NUREG-1719

Pipe Cracking in 

U.S. BWRs: A 

Regulatory History

U.S. Nuclear Regulatory Commission 

Office of Nuclear Reactor Regulation 

Washington, DC 20555-0001
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I I

I
NUREG-1719 

Pipe Cracking in 

U.S. BWRs: A 

Regulatory History 

Date Completed: November 1999 

Date Published: May 2000 

Prepared By 

J. R. Strosnider, Jr., W. H. Koo 

I. A. Davis, M. C. Modes 

Division of Engineering 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001
ABSTRACT

This paper was prepared, in part, to 

support discussions at an IAEA Regional 

Workshop on Environmentally Assisted 

Cracking of Nuclear Power Plant 

Austenitic Piping that was held in 

Slavutych, Ukraine from 22 to 26 June 

1998 '. The paper presents a regulatory 

history and offers a perspective on 

intergranular stress corrosion cracking 

(IGSCC) in U.S. boiling-water reactor 

piping.

The paper focuses on regulatory and 

industry actions taken to assure that U.S. 

licensees manage IGSCC in a manner 

that provides safe and reliable plant 

operation. Although the paper does not 

offer extensive theoretical details on the 

IGSCC phenomenon, it does discuss 

some of the key technical issues that 

influenced regulatory positions and 

industry actions.

1. "Report of a Regional Workshop on Environmentally Assisted Cracking of NPP Austenitic Piping," 

IAEA TC Project RER/9/052, RBMK-SC-060, Vienna, November 4, 1998.

NUREG-1719 iio
CONTENTS 

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

List of Tables ... .................................................... ix 

List of Appendices .. ................................................. xi 

Executive Summ ary ................................................. xiii 

Acknowledgment .................................................... xv 

1 Introduction ..................................................... 1-1 

2 Typical Design of U.S. Boiling Water Reactors .......................... 2-1 

3 Cause of Intergranular Stress Corrosion Cracking ........................ 3-1 

4 IGSCC Operating Experience ....................................... 4-1 

5 Chronology of Regulatory Actions .................................... 5-1 

5.0 O verview .................................................... 5-1 

5.1 First Pipe Cracking Study Group - October 1975 ...................... 5-1 

5.2 NUREG-0313 - July 1977 ....................................... 5-1 

5.3 Second Pipe Cracking Study Group and NUREG-0531 - February 1979 .... 5-1 

5.4 NUREG 0313, Revision 1 - July 1979 .............................. 5-2 

5.5 Third Pipe Cracking Study Group and NUREG-0691 - September 1980 .... 5-2 

5.6 Information Notice 82-39 - September 1982 ......................... 5-2 

5.7 Bulletin 82-03 - October 1982 .................................... 5-2

V NUREG-1719
5.8 Bulletin 83-02 - March 1983 ........................ 

5.9 NRC Orders - August 1983 ......................... 

5.10 Generic Letter 84-11 - April 1984 .................... 

5.11 NDE Coordination Plan on Inspection Qualification - July 1• 

5.12 Fourth Pipe Crack Study Group and NUREG-1061, 

Volume 1 - August 1984 ......................... 

5.13 NUREG-0313, Revision 2 - June 1986 

and Generic Letter 88-01 - January 1988 ............. 

5.14 Generic Letter 88-01, Supplement 1 - February 1992..... 

5.15 Appendix VIII and the Performance Demonstration Initiative 

M itigation M ethods .................................. 

6.0 Overview .................................... 

6.1 W ater Chemistry Control ........................... 

6.2 Stress Improvement ............................... 

Inspection ......................................... 

7.0 Overview ....................................... 

7.1 Qualification of Inspection Personnel and Methods ....... 

7.2 Scope and Frequency of Inspections ................... 

Flaw Evaluation Methods .............................

)84.

- 1996 ...

9 epair M ethods ...................... 

