellows, and other primary containment pene

trations. Two bellows penetrations, X-105A and X

107B, required repair and were replaced with a new 

design which provided increased space between the 

plies. These replacements allow the total surface of the 

bellows to be challenged during Type B Local Leak 

Rate Testing. 

Ouad Cities Unit I 

In the past, double-ply bellows have been Type B 

LLRT tested by pressurizing between the inner and 

outer plies of the bellows using a local test line. The 

X-25 bellows were tested using this method on Novem

ber 14, 1990, and the leakage rate was 4.3 SCFH. An

other test was conducted on February 26, 1991, by pres

surizing between the plies; the leakage rate was 6.0 

SCFH. The bellows were visually inspected for leak

age, and several tiny pinhole cracks were discovered in 

the outer ply by using soap bubble solution. 

During the pressurization phase of the Type A test 

(ILRT) on February 28, 1991, larger visible cracks were 

discovered in the outer layer of the X-25 bellows. Fol

lowing the ILRT, a LLRT was performed, and the leak

age rate of the bellows was 6.0 SCFH. Because the 

leakage of the bellows during the ILRT was believed to 

be much higher than that measured during the Type B 

LLRT test, a plate was welded across the purge exhaust 

pipe inside the drywell to allow testing the bellows un

der the same conditions as the ILRT. The bellows were 

tested from inside the containment on March 12, 1991, 

by pressurizing through the purge exhaust pipe to the 

inner ply of the bellows; measured leakage was 137.4 

SCFH. The X-25 bellows was then cut out and replaced 

with a new bellows. 

H.B. Robinson Steam Electric Plant Unit 2 

During the 1984 steam generator (SG) replacement out

age, the three SG blowdown lines (penetration numbers 

S-24, S-26, and S-30) were replaced with larger diame

ter blowdown lines to accommodate a higher blowdown 

rate for the new SGs, and the insulation was replaced 

with insulation sized to fit the larger diameter pipe. In 

1986, a penetration bellows failure occurred in SG 

blowdown line penetration S-24. The bellows was sub

sequently replaced during 1987; following an analysis 

of the defective bellows, the failure mechanism was

3-3
determined to be TGSCC. At that time, recommenda

tions were made to consider future replacement with 

materials resistant to TGSCC. During 1989, leaks were 

again detected in the penetrations bellows for all three 

penetrations, and repairs were scheduled for 1990. 

During these repairs, an undetermined quantity of water 

was discovered inside the S-30 penetration, and a chlo

ride-free insulation was installed on the piping in pene

tration S-30. However, TGSCC-resistant materials 

were not considered for the other two penetrations 

(S-24 and S-26). 

On July 21, 1995, inspection of the penetration S-24 

bellows revealed a crack approximately an inch or two 

in length; further analysis revealed the presence of chlo

rides inside the penetration. It was concluded that con

densation of water from the PPS-supplied air inside the 

penetration wetted the pipe insulation, and transported 

the chlorides contained in the insulation material to the 

penetration bellows. The presence of these chlorides on 

the stainless steel material of the penetration bellows 

caused the bellows to fail due to TGSCC. Penetration 

S-24 bellows assembly was replaced by July 31, 1995. 

Containment penetrations S-24, S-26, and S-30 were re evaluated and re-classified from 10 CFR 50, Appendix 

J, "Type B" penetrations (i.e., tested to detect and 

measure local leakage) to 10 CFR 50, Appendix J, 

"Type A" penetrations (i.e., tested to measure the over

all integrated leakage of the primary containment). 

3.1.5 Nuclear Plant Reliability Data 

System and License Event Report 

Comparison 

NPRDS and LER data compare well for the bellows. 

Both databases show a small number of failures (12 

from NPRDS and 16 from LERs), with similar aging 

mechanisms (e.g., TGSCC). Twelve bellows failures 

were reported in the NPRDS during the 1974-1992 time 

period, and sixteen bellows failures were reported in the 

LERs (9 bellows failures were reported by Sams and 

Trojovsky (1980) for the 1972-1978 period, and 7 bel

lows failures were found during a search for the 1980

1998 period). From reading the LER reports, it is obvi

ous that bellows have been replaced previously that are 

not included in the total count of 16 bellows failures 

discussed here.

3.2 Operational Experience 

Summary 

Four primary sources of information (Section 3.1) were 

used in this report to characterize penetration historical 

performance. Each provides a consistent perspective on 

bellows aging. No statistical inferences can readily be 

drawn concerning bellows failure probability or rate, 

because not all cases of bellows degradation have been 

identified. Despite this limitation, several general ob

servations can be stated: 

" A number of bellow failures have occurred 

throughout the industry, with these failures primar

ily caused by TGSCC. 

" Degradation of bellows usually has been identified 

through leak rate tests. 

" After leak tests have identified penetrations with 

excessive leak rates, visual methods have identified 

pin-holes and cracks (with the aid of PT in some 

bellows). Visual methods did not find TGSCC 

damage to some bellows in penetrations with leak

age that was within allowable limits, but larger than 

expected. (It was not clear from the literature 

whether the leakage was through the bellows, or 

another part of the penetration component, or 

whether the cause of leakage was even determined.) 

" After leak tests have identified penetrations with 

excessive leak rates, helium leak detection equip

ment has been utilized successfully to locate leak 

locations. 

" Given an aggressive environment (i.e., a humid 

environment and nearby insulation with chlorides, 

so that chlorides can leach out of the insulation and 

onto the bellows), bellows failure can occur over a 

period of a few years. 

" Bellows failures have occurred in nuclear power 

plants that are fairly new, as well as in older plants. 

Additionally, new bellows used to replace older 

bellows have also developed TGSCC and have had 

to be replaced themselves. However, many older 

bellows have not experienced TGSCC. Therefore, 

it appears that the environment (i.e., presence of 

chlorides or other chemical agents) is a major fac

tor, and bellows age, by itself, is not a good indi

cator of the likelihood of TGSCC damage.

3-4
4. AGING MECHANISMS AND EFFECTS ON BELLOWS

Bellows aging mechanisms have been identified, with 

consideration given to the design and operating his

tory of bellows, along with relevant industry research, 

information, and reports. For purposes of this docu

ment, the significance of an aging mechanism centers 

around whether the aging mechanism results in loss of 

bellows functionality during the current or license 

renewal period. 

4.1 Electrochemical Aging 

Mechanisms and Effects 

This report focuses specifically on bellows and the 

stainless steel used in their construction. Low-carbon 

steels are often used in the penetration component, 

and stainless steel sometimes is welded to the low

carbon steel. In carbon steel/stainless steel combina

tions that have surfaces inadvertently wetted or wet

ted under normal operation (i.e., in the fuel transfer 

tube), the carbon steel surface could be subjected to 

galvanic corrosion that can cause significant amounts 

of material loss over small areas. Therefore, corro

sion of any carbon steel material that is welded to 

stainless steel in the penetration assembly is poten

tially significant. However, evaluation of possible 

damage to non-bellows parts used in the penetration 

assembly is beyond this project's scope. 

The primary aging mechanism that is of concern for 

bellows is TGSCC-an aging mechanism that is as

sociated primarily with austenitic stainless steel mate

rials. All bellows are made from austenitic stainless 

steel. The three conditions necessary for TGSCC are 

a sensitized material (resulting from the manufactur

ing and/or installation process), a high tensile stress 

(residual or applied), and a suitably aggressive envi

ronment (moist or wetted environment, or in an envi

ronment contaminated with small amounts of chlo

ride, fluoride, or sulfates, or high temperature). 

Higher temperature encourages SCC, which generally 

will not occur below 121°C (250'F) unless aggressive 

contaminants, such as chloride, are present. 

The expansion joints are subjected to high tensile 

stresses, particularly in the convolutions of the bel

lows, where high local residual stresses are present at 

all times. These expansion joints may be subject to 

water environments part of the time, as well as be 

contaminated with materials (fluids, chlorides, sul

fides, etc.) that encourage initiation of TGSCC and 

crack growth. Several instances of TGSCC of piping

penetration expansion joint bellows have occurred at 

nuclear plants in the last 20 years. Therefore, 

TGSCC for the piping, vent line bellows, and fuel 

transfer tube stainless steel expansion joints can be 

significant. 

The presence of contaminants in the atmosphere or 

deposited on component surfaces can lead to acceler

ated aging. These contaminants include chlorides, 

fluorides, peroxides, and sulfur compounds. There 

are several potential sources of these contaminants at 

the bellows: 

"* inadequate cleaning after completion of the ini

tial containment penetration installation. 

"* in-service inspection, testing, and maintenance 

activities without adequate care or cleanup. 

"* cleaning compounds used in other areas of the 

containment inadvertently being spread into 

containment penetration components. 

" residual material, from welding processes (smoke 

and other particulates) associated with penetra

tion components during initial installation or in

service maintenance, remain on or within pene

tration components. The specific welding activ

ity of concern is associated with the installation 

and repair of stainless steel expansion joints for 

mechanical piping penetrations. 

" decomposition of nonmetallic components such 

as thermal insulation or surface coatings. For 

example, chloride ions present in some types of 

lagging may be leached from the insulation dur

ing exposure to moisture or wetting. 

4.2 Microbiologically Influenced 

Corrosion 

Although Microbiologically Influenced Corrosion 

(MIC) has been found in stainless steel piping, the 

bellows environment is not conducive to this type of 

damage. No report of MIC damage to bellows was 

found during the search of the LER and other data

bases.

4-1
4.3 Stresses in Bellows

Stresses in the bellows, caused by normal operation 

and environmental conditions, have a direct impact 

on the existence and progression of aging-related 

degradation of the bellows. Thus, it is important to 

understand the behavior of bellows materials when 

subject to various stresses in order to design and op

erate the bellows component satisfactorily, and to 

develop methods for detecting and mitigating compo

nent degradation. 

Stresses in the bellows result from static loads, cyclic 

loads, differential movement (settlement, thermal 

expansion/contraction, or misalignment during con struction) and residual stresses from forming the bel

lows. Physical damage (during installation, mainte

nance, or operation) can cause small dents or 

scratches that lead to local stress concentrations. 