10 Piping Replacement ................... 

NUREG-1719 vi

Page 

* 5-3 

5-3 

5-3 

* 5-3

5-4

6 

7 

8

. 5-4 

5-4 

5-5 

6-1 

6-1 

6-1 

6-3 

7-1 

7-1 

7-1 

7-3 

8-1 

9-1 

10-1

I I
11 Leak Detection and Leak-Before-Break ............................... 11-1 

12 Current Status of BW R Piping ...................................... 12-1 

13 Conclusions .................................................... 13-1 

14 References ..................................................... 14-1 

vii NUREG-1719
LIST OF TABLES

Table 4.1 

Table 4.2 

Table 12.1

Page 

IGSCC Incidents by Line Type in U.S. and Foreign BWRs .......... 4-1 

Summary of Inspection Findings on Large Piping 

in BWRs Inspected According to IEBs 82-03 and 83-02 ............ 4-4 

Status of U.S. BW R Piping .................................. 12-2

ix NUREG-1719
LIST OF APPENDICES 

Appendix A Abbreviations .......................................... A-1

xi NUREG-1719
EXECUTIVE SUMMARY

This paper presents a regulatory 

perspective on the history of intergranular 

stress corrosion cracking (IGSCC) in the 

piping of U.S. boiling-water reactor 

(BWR) plants. The paper focuses on 

regulatory and industry actions taken to 

assure that U.S. BWR licensees manage 

IGSCC in a manner that provides safe 

and reliable plant operation. 

Development of the U.S. regulatory 

framework for managing IGSCC in 

BWR piping was an evolutionary 

process driven by operating experience, 

research, and technological 

developments. IGSCC was initially 

observed in only small-diameter piping;

however, in time it affected both large

and small-diameter piping systems. A 

key element in the management of 

IGSCC was an aggressive inspection 

program utilizing qualified inspection 

techniques and personnel. In parallel, the 

development of qualified repair and 

mitigation methods, along with 

replacement of piping with IGSCC

resistant materials resulted in an effective 

and economical solution to the problem. 

Today, a combination of regulatory 

documents, code rules, and industry 

guidelines is available to support 

effective programs for managing IGSCC 

in the piping of U.S. BWRs.

Xlll NUREG-1719
ACKNOWLEDGMENT

The authors appreciate the assistance of 

Mr. James Medoff, U.S. Nuclear 

Regulatory Commission, Materials and

Chemical Engineering Branch, for his 

assistance in preparing this report.

xv NUREG-1719
1 INTRODUCTION

This paper presents a regulatory 

perspective on the history of 

intergranular stress corrosion cracking 

(IGSCC) in the piping of U.S. boiling

water reactor (BWR) plants. The paper 

is not intended to provide extensive 

theoretical details on the IGSCC 

phenomenon or other related subjects 

such as nondestructive testing and 

fracture mechanics. An extensive 

literature offers theoretical and 

engineering details on these subjects, 

and some selected references on these 

subjects are listed at the end of this 

paper. This paper focuses, rather, on 

regulatory and industry actions taken to 

assure that U.S. BWR licensees manage 

IGSCC in a manner that provides safe 

and reliable plant operation. Some key 

technical issues that influenced 

regulatory positions and industry actions 

are discussed. 

Chapter 2 of this report briefly describes 

the typical design of BWRs operated in 

the U.S. Chapter 3 presents a brief 

discussion of the causal factors of 

IGSCC and BWR piping design features 

of particular interest with regard to 

IGSCC. The information in these 

sections serves as background to assist in 

understanding issues discussed in later 

sections of the paper. Chapter 4 briefly 

summarizes BWR pipe cracking 

experience in the U.S. In retrospect, 

given the material selections, fabrication 

methods, and operating environments of 

BWRs, the occurrence of IGSCC in

BWR piping should not have been 

totally unexpected. However, the 

discovery of IGSCC in small-diameter 

piping in the late 1970s and in large

diameter BWR piping in the early 1980s 

required substantial efforts from the U.S. 

industry and the U.S. Nuclear Regulatory 

Commission to support development and 

implementation of the engineering and 

regulatory solutions to the problem. 

Chapter 5 summarizes the major 

regulatory initiatives taken in response to 

the BWR pipe cracking problem. 

Specific issues that had to be addressed 

were methods for mitigation, inspection, 

flaw evaluation, piping repair, and 

piping replacement. These issues are 

discussed in Chapters 6 through 10 of 

this paper, respectively. 

Chapter 11 of this report discusses leak 

detection and leak-before-break (LBB) 

in the context of BWR IGSCC. It is 

important to note that LBB is not 

accepted in the U.S. as a regulatory basis 

for assuring the integrity of piping 

systems subject to IGSCC. Nonetheless, 

it is an important element in "defense-in

depth" and should be given close 

attention in any program for maintaining 

piping and component integrity. 

Chapter 12 of this report provides a 

current status of BWR piping in U.S. 

plants and Chapter 13 present major 

conclusions of the report.

1-1 NUREG-1719
2 TYPICAL DESIGN OF U.S. BOILING WATER REACTORS

The direct-cycle boiling-water reactor 

(BWR) nuclear system consists of a 

nuclear core located inside a reactor 

vessel and a conventional turbine

generator and feedwater supply system. 