Although not directly producing stress, environmental 

conditions such as a chloride contaminant on the sur face of the bellows can exacerbate the effects of the 

stresses and result in more rapid deterioration. For 

example, chloride contaminants and residual stresses 

may combine to cause TGSCC. 

In Section 3, aging mechanisms were substantiated 

where possible through evaluation of operating his

tory. However, operating experience in benign envi

ronments is not totally conclusive. Accident envi

ronments may result in additional stresses. For 

example, Lambert and Parks (1995) tested bellows 

that were degraded in the laboratory which were leak 

tight when subjected to internal pressure with no axial 

or lateral displacements. However, when the de

graded bellows were subjected to large axial and lat

eral displacements, cracks initiated and grew in the 

bellows, and the bellows leaked. 

4.3.1 Static Loads 

Bellows are designed to withstand specific types of 

static loads. These include 

"* pressure loads from pressurization of the primary 

containment, 

"* forces from thermal movement of piping during 

system heatup and cooldown (only for hot piping 

penetrations), 

"* forces from differential thermal movement of 

structures through which the penetration extends,

"* forces from supporting the weight of penetration 

components or other adjacent components, and 

"* residual stresses in the bellows convolutions 

caused by the bellows forming process. 

Provided that these static loads are within those speci

fied for design of the penetration, they should not cause component degradation. However, there are 

two situations of further concern (detailed in Section 

4.3.3): (1) in-service differential settlement of the 

multiple structures that are in contact with the pene

tration joint, and (2) axial misalignment of the entire 

penetration or the penetration expansion during initial 

installation or subsequent maintenance activity. 

The most significant static load is the residual stresses 

in the bellows. These residual stresses, which are at 

the yield stress level in many locations, are a result of 

cold forming the bellows. Because the bellows are 

not annealed after they are formed, the residual 

stresses remain for the bellows' entire lifetime. These 

large residual stresses can combine with environ

mental conditions, such as chloride contamination on the surface of the bellows, and cause TGSCC. 

Several of the static loads described above can be 

applied periodically and are, therefore, also cyclic 

loads (with the exception of the forces from weight). 

These cyclic loads can affect many components in all 

types of penetrations. 

4.3.2 Cyclic Loads 

Stainless steel expansion bellows have been designed 

to withstand cyclic loads, and therefore, should not be 

subject to fatigue cracks, unless the number and/or 

magnitude of the loads exceeds those specified for 

design. Generally, the design basis loads (i.e., ther

mal loads from startup and shutdown) cause small 

cyclic plastic strains, but the plastic strains and ex

pected number of cycles are well below the number 

required to cause bellows degradation or failure. 

However, this may not be the case if the bellows has 

been degraded by TGSCC or subjected to mechanical 

abuse resulting in scratches in the bellows. When 

these types of degradation exist, fatigue cracking and 

failure may be accelerated by cyclic loads. 

Fatigue of metallic expansion bellows is primarily 

associated with the cyclic load stress described above. 

The cyclic loadings are from periodic application of 

pressure loads and forces due to thermal movement of 

piping transmitted through penetrations and structures

4-2
to which the penetrations are connected. The effects 

of aging (crack initiation and growth) are associated 

with the number and magnitude of the cyclic loads 

and the stress levels resulting from these loads. In 

general, the fatigue cracks will become larger on the 

bellows surface where large residual strains have al

ready caused TGSCC to occur. 

Pressure testing (ILRT) of the entire containment and 

LLRTs for individual penetrations require periodic 

application of pressure loads. Periodic piping ther

mal loads are applied to hot piping penetrations when 

the piping systems are heated up and cooled down 

and when significant changes in fluid temperature 

occur within the piping during operation. These peri

odic structure thermal loads are generally associated 

with steel containments where there is differential 

movement between the steel shell and the biological 

shield wall during heatup and cooldown of the pri

mary containment environment, which generally co

incides with power plant heatup and cooldown. 

Fatigue failures of expansion joint bellows have oc

curred at nuclear power plants. However, the root 

cause is not fatigue but damage by TGSCC that 

weakens the steel and allows fatigue to accelerate the 

crack growth. 

4.3.3 Differential Movement 

Differential movement that must be accommodated 

by the bellows can be caused by 

"* settlement of the structures to which the me

chanical penetrations are attached, 

"* misalignment during construction, and 

"* thermal expansion and contraction of the struc

tures the bellows connect. 

Because the bellows are designed to accommodate 

large relative displacements, small amounts of per

manent differential movement caused by settlement or 

misalignment will not cause a significant reduction in 

capability, unless TGSCC damage has already sig

nificantly degraded the bellows. Bellows can easily 

accommodate a permanent differential movement of 

some amount that is small in comparison to the total 

capability. 

For example, a permanent differential movement of 

0.5 in. would be small compared to an overall capa

bility of 10 in. In this example, the primary effect

would be that only 9.5 in. of additional movement can 

be accommodated. Because the travel that bellows 

are capable of accommodating is far greater that the 

design limits, a small permanent differential move

ment from construction misalignment or differential 

settlement will not negatively impact safety for DBAs 

unless other damage, such as TGSCC, has also accu

mulated. However, in the event of a SA that is far 

larger than the design level, there may be some small 

reduction in overall capacity. In the example above, 

if a bellows needed to accommodate 10 in. of dis

placement during a SA but was able to accommodate 

only 9.5 in., a small reduction in bellows capacity 

would result. 

A small permanent differential movement that the 

bellows must accommodate does not increase the 

stresses in the bellows. The convolutions of the bel

lows already have residual stresses that are at the 

yield stress level. Although small movements of the 

bellows cause additional yielding, the stress values do 

not increase. If damage to the bellows has occurred 

from another means, such as a scratch or TGSCC, 

then small movements could cause a crack to initiate. 

However, small differential movement would be a 

problem only if other significant damage such as 

TGSCC had already accumulated. In which case, 

bellows functionality had been compromised before 

the differential movement occurred. 

Misalignment adjustments are made when the bellows 

are installed new, and no TGSCC exists. Therefore, 

the issue of permanent differential movement from 

settlement or misalignment during construction will 

usually not be a concern for stainless steel expansion 

bellows. However, any relative movement that is 

accommodated by the expansion bellows should be 

evaluated to ensure that the measured movement is 

small compared to the total movement the bellows 

could accommodate. The motion in the axial direc

tion of the bellows, as well as lateral to the axis of the 

bellows, should be evaluated. 

Thermal expansion and contraction of the structures 

the bellows connect will also cause differential 

movement. This movement is quite small and will 

not cause a premature failure unless other damage to 

the bellows, such as TGSCC, has already occurred. 

This type of loading, which occurs during startup and 

shutdown cycles, is detailed in Section 4.3.2.

4-3
4.4 Degradation Related to 

Installation, Maintenance, 

or Operation 

Degradation during penetration assembly or compo

nent installation is not in itself an aging mechanism, 

but it can adversely affect the longevity of the pene

tration. Installation degradation normally is con

trolled or prevented through use of approved proce

dures, controls, and standards during installation or 

maintenance. Equipment damage usually occurs as a 

result of failure to use or properly implement proce

dures, or, in a limited number of cases, from improper 

equipment design. Examples of such degradation 

include accidentally scratching or denting expansion 

joint surfaces. 

Incidental mechanical degradation can, by itself, 

cause a functional failure of a component. Mechani

cal damage is considered because it could occur for 

some types of penetrations and may not be detected

and reported. This damage might then lead to accel

erated aging. 

4.5 Radiation Effects 

During normal operating conditions in containment 

areas, the normal radiation dose rates (all sources) 

may range from about 0.1 to 1000 Gy/hr (0.01 to 100 

rad/hr), depending on plant type (EPRI, 1986). Al

though some bellows are located in high-dose spaces, 

the stainless steel used in their construction has a high 

damage threshold. Metals generally have damage 

thresholds on the order of 1 x 10' 9

n/cm' (approxi

mately 2.5 x 108 Gy or 2.5 x I0 t rad) (NASA, 1969). 

A conservative threshold dose of 1 x 101i n/cm 2

, or 3 

MGy (300 Mrad) gamma, can be set for metals used 

in penetration components. Below this dose, few or 

no aging effects should occur in such materials. Few 

locations within primary containment would experi

ence this level of exposure. Therefore, radiation ef

fects are not a concern for the metallic bellows.

4-4
5. INSPECTION AND EVALUATION OF BELLOWS

5.1 Requirements for Continued 

Operation during Current 

License Period 

Maintenance, surveillance, and condition monitoring 

are performed to ensure that the characteristics or at

tributes of components that are essential for operation 

are maintained. The activities performed during the 

operation, maintenance, testing, and condition moni

toring of piping penetrations may be generally classified 

as 

"* visual or physical inspection, 

"* leak rate testing, 

"* cleaning, and 

"* component repair or replacement. 

In addition to individual maintenance, surveillance, and 

condition monitoring techniques, programs resulting in 

the management of aging effects on containment pene tration components were considered. For example, maintaining compliance with 10 CFR 50.49 (1983) can 

be considered an aging management program for quali fied containment penetrations because the net effect is 

to maintain adequate post-accident component func

tionality through establishment of a qualified-life stan

dards for components. All nuclear power plants have 

existing environmental-qualification programs that in

clude some portion of the total penetration population within their scope. Similarly, a structured bellows 

LLRT program based on plant-specific operating expe rience, and coupled with root cause analysis and appro priate corrective action, could be considered an aging 

management program. Note that some existing plant 

activities not specifically focused on containment pene tration aging management, may be credited with ad

dressing aging effects: 

Preventive Maintenance Programs 

Although periodic inspection of expansion joint bellows 

is performed under the surveillance programs (Section 

5.1.1), some periodic cleaning and refurbishment or 

replacement of these bellows may be performed under 

the preventive maintenance program. 

Maintenance Rule 

Effective July 1996, nuclear power plant owners have 

implemented the Maintenance Rule (10 CFR 50.65,

1996). Because of their safety-related function, con

tainment penetrations are included within the scope of 

the Maintenance Rule-a performance-oriented rule 

that establishes reliability goals for both systems and 

individual components. The investigation of mainte

nance preventable functional failures (MPFFs) will sup

port aging management for stainless steel bellows. An 

MPFF for a piping penetration may be a crack in the 

bellows that results in it being unable to meet the leak

age acceptance criteria during a LLRT. The program 

requires that MPFFs be monitored, that the root cause 

of the failure be investigated, and that a corrective ac

tion plan be developed and implemented. 

Table 5-1 summarizes programs used to detect damage 

to bellows. 

5.1.1 Surveillance Programs 

Surveillance programs provide for visual inspection of 

primary containment accessible surfaces, including the 

mechanical penetrations, once each operating cycle to 

satisfy Technical Specification requirements. Bellows 

are inspected for discoloration, scratches, cracking, leaking, and any other abnormal appearance. Addition

ally, the presence of dirt, dust, contamination, mois

ture/humidity, chemicals, or corrosion products may indicate degradation. Inspection results are recorded 

and corrective action is taken as necessary. 

Many of the nuclear power plants commissioned after 

1980 have primary containment vessels and mechanical 

penetrations designed in accordance to Section III, Class MC and Class CC of the ASME Boiler and Pres

sure Vessel (B&PV) Code (ASME, 1992a). In 1986, 

the ASME amended Section XI of the B&PV Code, 

adding Subsection IWE, to include in-service inspection requirements for containment vessels, including me

chanical penetrations. Although Subsection IWE is a 

formalization of the Surveillance Programs already be

ing performed by operating nuclear power plants, the 

inspection requirements are significantly more rigorous. 

In 10 CFR 50.55a(b) (2)(ix - x) (1997), the NRC for

mally required licensees to use the 1992 edition with the 

1992 Addenda of Subsection IWE in their nuclear 

power plant ISI programs. The requirements of Sub

section IWE do not have to be implemented until Sep tember 9, 2001. 

Although periodic visual inspections provide useful 

observations for many penetrations, they have not been

5-1
Table 5-1. Aging Management Programs for Bellows

N. Aging Management No. Activity When Activity is PlannedfFrequency of Activity 

Common Programs

Once each oneratine cycle C I -*-- fljr.fluthI I r----c -. ' -

10 CFR 50, Appendix J 

leak rate testing

ASME Section XI 

In-service Inspection 

Program, Article IWE 

Preventive Maintenance 

Program 

Periodic plant walkdowns

ILRT: three times in a 10-year period 

LLRT: every refueling outage

Refer to ASMIE B&PV Code, Section XI, Article IWE-2000, "Examina

tions and Inspection" (ASME, 1992).

Frequency based on written plant procedures and depends on component.

DLesA ;omo nA • r, oro ant restrt ater refueling etc.

Less Common Programs

6 Settlement Trending Settlement monitoring based on accepted methods (EPRI, 1994a; EPRI, 

Program 1994b). When settlement approaches design or acceptance criteria, a re

evaluation is performed. 

Detailed fatigue analysis Performed when significant cracking is found or suspected. 

using current conditions

effective in detecting SCC degradation in stainless steel 

bellows. Most degraded bellows are found during leak 

tests. 

5.1.2 Plant Walkdowns 

Periodic walkdown inspections of nuclear power plant 

areas are performed daily and before plant restart. The 

areas included in plant walkdowns include those acces

sible portions on the outside of the containment where 

many of the penetrations exist. The purpose of these 

routine walkdowns is to notice any abnormal physical 

conditions (such as rust, dents, cracks, deformation, 

discoloration, insulation damage, vibration, leaks, and 

spills) and any unusual noises. 

Plant walkdowns are also performed after refuelings 

and other planned plant shutdowns and just before or 

during plant restart. The purpose of these walkdowns is 

the same as that for the daily walkdowns; however, 

during pre-start walkdowns, personnel must pay par

ticular attention to areas or components that were modi

fied, refurbished, or replaced during the outage and to 

general "housekeeping" conditions. These pre-start 

plant walkdowns are performed within the primary 

containment, as well as other plant areas. Therefore, all 

reasonably accessible portions of the penetrations are 

observed during these walkdowns. 

All abnormal conditions identified during these walk

downs are documented and investigated and corrective

action is taken as necessary. The periodic plant walk

downs are particularly useful for identifying corrosion 

and deformation of the mechanical penetration compo

nents. However, walkdown inspections are unlikely to 

detect cracks or pinholes in bellows caused by TGSCC. 

5.1.3 Leakage Testing, 10 CFR 50, 

Appendix J 

The primary containment, as a whole, and individual 

penetrations are periodically leak tested in accordance 

with Technical Specification requirements and the 

regulatory requirements of 10 CFR 50, Appendix J 

(1995). Although this leak testing is also part of the 

Surveillance Program discussed in Section 5.1.1, it is 

considered to be a separate program. The testing fre

quency can be generalized as three times during a 10

year period for the ILRT of the entire primary contain

ment and every refueling outage for LLRTs of individ

ual penetrations. An LLRT of a bellows pressurizes the 

volume between the plies and the associated test con

nection piping and valves, while the ILRT checks the 

containment and penetration as a whole. The ILRT and 

LLRT have specific maximum allowable leakage rates 

that are specified in 10 CFR 50, Appendix J, the Tech

nical Specifications, and the test procedures. When 

leakage exceeds the allowable rates, an investigation of 

the leakage cause is performed and corrective action is 

taken as necessary. For bellows, corrective action gen

erally will require replacing the entire bellows. Fol5-2

e rom am I r)nce each o eratimy cvcle

I

I

I

|

I er rme . Z r

C....;11. Dr r r

ýurve anclll
lowing the repair, a retest is performed to ensure that 

the bellows is leak tight. 

In 1992, the NRC issued modifications to 10 CFR 50, 

Appendix J to provide options for performing these 

leakage tests. In general, this option would allow the 

frequencies of the ILRTs and LLRTs to be reduced 

based on successful experiences with prior testing and 

the establishment of leakage test performance goals. 

Although the option to reduce test frequency is, to some 

extent, a result of generally good performance with re

gard to the results of ILRTs and LLRTs over the past 20 

years, it also reflects a concern for the aging effect on 

the primary containment and mechanical penetrations 

caused by frequent testing. 

The LLRTs associated with mechanical penetrations 

that contain stainless steel expansion joints made with 

two-ply bellows typically have included interply leak

age tests for these bellows. Now it is clear that, because 

of bellows manufacturing methods, this type of test is 

invalid and can provide nonconservative leakage rate 

results (Brown and Tice, 1993; Berg and Brown, 1995). 

The NRC notified licensees of this concern (NRC, 

1992a). Complete resolution of this issue is plant

specific because not all plants have this type of two-ply 

bellows. 

5.1.4 Settlement Trending Program 

The Settlement Monitoring and Trending Program is 

identical to the programs described in the containment 

industry reports (EPRI, 1994a; EPRI, 1994b) to verify 

primary containment settlement. A similar program is 

used for the spent fuel pit foundation. The purpose is to 

check for differential settlement, which may increase 

loads on the fuel transfer tube penetration components. 

5.1.5 Detailed Fatigue Analysis Using 

Current Condition 

Fatigue potentially is a significant degradation mecha

nism for some components in each type of mechanical 

penetration. For early designs, an explicit fatigue 

evaluation was not required; however, if the contain

ment is a Section III/Subsection NE design (ASME, 

1992b), then fatigue analysis was required. Fatigue 

analyses that were performed for bellows used an unde

graded component basis. This program requires an up

dated fatigue analysis when significant degradation is 

found for any of the components. The ASME B&PV 

Code (1992a), Section XI, In-service Inspection, in

cludes an appendix to address fatigue when the analysis 

includes the calculation of cumulative fatigue usage

factor. This nonmandatory appendix provides guidance 

for usage factors that approach the limiting value of 1.0 

or other fatigue concerns. In the event that a bellows is 

degraded by TGSCC, however, the fatigue curves in the 

code are very likely not conservative. Fatigue tests on 

low-carbon steels with corrosion damage show a sig

nificant reduction in the number of cycles before failure 

(Shigley, 1983; Cherry and Smith, 2001a; Bruneau and 

Zahrai, 1997). Although the reduction in cycles caused 

by TGSCC has not been quantified, a similar result will 

certainly occur. 

5.1.6 Cleaning 

Preventive cleaning of expansion joint bellows may be 

performed coincident with periodic inspections or com

ponents or during related maintenance. The following 

substances have been found generally acceptable by 

various manufacturers for cleaning of penetration com

ponents. In all cases, the manufacturer's guidance 

should be consulted before using any substance for 

cleaning penetration components (Conax, 1987ab; Co

nax, 1987bc). 

"* "freon" TF (trichlorotrifluoroethane) 

"* "freon" TWD 602 ("freon" TF with water and de

tergent) 

"* mild detergents and soaps 

"* denatured alcohol ("isopropyl") 

These substances generally are not recommended for 

use on penetrations:

S 

S 

0 

0 

0 

0 

0 

0 

0 

0

1, 1, 2, 2 - tetrachloroethane 

1, 1, 1 - trichloroethane (e.g., "vythene") 

acetone 

alcohols (except as noted above) 

benzene (benzol) 

carbon tetrachloride 

cellosolve solvent 

chlorobenzene 

chloroform 

cyclohexanone

Cherry, J.L. and Smith, J.A. 2001 (in press). Capacity of Steel 

and Concrete Containment Vessels with Corrosion Damage. 

Albuquerque, NM: Sandia National Laboratories. 

b Conax Buffalo Corporation, 1987a, Proprietary Report, IPS-981, 

Rev. B, March 1987. c Conax Buffalo Corporation, 1987b, Proprietary Report, IPS-725, 

Rev. H, July 1987.