Associated with the nuclear core are 

auxiliary systems to accommodate the 

operational and safeguards requirements. 

Water circulating through the reactor 

core produces saturated steam, which is 

separated from recirculation water, dried 

in the top of the vessel, and directed to 

the steam turbine-generator. The turbine 

employs a conventional regenerative 

cycle with condenser deaeration and 

condensate demineralization. 

The nuclear core, the source of the heat, 

consists of fuel assemblies and control 

rods contained within the reactor vessel 

and cooled by the recirculating water 

system. The power level is maintained 

or adjusted by positioning control rods 

up and down within the core and by 

changing the recirculation flow rate 

through the core without changing the 

position of the control rods. The 

recirculation system utilizes jet pumps 

inside the reactor vessel. These pumps 

generate about two-thirds of the 

recirculation flow within the reactor 

vessel. The external recirculation system 

piping is made up of two separate 

parallel pump loops, each loop 

consisting of a reactor recirculation 

pump and associated piping. The pump 

discharge line in each loop feeds a pipe 

riser manifold that has five riser outlets.

Each outlet supplies a pair of jet pumps. 

Connections are provided on the suction 

side of one loop to supply reactor 

coolant during the shutdown mode of the 

residual heat removal (RHR) system and 

to the reactor water cleanup system. The 

discharge piping of both loops has 

connections from the RHR system 

discharge to support both the shutdown 

cooling and low-pressure coolant 

injection modes of operation. 

The BWR operates at a constant steam 

pressure of about 8.62 MPa (1250 psi) 

and a temperature of about 287.8 *C 

(550'F). U.S. BWRs have fast-acting 

emergency core cooling systems 

(ECCSs). 

The nuclear boiler system is supported 

by the following specialized functions of 

its auxiliary systems, which are used for 

normal plant operation: 

"* reactor water cleanup (RWCU) system 

"* shutdown cooling function of residual 

heat removal (RHR) system 

"* fuel building and containment pool 

cooling and filtering system 

"* closed cooling water system for reactor 

service 

"* radioactive waste treatment system 

The following auxiliary systems are used 

as backup (standby) or emergency 

systems: 

- standby liquid control (SLC) system

2-1 NUREG-1719
"* reactor core isolation cooling (RCIC) 

system 

"* high-pressure core spray (HPCS) 

system 

"* low-pressure core spray (LPCS) system 

"* automatic depressurization function 

"* residual heat removal (RHR) system 

- low-pressure coolant injection 

(LPCI) function 

- steam condensing function 

- containment spray function 

- suppression pool cooling function 

The emergency systems provide diverse 

and redundant means for achieving and 

maintaining safe shutdown of the reactor 

assuming design-basis pipe failures up to 

a full double-ended, offset break of the 

largest diameter pipe in the system. 

NUREG-1719

BWRs in the U.S. are provided with 

multi-barrier pressure suppression 

containments. The primary containment 

consists of a drywell, which encloses the 

reactor vessel and recirculation system, 

and a pressure suppression chamber. The 

primary containment is designed to 

prevent the release of radioactive fission 

products to the environment in the event 

of a design-basis pipe break. A 

secondary containment consists of a 

reactor building that completely encloses 

the pressure suppression primary 

containment. This structure provides 

secondary containment when the primary 

containment is in service, and provides 

primary containment during periods 

when the primary containment is open, 

as during refueling.

2-2

I 1 1
3 CAUSE OF INTERGRANULAR STRESS 

CORROSION CRACKING

Intergranular stress corrosion cracking 

(IGSCC) occurs as a result of the effects 

of stress and environment on a 

susceptible material. These three factors 

- material susceptibility, environment, 

and stress - are discussed below. 

The original piping material in U.S. 

BWRs was type 304 or 316 austenitic 

stainless steel. The corrosion resistance 

of austenitic stainless steel results from 

the addition of more than 12% 

chromium which combines with oxygen 

to form a passive chromium-oxide film 

on the surface of the material. However, 

if austenitic stainless steels that contain 

more than 0.03 % carbon are heated to 

or slowly cooled through the temperature 

range of 520 to 820 °C (900 to 1450 

OF), chromium carbides will form at the 

grain boundaries. The regions adjacent 

to the precipitated chromium carbides 

will be depleted of chromium to less 

than 12%, resulting in a loss of corrosion 

resistance in those areas. Stainless steel 

in this condition is referred to as 

"sensitized." The most common cause 

of sensitization in BWRs is welding 

followed by a slow cooling through the 

sensitizing temperature range. The 

sensitization in these cases is contained 

within the weld heat-affected zone. 