5-3
"a esters (e.g., ethylacetate) 

"* "freon 22" 

"* "freon" TA ("Freon" TF + acetone) 

"* "freon" TE ("Freon" TF + ethanol) 

"* heptane 

"* kerosene 

"* methylene chloride 

"* methyl ethyl ketone 

"* mineral spirits (e.g., "varson") 

"* most regular gasolines 

"* naphtha 

"* tetrachloroethylene ("perclene") 

"* toluene 

"* trichloroethylene (a degreasing solvent) 

"* turpentine 

"* xylene (xylol) 

5.1.7 Repair 

Degradation of bellow elements and the inability to 

effectively repair these components has caused the en

tire expansion joint to be replaced for feedwater and 

mainstream piping penetrations at several nuclear 

plants. These replacements are difficult and costly be

cause of poor access around these large-diameter piping 

penetrations. Some of the replacements have longitudi

nal closure welds (through the expansion joint spacer 

and bellows areas). Without employing these expansion 

joint closure welds, it would have been necessary to 

remove an entire section of process pipe and all other 

penetration components to replace the expansion joint. 

This clearly is impractical for many physical configura

tions. 

5.2 License Renewal Requirements 

The current requirements for the technical content of a 

license renewal application are described in 10 CFR 

54.21 (1995) and Nuclear Energy Institute (NEI) report 

NEI 95-10 (NEI, 1995). A license renewal application 

must contain: 

I. an Integrated Plant Assessment (IPA), 

2. a list of CLB changes during NRC application re

view, 

3. an evaluation of time-limited aging analyses 

(TLAAs), and 

4. an FSAR supplement. 

An IPA must

1 for those SSCs within the scope delineated in 10 

CFR 54.4 (1996), identify and list those structures 

and components subject to AMR; 

2. describe and justify the methods used in item I 

(scope determination) of the IPA; and 

3. for each structure and component identified in item 

1 of the IPA, demonstrate that the effects of aging 

will be managed so that the intended function(s) 

will be maintained for the period of extended op

eration. 

5.2.1 Aging Management Review 

The revised version of the license renewal rule, 

(10 CFR 54, 1996), discusses the concept of an "aging 

management review." The IPA must list structures and 

components subject to an AMR, and must demonstrate 

that aging concerns are being managed. It is necessary 

to maintain the CLB in a way so that there is an accept

able level of safety during the period of extended op

eration. Furthermore, the TLAAs required for some 

penetrations, (10 CFR 50.49, 1983; 10 CFR 54.21, 

1995) which form the basis of a plant operator's conclu

sion regarding the capability of those penetrations, must 

consider the effects of aging and be based on explicit 

assumptions defined by the current operating term of 

the plant. The specific methods to be used by the plant 

operator in meeting these requirements are not man

dated; each operator can select the method(s) or tech

nique(s) most appropriate for their individual AMR. 

The 10 CFR 54 (1996) requires that the IPA list those 

structures and components subject to an AMR and 

demonstrate that the effects of aging on the functionality 

of such structures and components will be managed to 

maintain the CLB so that there is an acceptable level of 

safety during the period of extended operation. Fur

thermore, the TLAAs required by 10 CFR 54.21 (1995), 

which form the basis of a licensee's conclusion regard

ing the capability of SSCs, must consider the effects of 

aging and be based on explicit assumptions defined by 

the current operating term of the plant. The methods by 

which the licensee meets these requirements may vary; 

each licensee can select the method(s) or technique(s) 

most appropriate for their individual AMR. 

5.2.2 Special Considerations for Stainless 

Steel Bellows 

The problems with interply leak testing of two-ply 

stainless steel bellows were identified in NRC Informa

tion Notice 92-20 (1992). This issue has been explored

5-4
in greater detail by Brown and Tice (1993) and Berg 

and Brown (1995). The interply pressurization method 

for leak testing two-ply bellows provides invalid and 

nonconservative results because the air path between 

the bellows from one end to the other and around the 

circumference is essentially blocked by the stainless 

steel mesh between the plies. The function of the mesh 

was to hold the spacing between the two plies; however, 

it also acts as a gas seal, at least for an extended period 

of time. To perform a valid leakage test across the ex

pansion joint bellows, a pressure must be applied to the 

exterior or interior surface and the level of pressure 

decay measured over a period of an hour or two. This 

test can be accomplished with a permanent modification 

to the penetration. A second bellows expansion joint is 

placed outside the existing joint, and the leakage test is 

performed by pressurizing the area between the primary 

and secondary (or outer) expansion joint. Localized 

permanent test enclosures around bellows can be de

signed and used for leak testing as described by Brown 

and Tice (1993) and Berg and Brown (1995).

Valid LLRTs for these two-ply bellows expansion joint 

penetrations, as well as single-ply bellows, can be ac

complished with a temporary test fixture. This fixture, 

installed inside the containment, would result in testing 

all pressure-retaining penetration components, including 

the expansion joint, simultaneously. Although it obvi

ously is possible to build such fixtures for testing as 

noted by Brown and Tice (1993) and Berg and Brown 

(1995), it is time consuming and expensive, especially 

at plants with approximately 20 penetrations of this type 

for certain containment designs. A reusable test assem

bly that can be installed and removed rapidly would be 

preferred for periodic penetration LLRTs. Where space 

does not allow use of local bellows test assemblies, the 

replacement of a close-rolled two-ply bellows with two 

independent, single-ply bellows is one of the few alter

natives remaining to satisfy the LLRT requirements of 

10 CFR 50, Appendix J (1995).

5-5
6. REGULATORY AND RISK SIGNIFICANCE 

OF DEGRADED BELLOWS

This section explores the potential regulatory and risk 

significance of degraded bellows in the context of 

regulatory requirements for DBA and in the context of 

risk goals related to SAs. The primary focus is on 

BWRs with Mark I containments because they typically 

have a large number of two-ply bellows associated with 

containment penetrations and a single-ply bellows asso

ciated with each of the vent line pipes. However, the 

insights developed here for Mark I containments can be 

adapted to other containments that employ bellows. 

The section first examines the decision logic, represen

tative of at least some part of the industry, relating to 

the identification and replacement of two-ply bellows. 

Insights from SNL's containment bellows testing pro

gram are then reviewed in the context of industry logic 

for bellows replacement. Finally, a number of issues 

related to the risk significance of containment bellows 

are reviewed. 

6.1 Representative Industry Logic 

for Bellows Replacement 

In typical installations of new bellows assemblies, initial 

LLRT values are zero or very small (Figure 6-1). 

Trending of LLRT data at some domestic BWR plants 

indicates that the measured leak rate remains stable for 

a period of time, suggesting that the condition of the 

bellows has remained essentially unchanged. After a 

period of 5 to 15 years, an increase in leakage rates has 

been observed at some drywell penetrations. Initially 

these leak rates tend to increase linearly with time, indi

cating that whatever is causing the rates to change is 

progressive. 

Figure 6-2 is a simplified logic diagram (Brown and 

Tice, 1993) for replacement of penetration bellows. 

Implicit in this logic is an assumption that an interply 

LLRT is an effective leak detection method while ac

knowledging that an interply LLRT is not a reliable leak 

rate measurement method. It is clear from Figure 6-1 

that the onset of leakage can be detected when the 

measured leak rate exceeds some threshold value, al

though Brown and Tice (1993) do not suggest a thresh

old value. However, examination of Figure 6-1 sug

gests that a measured leak rate of -0.05 m3

/hr (1.20 

m3

/day) might be adequate for detecting departure from 

a stable state.

Figure 6-2 shows that the bellows may be judged to be 

an acceptable pressure boundary even if one or both 

plies are known to be degraded and leaking. When both 

plies are degraded, all observable flaws in the outer ply 

are mapped and measured as input to a calculational 

methodology, which involves predicting the gas flow 

rate through crack-like orifices. In an effort to account 

for radial expansion from pressure (which causes the 

cracks to open slightly), a finite element model is used 

to simulate the bellows response to pressure loads. 

Empirical adjustments are made to reflect known flow 

rates through known flaws. The calculational method

ology is used to predict bellows response during a 

DBA, and as such, is the foundation on which a deci

sion is made to replace (or defer replacement). 

The logic in Figure 6-2 is applied to each penetration 

bellows separately. No adjustments are made to situa

tions where multiple bellows are simultaneously in a 

degraded state. Such common mode or concurrence 

states likely existed at Dresden where a number of bel

lows were replaced over two consecutive refueling out

ages. 

6.2 Sandia National Laboratories 

Testing of Like-New and 

Degraded Bellows 

SNL has tested containment piping bellows that were in 

like-new condition (Lambert and Parks, 1994) and in 

degraded condition (Lambert and Parks, 1995). All 

bellows were conducted to full compression (or exten

sion) with an appropriate lateral displacement. Table 

6-1 summarizes the tests conducted with like-new bel

lows subjected to SA conditions. These tests spanned 

conditions ranging from ambient pressure and tem

perature to elevated pressure and temperature. Table 

6-2 summarizes the tests conducted with degraded bel

lows. The degraded bellows were subjected to acceler

ated corrosion environments followed by combinations 

of elevated pressure, temperature, and axial (compres

sive) and lateral deformations. 

Table 6-3 summarizes key insights from SNL's testing 

of containment bellows. Under DBA conditions, the 

expected deformation on any individual bellows is 

-10% of full compression (or extension). Under DBA 

conditions, SNL's experiments showed no leakage from

6-1
31 fL h rn2 h 

L 20•57

1 15: _4 

7i 28 14: 

5 1.4

"73 '74 '75 *76 "77

BELLOWS 

REPL.kCED

"78 '79 '80 '81 '82 '83 '84 '85 '86 '87 '88 '89 '90 '91

TEST DATE - YEAR 

Figure 6-1. LLRT Flow Rate vs. Time (from Brown and Tice, 1993).

like-new bellows. Also, the bellows with laboratory

induced TGSCC (degradation not extensive enough to 

cause an initial leak), did not leak when compressed (or 

extended) about 10% to correspond to a DBA. 

These three categories of tests covered in Table 6-3 

demonstrate that if a bellows with TGSCC does not leak 

initially, then the relative movement caused by a DBA 

probably would not cause a leak to develop. However, 

only three of the bellows tested that had TGSCC dam

age did not leak initially. Obviously, three tests do not 

comprise a statistical sample. Probably the most im

portant factors are the crack growth rate and the associ

ated increase in the leak rate as the bellows in the test 

were displaced. Up to about 10 or 20% compression, 

the leak growth rate increased, but at a stable rate. 

Therefore, if a bellows was not leaking initially, a de

sign DBA would be unlikely to cause a leak. In the 

unlikely event that a leak did develop, it can be stated 

with very high confidence that the leak rate would be 

small enough to pass the 10 CFR 50, Appendix J (1995) 

leak rate criteria. Consequently, DBA criteria are de

monstrably satisfied. 

No such conclusions can be drawn when the bellows is 

corroded to the point that they leak initially (before be

ing compressed or elongated). The data indicate that 

the initial leaks increase, but the measurement tech

niques are not adequate to quantify if DBA criteria are 

satisfied. The database also is not adequate to validate 

industry models that predict bellows performance under 

DBA conditions from a prior characterization of dam

age with the bellows in the neutral position.

Table 6-3 shows that any bellows, like-new or de

graded, can develop leaks for SA where the bellows are 

completely compressed (or extended). The leaks take 

the form of cracks, localized circumferential tears, or 

tears where a convolution buckles. The welded seam 

appears to be a preferred site for buckle tears. Leakage 

increases with the degree of initial degradation to the 

bellows. Any bellows failure at full compression would 

generally be classified as a small leak (-10 times DBA 

criteria). The IPA treatment of containment fragility 

suggests that containment deformations capable of fully 

compressing a bellows will have a significant likelihood 

of inducing other failures that will produce much larger 

hole sizes. Table 6-1 also suggests that expansion bel

lows may develop more risk-significant leaks when 

fully extended. The available data do not comprise a 

statistical sample, and there are no data for degraded 

expansion bellows. 

The pressure history on a Mark I drywell has a similar 

response for both DBA and SA. The drywell pressure 

rises rapidly during the duration of the blowdown event 

(minutes) and then quickly decays. The magnitude of 

the pressure rise is controlled by the rate of blowdown 

(in competition with flow through the vent lines into the 

wetwell); consequently, the pressure rise could be larger 

in a SA where the lower ruptures with the system at 

pressure. For both DBA and SA, the drywell pressure 

will again increase slowly over a period of several 

hours.

6-2

A 

K 

R 

A 

T 

E
Figure 6-2. Representative of Industry Logic for Replacement of Bellows.

6-3
Table 6-1. Experiment Results from Like-New Containment Piping Bellows Subjected to 

Severe Accident Conditions

Leak Tight at Full Compression 

Test IDm 1() Direction Pressure Temperature (Elongation) of First Loading (Dis

Specimen ID~ placement of First Leak If It Occurs) 

I- I1-CL Compression Ambient pressure Leak tight 

A-8-2-1 Ambient temperature 

I-12-LC Compression Ambient pressure Leak tight 

A-8-2-2 Ambient temperature 

1-3-SC Compression 0.52 MPa Leak tight 

B-8-2-1 Ambient temperature 

I-4-SCT Compression 0.45 MPa Leak at 86% compression 

B-8-2-2 213 0C 

I-2-SCT Compression 1.03 MPa Leak tight 

B-12-2-1 216 0C 

I-10-SCT Compression 0.47 MPa Leak at 91% compression 

B-12-2-1 210 0C 

1-9-SC Compression 0.50 MPa Leak at 99% compression 

B-12-2-2 Ambient temperature 

HI-19 Compression 1.03 MPa Leak tight 

VL-2 213 0C 

11-17 Compression 0.51 MPa Leak tight 

PP-2 204°C 

I- 13-SE Elongation Ambient pressure Leak at 88% elongation 

A-8-1 -e Ambient temperature 

I-14-SE Elongation Ambient pressure Leak tight 

A-8-2-e Ambient temperature 

1-15-SE Elongation Ambient pressure Leak tight 

B-8-1-e Ambient temperature 

1-16-SE Elongation Ambient pressure Leak tight 

B-8-2-e Ambient temperature I

t I S T.ct Ii) flierrmntann

CL Axially compress the bellows until all convolution roots are fully compressed. If the bellows are still leak tight, apply lateral defor

mation until a tear in the bellow occurs. No internal pressure will be applied during this test. 

LC Apply lateral offset until the bellows becomes unstable (i.e., very small increase in lateral load produces relatively large lateral defor

mation). If the bellows is still intact, apply axial compression until all convolutions are fully compressed. If the bellows is still intact 

after being fully compressed, apply additional lateral deformation until a tear develops in the bellows. No internal pressure will be 

applied during this test. 

SC Simultaneously apply internal pressure, axial compression, and lateral deformation. Internal pressure, axial compression, and lateral 

deformation shall be increased linearly so that the internal pressure reaches the maximum test pressure level when the total applied 

lateral deformation is 2 inches when the bellows are fully compressed. If the bellows are still intact after being fully compressed, re

verse the applied test conditions by removing axial and lateral deformation, as well as internal pressure at the same rate at which each 

was originally applied. Continue unloading until either a crack develops or until all originally applied displacement and pressure 

have been removed. 

SCT Same as SC except test temperature is 425°F (±25°F) throughout the specimen for the entire test. 

SE Simultaneously apply axial elongation and lateral deformation. Axial elongation and lateral deformation shall be increased linearly 

so that total applied lateral deformation is 2 inches when the bellows is fully elongated. If the bellows are still leak tight after being 

fully elongated, reverse the applied test conditions by removing elongation and lateral deformation at the same rate at which each was 

originally applied. Continue this process until either a crack develops or until all originally applied deflection has been removed. No 

internal pressure will be applied during this test. 

(2) Specimen ID numbers denote the bellows construction. The first letter (A or B) indicates convolution depth (A = 0.5 in. and B= 1.25 in.). 

The first number indicates the number of convolutions (8 or 12), and the second number indicates the number of plies. Specimens with A 

or B designation are single bellows: those with VL (vent line) and PP (process piping) are universal bellows.

6-4

s
Table 6-2. Experiment Results from Degraded Bellows Subjected to Severe Accident Conditions 

Test ID&' Surface Condition Compressive Phase Observations Measured Interply 

Specimen IDt 1 2 Pretest Leak CompressivePhaseObservations _Leak Rates 

1-7-SC * Severely corroded with * Pretest cracks began to grow at 10% compression. 

A- 12-2-1 numerous small cracks and * Cracks merged and expanded at 25% compression. 

a few larger ones. * Severe damage to the bellows with numerous large 

9 Pretest leaks > App J tears at full compression. Wire mesh clearly visi

ble. Severe cracking in other places. Many fail

ures along welded seam. 

* Estimated flow area from archival photos 

-1.4x10-3 m2 

e Estimated volume flow rate at 293 K 

Vý = A.yW = 1686 m3

/day 

I-8-SCT a Corrosion visible to naked 9 Leak rate increased with additional deformation Compression Leak Rate 

A-12-2-2 eye, but cracks only visible from -25% compression until 50-75% compres- (sm3

d) 

only with magnification. sion. @0.14 MPa 

"* Pretest leaks < App J 0 Wire mesh clearly visible in a number of locations. 0.00 0.12 0.21 0.40 

" Pretest leak -0.12 slpm Cracking in other places. 0.26 1.87 

-8x10- m3 /day 0 Estimated flow area from archival photos 0.31 5.83 

-2.2x10-e m' 0.36 12.2 

e Estimated volume flow rate at 293 K 0.41 34.7 

V = =265 m'lday 1.00 39.6 

Max leak 73.4 sm 3

d after 

return to full extension 

1-5-SCT 9 Well-defined cracks easily 0 Leak rate increased with additional deformation Compression Leak Rate 

C-6-2-1 visible, from -25% compression until 50-75% compres- (sm 3

d) 

"* Pretest leak < App J sion. @0.14 MPa 

"* Pretest leak -0.005 slpm * Wire mesh clearly visible in a few locations. 0.00 0.01 

0.11 0.04 

"* Pretest leak -0.12 slpm * Estimated flow area from archival photos -2.2x 10- 0.21 0.13 

M3"3 m32day21 18 m1 

-0.3X08 in/day 0.35 0.40 

* Estimated volume flow rate at 293 K 0.40 1.68 

V =AfyjRiT =265m'/day 0.45 7.08 

0.55 12.1 

0.61 24.5 

0.64 41.6 

0.76 54.6 

1.00 54.6 

Extension after full com

pression 

Compression Leak Rate 

(sm 3d) 

@0.14 MPa 

0.94 59.1 

0.60 83.6 

I-1-SCT * Visible corrosion with * Small leak at last increment to full compression. Compression Leak Rate 

A-8-2-4 minute cracks visible with (sm 3

d) 

magnification. 0.014 MPa 

* No pretest leak. 0.00 0.00 

1.00 0.18 

Extension after full comp. 

Compression Leak Rate 

(sm 3

d) 

@0.14 MPa 

0.85 

0.71 24.5

6-5
Table 6-2. Experiment Results from Degraded Bellows Subjected to Severe Accident Conditions 

(continued)

Table 6-3. Insight from Sandia National Laboratories' Testing of Like-New and Degraded 

Containment Bellows 

Like-New Bellows and Degraded Bellows Degraded Bellows 

Degraded Bellows Initial Leak Rate < Initial Leak Rate > Not Leaking Initially Appendix J Criteria* Appendix J Criteria* 

DBA * No measurable LLRT e Interply leak rate meas- * Interply leak rate meas Conditions leaks. urements (LLRT) suggest urements (LLRT) suggest 

< 10% Corn- * DBA criteria demonstra- that small deformations that the small deforma

pression bly achieved, increased the leak rate. tions increased the leak 

Not possible to quantify if rate. 

DBA criteria are met e Not possible to quantify if 

(ILRT leak rates cannot be DBA criteria are met 

determined from the (ILRT leak rates cannot be 

LLRT). determined from the 

LLRT). 

Industry practice leaves * Industry practice is to * Industry practice is to 

bellows in service, leave bellows in service if leave bellows in service if 

projected leak rate through projected leak rate through 

either ply is less than 10 either ply is less than 10 

CFR 50, Appendix J CR 50 Appendix J (1995) 

(1995) criteria, criteria. 