Sensitization can be avoided by 

controlling the carbon content to below 

0.03%, by cooling quickly through the 

sensitizing temperature range, or by

solution annealing sensitized material 

and rapidly cooling. In the solution 

annealing process, the material is heated 

to 1100 to 1150 °C (1950 to 2050 OF) 

until the chromium carbides dissolve. 

The material is subsequently quenched to 

prevent the carbides from reforming. 

Another approach to reduce sensitization 

is to add strong carbide formers such as 

titanium or niobium to the steel. The 

carbon in the steel will preferentially 

react with the titanium or niobium to 

form titanium or niobium carbides 

instead of reacting with the chromium. 

Stainless steels with additions of 

titanium or niobium are called 

"stabilized." IGSCC can still occur in 

stabilized stainless steels if the carbon 

content is too high. Stabilized stainless 

steels generally have not been used in 

U.S. BWRs. 

Environment plays a key role in the 

occurrence of IGSCC. The IGSCC 

mechanism is an electrochemical 

process; therefore the propensity for 

IGSCC is very dependent on the 

electrolytic nature of the environment in 

which the material is placed. Impurities 

in the BWR reactor coolant provide the 

electrolytic environment necessary to 

support IGSCC. These impurities 

include oxygen, which may be present 

as a result of radiolysis or the ingress of 

air into the reactor coolant, or anionic 

impurities such as sulfates or chlorides,

3-1 NUREG-1719
which may be present as a result of 

leakage from the condenser into the 

reactor coolant. Higher levels of oxygen 

and impurities increase the potential for 

IGSCC. Controlling the oxygen 

concentration below 20 ppb by the 

addition of hydrogen will normally 

prevent IGSCC from initiating. 

Although it is important to control the 

bulk coolant chemistry, once a crack has 

initiated, the microchemistry within the 

crack is very difficult to change or 

control; thus, IGSCC may continue even 

after extensive efforts to improve the 

bulk chemistry have been initiated. 

The third component necessary for 

IGSCC to occur is stress. Stress is 

generally considered to contribute to the 

IGSCC process through rupturing of the 

protective oxide film. The major stress 

contribution in BWR piping comes from 

NUREG-1719

residual welding stresses, which can be 

on the order of the yield stress of the 

material. Although lower in magnitude, 

other stresses - such as cyclic stresses 

from plant startup and shutdown or 

vibration - may play an important role. 

The three factors discussed above, 

material susceptibility, environment, and 

stress, interact in a synergistic manner to 

produce IGSCC. The degree of 

sensitization, oxidizing potential of the 

environment and level of stress can be 

varied in different combinations to 

produce a range of initiation times and 

crack growth rates for IGSCC. This 

makes it very difficult to extrapolate 

laboratory data to field experience or to 

develop reliable models for predicting 

inservice performance, and care should 

be taken in utilizing such approaches to 

managing IGSCC.

3-2

I II
4 IGSCC OPERATING EXPERIENCE

In this section, the staff summarizes 

IGSCC experience in U.S. plants. 

Reference is made to several pipe crack 

study groups, NUREG reports, and NRC 

bulletins that resulted in augmented 

inspections and other activities to 

address IGSCC. These regulatory 

activities are discussed in greater detail 

in Chapter 5 of this report. 

Consideration, in concert, of the contents 

of this Chapter and Chapter 5 provides 

greater insight to the interaction between 

operating experience and regulatory 

actions that ultimately lead to a 

comprehensive approach to managing 

IGSCC. 

In NUREG-0531, "Investigation and 

Evaluation of Stress-Corrosion Cracking 

in Piping of Light Water Reactor Plants" 

the NRC staff and its contractors

summarized information regarding 

cracking in BWR piping from 1965 

through 1978. IGSCC was identified in 

BWR piping as early as 1965. In 

December 1965, during a hydrostatic 

pressure test, a leak was observed in a 6

inch bypass line of the recirculation loop 

at Dresden Unit 1. From December 

1965 until September 1974, cracks were 

found in the piping of six U.S. BWRs. 

All cracking was found in the heat

affected zone of welds in type 304 

stainless steel piping of 8-inch or smaller 

diameter. From 1975 through 1978, 

IGSCC was found in recirculation 

bypass, core spray, reactor water cleanup 

and control rod drive return piping. 