Often, bellows must be 

replaced. Severe * Possible tears initiate near a Leak rates increase as a Leak rates increase as Accident full compression or bellows is compressed. bellows is compressed. Conditions extension. 9 Estimated leak rates - 12 * Estimated leak rates -80 

Up To 100% * Possible leaks develop times DBA criteria when times DBA criteria when 

Compression near full compression or bellows is fully corn- bellows is fully com

extension, with larger leak pressed (based on a small pressed (based on a small 

rates in expansion bellows number of test results). number of test results). 

than in compression bel

lows. 

"10 CFR 50, 1995

6-6

Test ID..( Surface Condition Measured Interply Specimen IDW Pretest Leak Compressive Phase Observations Leak Rates 

II-20-SCT a Visible corrosion with 0 No leak to full compression. 

VL-1 minute cracks visible with * No leak for two full cycles Single-ply Vent line bem- magnification s gNificati l Cracks became wider without leaking by end of Vetlnebl No pretest leak. test. 

low 

I-6-SCT * Pretest pitting but no visi- * No leak prior to full compression. 

D-6-2 ble surface corrosion or 

cracking

0 No pretest leak.

Notes: 

(1) Same as Table 6-1 

42) Specimen ID numbers denote the bellows construction, The first letter (A, C, or D) indicates convolution depth (A and C = 0.5 in, and 

D=1.25 in.). The first number indicates the number of convolutions (6, 8, or 12), and the second number indicates the number of plies. Specimens with A designation are single bellows; those with C, D, or VL (vent line) are universal bellows.
This behavior is significant with regard to potential ra

dionuclide release from the containment for two rea

sons. First, any leaks in the containment bellows are 

insignificant compared to the flow into the wetwell 

where the fission products will be scrubbed. Second, 

the period when short-term drywell pressures are high is 

limited, which further reduces the release from the con

tainment. Furthermore, for the more common sequences 

where there is power in the plant, the operators have the 

means to vent the wetwell to reduce long-term pressures 

in the drywell, reducing even further any possible fis

sion product release. SNL also notes that the core is not 

severely damaged in a DBA, so that the fission product 

release to the drywell is also quite limited. 

6.3 Severe Accident Risk 

Perspective Regarding Bellows 

Leakage 

A number of probabilistic risk assessments (PRA) have 

been performed that have quantified the potential risk 

from various accident scenarios. In particular, the 

NUREG/CR-l 150 (NRC, 1990) study has been re

viewed to determine what information from this previ

ous study can be applied to the bellows leakage issues. 

Since bellows are used primarily in steel containments, 

the current discussion will focus on the two 

NUREG/CR-1 150 containments that are of steel con

struction, the Peach Bottom plant (BWR) and the Se

quoyah plant (PWR) (Payne et al., 1990; Gregory et al., 

1990, respectively). 

Bellows leakage is addressed in terms of containment 

failures, source terms, and consequences that were in

cluded in the NUREG/CR-1 150 study (NRC, 1990). 

Three primary aspects of the release should be ad

dressed 

"* size of the leak, 

"* source term (which radionuclides and what quanti

ties) to the environment, and 

"* timing characteristics associated with the release. 

If a bellows leak were to occur because of TGSCC 

damage, the leak will generally be one or more line 

cracks that form in the convolutions of the bellows. In 

terms of the NUREG/CR-1 150 PRA (NRC, 1990), the 

size of a bellows leak would probably lie somewhere 

between nominal leakage, applied in the NUREG/CR

1150 study during times of no containment failure 

(NCF), and a small leak. As applied in NUREG/CR1150, the NCFs nominal leakage was estimated at 0.1% 

per day for five days, and the small leak was estimated 

as a 1 ft2 hole. Based on the limited data available, it 

appears that any leak through a bellows would be much 

closer to the nominal leakage category than it would be 

to the small leak category. In any event, it is assumed 

that the NUREG/CR-1150 small leak (I ft2 hole) is an 

upper bound, with a large margin for error, for a worst

case bellows leak. 

The magnitude of the source term release is dependent 

on many factors, including: 

"* the accident initiator; 

"* the amount of material released from the reactor 

core; 

"* the core release that is retained in vessel; 

"* the vessel release that is retained in containment; 

and 

"* the decontamination by mitigating factors such as 

suppression pools, sprays, ice condensers, natural 

agglomeration, etc. 

It is beyond the scope of this discussion to investigate 

all of these features of the source term. The feature that 

is central to the bellows leakage issue involves the con

tainment retention, which can be investigated by focus

ing on containment failure size and timing. The re

maining factors are indirectly included in the 

distribution of the source terms within the range that is 

investigated herein. 

For timing considerations, because a bellows failure 

could exist throughout the entire accident progression, 

source terms for early containment leaks (ECLs) are 

considered. In addition, to attempt to account for 

smaller leaks (with lower leak rates), source terms that 

involve late containment leaks (LCLs) are considered 

because they feature longer containment retention times. 

Finally, source terms involving NSF that were modeled 

in NUREG/CR- 1150 (NRC, 1990) with the 1% per day 

leakage rate are considered. Note that all source terms 

discussed here will include only those core damage ac

cidents that proceed to vessel breach (there were some 

accidents modeled in NUREG/CR- 1150 in which core 

damage was arrested in vessel). As mentioned earlier, a 

bellows failure could exist throughout the entire acci

dent progression. Hence, it is important to look at 

emergency response scenarios in which the population 

evacuates after the release has begun.

6-7
The NUREG/CR-1 150 PRA (NRC, 1990) implemented 

a partitioning method in which the thousands of source 

terms estimated in each plant analysis were placed into 

source term groups based on their potential for early 

and latent health effects. Consequence calculations 

were then performed for the frequency-weighted aver

age source term for each group and each of three sub

groups. Subdivision into the three subgroups was 

based on evacuation timing: 

Subgroup 1: evacuation starts at least 30 minutes be

fore the release begins, 

Subgroup 2: evacuation starts between 30 minutes be

fore and 1 hour after the release begins, 

Subgroup 3: evacuation starts more than 1 hour after 

the release begins. 

In NUREG/CR-I 150 (NRC, 1990), source terms for 

some subgroups were not calculated because the sub

group had zero probability. Therefore, for this study, 

the source term subgroups are selected on the basis of 

evacuation timing in the following order: Subgroup 2, 

Subgroup 3, and Subgroup 1. The source term sub

group designation is the final number in the source term 

group descriptor. (For example, the source term group 

PB-XX-l is Peach Bottom source term group XX, Sub

group 1, and the source term group SEQ-XX-2 is the 

Sequoyah source term group XX, subgroup 2.) 

Based on the selection criteria of containment failure 

leak size, containment failure time, and evacuation 

timing, relevant source term groups from the Peach 

Bottom and Sequoyah NUREG/CR- 1150 (NRC, 1990) 

analyses have been chosen to establish the discussion of 

the bellows leakage event. These source term groups 

are the most probable for the containment failure con

ditions noted. For Peach Bottom, this results in the 

selection of the following source term groups: 

I. For ECL failures, the most probable source term 

groups are PB-05 and PB-01 (subgroups PB-05-3, 

PB-01-3). 

2. For LCL failures, the most probable source term 

groups are PB-19 and PB-01 (subgroups PB-19-3, 

PB-0 1-3). 

3. For NCF, the most probable (as well as the only) 

source term groups are PB-18 and PB-17 (sub

groups PB-18-1, PB- 17-1).

For Sequoyah, these are the selected source term 

groups: 

1. For ECL failures, the most probable source term 

groups are SEQ-18 and SEQ-02 (subgroups SEQ

18-2. SEQ-02-2). 

2. For LCL failures, the most probable (as well as the 

only) source term groups are SEQ-18 and SEQ-17 

(subgroups SEQ-18-2, SEQ-17-2). 

3. For NCF, the most probable (as well as the only) 

source term groups are SEQ-17 and SEQ-16 (sub

groups SEQ-17-2, SEQ-16-2). 

The consequences calculated in NUREG/CR-I 150 

(NRC, 1990) for the selected source terms are provided 

in Table 6-4. The consequences are conditional on the 

occurrence of each source term group. Hence, to obtain 

an indication of risk, each consequence must be multi

plied by the estimated frequency of occurrence of the 

source term group (e.g., the frequency of a SA coinci

dent with a bellows leak). Also note that these results 

are site-specific with respect to such considerations as 

population density and distribution, meteorology, core 

size, etc. The results indicate that the overall predicted 

consequences are moderate, in terms of the entire SA 

spectrum of both early and latent health effect conse

quences, for the worst case surrogate scenario of ECL 

for both Peach Bottom and Sequoyah. 

In most PRAs, the source term and consequence analy

ses typically focus on accidents that will result in the 

highest risk (the product of a consequence and its prob

ability). Hence, analytical resources are usually di

rected to events that result in either high or moderate 

consequences that have some non-zero probability of 

occurrence. Conversely, the modeling of low conse

quence events (such as for source term groups PB-17 

and SEQ-16) is usually performed in a manner that is 

limited in detail, regardless of the probability of occur

rence. 

Even though there may have been limited resources 

directed toward the calculation of the source terms and 

consequences for some of the source term groups, it is 

believed that this discussion forms a valid preliminary 

study of the potential consequences from bellows leak 

failure. In addition, it is recommended that because this 

study indicates a potential for both early and latent 

health effects, that from a risk perspective, the bellows 

leak failure should be further investigated.

6-8
Table 6-4. Consequences for Selected Source Terms from NUREG/CR-1150 (NRC, 1990) 

Individual 

Source Containment Early Latent Population Population Early Late 

Term Failure Cancer Dose within Dose Entire Fatalities Cancer 

Subgroup Conditions Fatalities 50 miles (Sv) Region (Sv) Risk, 1 mi. Fatalities 

Risk, 10 mi. 

PB-01-3 ECL,LCL 9.OOE-7 3.94E+1 1.32E+3 2.50E+3 3.25E-9 7.39E-5 

PB-05-3 ECL" 1.87E-2 5.36E+2 1.86E+4 3.79E+4 5.85E-5 1.34E-4 

PB-17-1 NCF 0.OOE+O 1.33E-2 5.22E-1 9.28E-1 0.OOE+0 3.98E-9 

PB-18-1 NCF" 0.OOE+O 7.21E-1 3.22E+1 5.71E+l 0.00E+0 4.78E-7 

PB-19-3 CL'L 0.OOE+0 5.48E+l 1.74E+3 3.28E+3 0.OOE+0 8.88E-5 

SEQ-02-2 ECL 7.24E-5 6.09E+l 1.26E+3 3.78E+3 1.82E-7 8.12E-5 

SEQ-16-2 NCF O.OOE+0 6.02E-1 !.98E+l 3.21E+1 O.OOE+0 3.03E-6 

SEQ-17-2 LCL,NCF" 0.OOE+0 2.53E+0 8.09E+1 1.82E+2 0.OOE+0 7.71E-6 

SEQ- 18-2 ECL,LCL* 0.OOE+O 1.84E+2 3.06E+3 1.05E+4 0.OOE+0 1.34E-4 . Denotes most probable source term group for this containment failure condition.

6.4 Future Research 

Recommendations 

The previous discussion suggests that the following 

activities would form the basis for a comprehensive and 

defensible resolution to the degraded bellows issues. 

1. There is a need to quantify the leak rate from de

graded bellows, with measurable leak rates in ex

cess of a threshold value, under DBA conditions 

(less than approximately 10% compression or ex

pansion). To quantify the leak rates, the experi

ments should be performed with single-ply bellows 

at temperatures and pressures representative of 

DBAs. 

2. These experiments can be used to benchmark in

dustry models for predicting leak rates under DBAs 

given attempts to characterize the damage state un

der neutral conditions. This activity requires coor

dination and cooperation with industry for maxi

mum benefit. 

3. Calculations using a system level code, such as 

MELCOR, should be performed using realistic leak 

rates for DBA and SA conditions. The calculations 

should also take credit for realistic pressure transients, given the passive design features and acci

dent management procedures that are likely to be 

invoked during an accident. These elements can be 

integrated into a probabilistic framework to give a 

balanced risk informed perspective to the impact of 

degraded bellows. 

The following two tasks would also provide useful in

formation, but are lower in priority than the first three 

listed above. However, if DBA leak rate testing is per

formed, the incremental cost to do one of the two tests 

that follow would be minimal. Because both of these 

tests cause the bellows to fail, only one or the other 

could be performed using the bellows from the previous 

DBA test. 

a. Cycle the degraded bellows under DBA conditions 

(i.e., an earthquake). For example, compress the 

bellows to 10% compression, elongate to the neu

tral position, compress to 10%, elongate to neutral, 

etc., and monitor the leak rate as TGSCC cracks 

develop and grow. 

OR 

b. Quantify the leak rate from degraded bellows from 

a single SA excursion (i.e., 100% compression).

6-9
7. SUMMARY AND CONCLUSIONS

This report has presented a comprehensive examination 

of currently available information regarding bellows 

used in containment penetrations, materials, the condi

tions that cause aging, applicable aging mechanisms, 

and aging effects, and the currently available methods 

that can be used to monitor aging. 

7.1 Summary 

1. The aging of bellows is routinely assessed and 

monitored through a combination of leak rate test

ing and surveillance programs. 

2. The primary containment, as a whole, and individ

ual bellows are periodically leak tested in accor

dance with Technical Specification requirements 

and the regulatory requirements of 10 CFR 50, Ap

pendix J (1995). When leakage exceeds the allow

able rates, an investigation of the leakage cause is 

performed and the bellows is replaced if the pro

jected leak rate (for the next outage) is greater than 

Appendix J criteria allow. Usually an immediate 

retest is performed to ensure that corrective action 

was appropriate. 

3. The Type B LLRTs associated with penetrations 

that contain stainless steel expansion joints made 

with two-ply bellows typically have included inter

ply leakage tests for these bellows. It is now clear 

that this type of test is invalid and can provide non

conservative leakage rate results. 

4. Plants are not required to replace a bellows until 

the leak rate is large enough to cause the penetra

tion to fail either the LLRT or ILRT. 

5. Routine daily plant walkdowns are conducted in all 

plant areas, except within primary containment, to 

detect any abnormal physical conditions such as 

rust, dents, cracks, deformation, discoloration, in

sulation damage, vibration, leaks, spills, or any un

usual noises. The accessible portions of mechani

cal penetrations outside of primary containment are 

observed during these walkdowns. 

6. Plant walkdowns of all areas are performed after 

refueling and other planned plant shutdowns and 

just before plant restart (in this case, personnel pay 

particular attention to areas or components that 

were modified, refurbished, or replaced during the 

outage). All reasonably accessible portions of the 

penetrations are observed during these walkdowns.

All abnormal conditions identified during these 

walkdowns are documented and investigated and 

corrective action is taken as necessary. Although 

many kinds of damage can be detected during 

walkdowns, it is unlikely that TGSCC will be ob

served during a walkdown. 

7. Degradation of bellows and the inability to effec

tively repair them has caused the entire expansion 

joint to be replaced for piping penetrations at sev

eral nuclear plants. These replacements are diffi

cult and costly because of poor access around the 

large diameter penetrations. 

8. A structured penetration inspection program based 

on plant-specific operating experience and coupled 

with root cause analysis and appropriate corrective 

action could be considered an aging management 

program. Note that many existing plant activities 

not specifically focused on penetration aging man

agement may often be credited with addressing 

aging effects. 

9. Effective July 1996, plants have implemented the 

Maintenance Rule (10 CFR 50.65, 1996). The in

vestigation of MPFFs and the actual monitoring of 

penetration performance will support aging man

agement for bellows. 

7.2 Conclusions 

The following conclusions regarding bellows historical 

performance can be supported: 

1. Bellows have primarily failed because of TGSCC. 

2. Degradation of bellows has been identified primar

ily through integrated and LLRTs. 

3. TGSCC of stainless steel bellows is not easily de

tectable using visual inspection techniques. After 

leak tests have indicated leaks in a penetration, PT 

tests have been used to examine the bellows. PT 

can only identify damage that is on the outside of 

the accessible surface. It is not clear how signifi

cant the damage must be before it can be observed 

using PT. 

4. Given an aggressive environment (i.e., a humid 

environment and nearby insulation with chlorides, 

such that chlorides leach out of the insulation and 

onto the bellows), bellows failure can occur over a

7-1
period of a few years. Bellows failures have oc

curred in plants that are fairly new, as well as in 

older plants. New bellows that have replaced older 

bellows have also developed TGSCC, and had to 

be replaced themselves. On the other hand, many 

older bellows have not experienced TGSCC. 

Therefore, it appears that the environment (i.e., 

presence of chlorides or other chemical agents) is a 

major factor, and bellows age, by itself, is not a 

good indication of the likelihood of TGSCC. 

5. No aging mechanisms other than those identified in 

this report are known or expected to occur during 

this period (i.e., no mechanisms unique to the li

cense renewal period have been identified). 

6. Aging and degradation mechanisms of stainless 

steel expansion joint bellows are not fully con

trolled through current maintenance, surveillance, 

and condition monitoring techniques employed by 

operating plants. This is from the lack of any reli

able means for detection of TGSCC. 

7. Plants are not required to replace a bellows until 

the leak rate is large enough to cause the penetra

tion to fail either the LLRT or ILRT. However, 

testing (Lambert and Parks, 1995) has shown that 

when subjected to large displacements, bellows 

with TGSCC damage develop leaks, or develop 

larger leaks, than would occur for a bellows with

out TGSCC. 

8. Once a bellows ply develops cracks large enough to 

cause it to fail a leak test, there is no way to repair 

it and still maintain the design characteristics re

quired. The bellows must be replaced, although 

this is relatively difficult and costly. 

9. Bellows components require little in the way of 

periodic maintenance. 

7.3 Additional Research Needed 

1. The capability to determine the aging condition of 

bellows using a nondestructive test would be use

ful. Currently, the detection of SCC is extremely

difficult, and evaluating the baseline condition is 

limited to "leak" or "no leak" determinations. De

tecting damage, and quantifying the extent of the 

damage, is important for the long-term reliability of 

bellows. 

2. It is also important to understand how TGSCC 

damage affects leakage through the bellows during 

both design and SA scenarios. For example, a 

bellows degraded by TGSCC may be able to pass a 

leak test (i.e., leakage is less than allowable limit), 

if relative movement is minimal, but cracks and 

leak rate could increase significantly during an ac

cident that required significant movements be ac

commodated by the bellows. Testing to quantify 

the leak rate under DBA conditions would either 

verify that current industry practice (i.e., allowing 

leaking bellows to remain in service if the leak is 

small enough) is acceptable, or provide information 

about how large the leak rate could become during 

the DBA. (See Section 6.3 for additional recom

mended leak rate testing.) 

3. Calculations using a system level code, such as 

MELCOR, could be performed using realistic leak 

rates for DBA and SA conditions. The calculations 

could take credit for realistic pressure transients, 

given the passive design features and accident 

management procedures that are likely to be in

voked during an accident. These elements can be 

integrated into a probabilistic framework to give a 

balanced risk informed perspective to the impact of 

degraded bellows. 

4. Are replacement bellows as tough as the original 

bellows were? Some bellows have been replaced 

that consist of two halves, with longitudinal welds 

that join the two parts. Can these "new" bellows 

withstand, without leaking, the same amount of 

relative movement that the original bellows could 

when it was new? The principal concern is the dif

ficulty of making a longitudinal weld along the 

convolutions of the bellows, coupled with very 

large strains that occur in the bellows convolutions. 

Limited testing could evaluate whether the "new" 

bellows are as robust as the "old" bellows.

7-2
8. REFERENCES

10 CFR (Code of Federal Regulations) 50.49. 1983. 

Environmental Qualification of Electrical Equipment 

Important to Safety for Nuclear Power Plants. Wash

ington, DC: US Nuclear Regulatory Commission. 

10 CFR 50.55a(b)(2)(ix-x). 1997. Codes and Stan

dards. Washington, DC: US Nuclear Regulatory 

Commission. 

10 CFR 50, Appendix J. 1995. Primary Reactor 

Containment Leakage Testing for Water-Cooled 

Power Reactors. Washington, DC: US Nuclear 

Regulatory Commission. 

10 CFR 50.73. 1995. Licensee Event Report System. 

Washington, DC: US Nuclear Regulatory Commis

sion. (Cited in this NUREG on p. 14.) 

10 CFR 54. 1996. Requirements for Renewal of Op

erating Licenses for Nuclear Power Plants. Washing

ton, DC: US Nuclear Regulatory Commission. 

10 CFR 54.4. 1996. Scope. Washington, DC: US 

Nuclear Regulatory Commission. (Cited in this 

NUREG on p. 23.) 

10 CFR 54.21. 1995. Contents of Application-Tech

nical Information. Washington, DC: US Nuclear 

Regulatory Commission. 

10 CFR 100. 1962. Reactor Site Criteria. Washing

ton, DC: US Nuclear Regulatory Commission. 

ASME (American Society of Mechanical Engineers). 

1992a. Boiler and Pressure Vessel Code. Subsection 

IWE, "Requirements for Class MC and Metallic Lin

ers of Class CC Components of Light-Water Cooled 

Plants." New York, NY: ASME. 

ASME. 1992b. Boiler and Pressure Vessel Code. 

Section III, Subsection NE, "Class MC Components." 

New York, NY: ASME. 

Berg, P. and Brown, J.A. 1995 "Testability of Con

tainment Flexible Metallic Expansion Joints," Nu

clear Plant Journal, September-October 1995. 

pp. 46-50. 

Brown, J.A. and Tice, G.A. 1993. "Containment 

Penetrations-Flexible Metallic Bellows Testing, 

Safety, Life Extension Issues," Nuclear Engineering 

and Design, 145.

Bruneau, M. and Zahrai, S.M. 1997. "Effect of Se

vere Corrosion on Cyclic Ductility of Steel," Journal 

of Structural Engineering, Vol. 123, no. 11. 1478

1486. 

EJMA (Expansion Joint Manufacturers Association). 

1985. Standards of the Expansion Joint Manufactur

ers Association, Inc. 5th edition. Addenda. Tarry

town, NY: Expansion Joint Manufacturers Associa

tion. 

EPRI (Electric Power Research Institute). 1986. 

"Natural versus Artificial Aging of Nuclear Power 

Plant Components," NP-4997, prepared by University 

of Connecticut, December 1986. 

EPRI. 1994a. "PWR Containment Structures License 

Renewal Industry Report, Rev. 1," TR-103835, July 

1994. 

EPRI. 1994b. "BWR Containment License Renewal 

Industry Report, Rev. 1," TR-103840, July 1994. 

Gregory, J.J. et al. 1990. Evaluation of Severe Acci

dent Risks: Sequoyah, Unit 1. NUREG/CR-4551/5, 

SAND86-1309/5, Rev. 1, Part 1. Albuquerque, NM: 

Sandia National Laboratories. 

Greimann, L. et al. 1991. Analysis of Bellows Expan

sion Joints in the Sequoyah Containment. 

NUREG/CR-5561, SAND/90-7020. Albuquerque, 

NM: Sandia National Laboratories. 

Kulak, R.F. et al. 1985. Structural Response of Large 

Penetrations for Containment Vessels Subjected to 

Loadings Beyond Design Basis. NUREG/CR-4064, 

SAND84-7177, ANL-84-41. Albuquerque, NM: San

dia National Laboratories. 

Lambert, L.D. and Parks, M.B. 1994. Experimental 

Results from Containment Piping Bellows Subjected 

to Severe Accident Conditions: Results from Bellows 

Tested in 'Like-New' Conditions. NUREG/CR

6154/1, SAND94-171 1/1. Albuquerque, NM: Sandia 

National Laboratories. 

Lambert, L.D. and Parks, M.B. 1995. Experimental 

Results from Containment Piping Bellows Subjected 

to Severe Accident Conditions: Results from Bellows 

Tested in Corroded Conditions. NUREG/CR-6154/2, 

SAND94-1711/2. Albuquerque, NM: Sandia National 

Laboratories.

8-1
NASA (National Aeronautics and Space Administra

tion). Space Materials Handbook. 1969. 3rd edition. 

NASA SP-305 1. Washington, DC: NASA. 

NRC (Nuclear Regulatory Commission). 1990. Se

vere Accident Risks: An Assessment for Five U.S. 

Nuclear Power Plants. NUREG/CR- 1150. Washing

ton, DC: NRC. 

NRC. 1992. "Inadequate Local Leak Rate Testing," 

Information Notice 92-20, March 3, 1992. Washing

ton, DC: NRC. 

NRC. 2000a. Final Safety Analysis Report, Grand 

Gulf Nuclear Power Station, MS. NRC Docket File 

050-00416. Washington. DC: NRC. 

NRC. 2000b. Final Safety Analysis Report, Mon

ticello Nuclear Power Station, MN. NRC Docket File 

050-00423. Washington, DC: NRC. 

NRC. 2000c. Final Safety Analysis Report, Nine Mile 

Point Nuclear Power Station, Unit 2, NY. NRC 

Docket File 050-004 10. Washington, DC: NRC. 

NRC. 2000d. Final Safety Analysis Report, Prairie 

Island Nuclear Power Station, MN. NRC Docket File 

050-00306. Washington, DC: NRC. 

NRC. 2000e. Final Safety Analysis Report, Surry 

Nuclear Power Station, VA. NRC Docket File 050

00281. Washington, DC: NRC.

NRC. 2000f. Final Safety Analysis Report, Watts Bar 

Nuclear Power Station, TN. NRC Docket File 050

00390. Washington, DC: NRC. 

NEI (Nuclear Energy Institute). 1995. "Industry 

Guideline for Implementing the License Renewal 

Rule, 10 CFR Part 54," NEI 95-10. Washington, DC: 

Nuclear Energy Institute, February 1995. 

Payne, A.C. et al. 1990. Evaluation of Severe Acci

dent Risks: Peach Bottom, Unit 2. NUREG/CR-4551, 

SAND86-1309, Vol. 4, Rev. 1, Part 1. Albuquerque, 

NM: Sandia National Laboratories. 

Sams, D.W. and Trojovsky, M. 1980. Data Summa

ries of Licensee Event Reports of Primary Contain

ment Penetrations at US Commercial Nuclear Power 

Plants. NUREG/CR-1730, EGG-EA-5188. Idaho 

Falls, ID: EG&G Idaho, Inc. 

Shackelford, M.H., Bump, T.R., and Seidensticker, 

R.W. 1985. Characterization of Nuclear Reactor 

Containment Penetrations. NUREG/CR-3855, ANL

84-87, SAND84-7139. Albuquerque, NM: Sandia 

National Laboratories. 

Shigley, J.E. 1983. Mechanical Engineering Design. 

2nd edition. New York, NY: McGraw-Hill.

8-2
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

(2-89) (Assigned by NRC, Add Vol., Supp., Rev., NRCM 1102, and Addendum Numbers, If any.) 3201.3202 BIBLIOGRAPHIC DATA SHEET 

(See n sctions the revss) NUREG/CR-6726 

2. TITLE AND SUBTITLE SAND 2001-1147P 

Aging Management and Performance of Stainless Steel Bellows in 

Nuclear Power Plants 3. DATE REPORT PUBLISHED 

MONTH YEAR 

May 2001 

4. FIN OR GRANT NUMBER 

W6690 

5. AUTHOR(S) 6. TYPE OF REPORT 

J.L. Cherry, J.M. Clauss, M. Pilch, J.J. Gregory, J.A. Smith Technical 

7. PERIOD COVERED (Inclusive Dates) 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (if NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and maihng address; if contractor, 

provide name and mailing address.) 

Sandia National Laboratory 

Albuquerque, NM 87185-0744 

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type "Same as above'; if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission, 

and ma&n address.) 

Division of Engineering Technology 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

10. SUPPLEMENTARY NOTES 

Herman L. Graves Ill, NRC Project Manager 

11. ABSTRACT (200 words or less) 

In commercial nuclear power plants, many containments use stainless steel bellows at piping penetrations, in dry-well-to -wetwell vent lines, or as a part of the fuel transfer tube. Bellows are an integral part of the primary containment pressure boundary in nuclear power plants. These bellows allow for thermal expansion and contraction of the pipe, vent line or fuel 

transfer tube, and accommodate relative motion between the containment and other structures, such as a shield wall. 

Ensuring that bellows in operational plants remain leak tight is an important safety issue. A number of bellows have been 

replaced when transgranular stress corrosion cracking caused them to leak. Determining how to evaluate bellows for 

age-related damage is a critical issue that must be addressed for plants operating under their original license, as well as for 

plants applying for license extension. 

The types of bellows, locations in a containment, and required bellows leak tests are described for major categories of 

boiling-water reactor (BWR) and pressurized-water reactor (PWR) power plants. Compilation of instances of bellows cracking, excessive leakage, or structural integrity questions, as well as testing issues and other degradation problems, are reviewed 

and summarized. The aging mechanisms that caused the degradation are discussed. Inspection methods currently in use, and a typical industry approach to managing bellows aging are described. Degradation that resulted from corrosion, misalignment, and other damage was considered. The risk and regulatory significance of containment bellows when subjected 

to Current License Basis (CLB), degraded, and severe accident (SA) conditions are also addressed.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.) 13. AVAILABILITY STATEMENT 

Bellows, Stainless Steel, Corrosion, Degradation, Transgranular Stress, and Corrosion Cracking unlimited 

14. SECURITY CLASSIFICATION 

(This Page) 

unclassified 

(This Report) 

unclassified 

15. NUMBER OF PAGES 

16. PRICE

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NUREG/CR-6726 AGING MANAGEMENT AND PERFORMANCE OF STAINLESS STEEL BELLOWS IN 

NUCLEAR POWER PLANTS

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, DC 20555-0001 

OFFICIAL BUSINESS 

PENALTY FOR PRIVATE USE, $300

MAY 2001