NUREG-0531 contains a summary of 

pipe cracking in both domestic and 

foreign BWRs; this information is also 

summarized in Table 4.1 below.

and Foreign BWRS(a)

Number IGSCC Incidents

System Component 

(pipe diameter)

Before 

July 1975

-..-- I -

July 1975 to 

January 1979

Totals

Recirculation Bypass Line (4-inch) 30 12 42 

Core Spray Pipe (10-inch) 16 17 33 

Control Rod Drive System Small Bore Pipe (CRD, 3-inch) 1 1 2 

Reactor Water Cleanup (RWCU; 3- to 8-inch) 10 14 24 

Large Recirculation (>_ 12-inch ) 0 13 13 

Small Bore Pipe (_• 3-inch) other than CRD and RWCU 0 6 6 

Other Piping Systems 7 6 (b) 13

Notes: (a) Cracking incidents reported to the NRC 

(b) Cracking incidents in a large diameter pipe of a German BWR

4-1 NUREG-1719
Many of the cracks discovered after 

1975 were found as the result of 

augmented inservice inspections 

performed consistent with the 

recommendations of the first Pipe Crack 

Study Group and the NRC positions in 

NUREG-0313, 'Technical Report on 

Material Selection and Processing 

Guidelines for BWR Coolant Pressure 

Boundary Piping," as discussed in 

Chapter 5 of this report. 

In NUREG-0531, the staff also noted 

that IGSCC had occurred in BWR 

Inconel safe-ends. Safe-ends are 

attached to the vessel to serve as 

transition pieces to facilitate welding of 

the austenitic stainless steel piping to the 

ferritic steel pressure vessel nozzles. 

The safe-end material can be sensitized 

when the ferritic steel components to 

which they are attached are given a post

weld heat treatment. In 1975, only three 

BWRs with sensitized safe-end material 

remained in operation in the U.S., and 

these plants were performing periodic 

inspections of the safe-ends. 

One of the most significant IGSCC 

incidents reported in the U.S. occurred at 

Duane Arnold on June 17, 1978. During 

this event, a leak of approximately 

684 L/hr (3 gpm) was identified from 

one of the eight 10- inch diameter 

recirculation inlet nozzle safe-ends. 

Each of these safe-ends is fabricated 

with an internal thermal sleeve that 

creates a deep crevice in the design. The 

safe-end had been leaking for about 3 

days before the source of the leak was 

NUREG-1719

identified following an unrelated reactor 

scram. Following identification of the 

leaking safe-end, the licensee performed 

ultrasonic testing (UT) of the remaining 

safe-ends and subsequently replaced all 

of the safe-ends in the system. 

Destructive examinations of the affected 

safe-ends revealed that all eight safe

ends had inside surface cracks which 

extended essentially completely around 

the circumference of the design; these 

cracks were located in the creviced 

region of safe-end design. The depth of 

the cracks typically ranged from 50 

percent to 75 percent of the wall 

thickness, except for the leaking safe

end where an through-wall crack was 

present in an 80-degree segment of the 

circumference. The licensee attributed 

the cracking to IGSCC and concluded 

that the cracking was a result of the 

combined effects of an oxygenated 

reactor coolant environment, high 

residual stresses from welding, use of a 

sensitized material during the fabrication 

of the components, and the presence of 

the deep crevice in the design. However, 

the licensee concluded that the 

conditions favorable to supporting 

IGSCC existed even without the 

contributions from However, the 

licensee concluded that the conditions 

favorable to supporting IGSCC existed 

even without the contributions of these 

two factors. 

The cracking at Duane Arnold was 

significant in that it demonstrated the 

potential for IGSCC to develop deep,

4-2

I I
360-degree flaws. This type of flaw 

challenges the concept of LBB (refer to 

Chapter 11) and emphasized the need 

for comprehensive inspections. 

In NUREG- 1061, Volume 1, "Report of 

the U.S. Nuclear Regulatory 

Commission Piping Review 

Committee," the fourth Pipe Crack Study 

Group, as discussed in Chapter 5, 

updated both domestic and foreign BWR 

operating experience through March 

1984. Augmented inspections 

performed in accordance with NUREG

0313, Revision 1, "Technical Report on 

Material Selection and Processing 

Guidelines for BWR Coolant Pressure 

Boundary Piping," published in July 

1980, had revealed a significant amount 

of cracking in small-diameter piping in 

systems such as RWCU and core spray. 

The results of inspections of large-bore 

piping conducted in response to NRC 

Generic Letters 82-03 and 83-02 were of 

particular interest. NUREG-1061, 

Volume 1, provides a tabular summary 

of these inspection results. These 

inspection results are also summarized in 

Table 4.2. 

Table 4.2 indicates that a wide variety in 

the extent of cracking occurred from 

plant to plant without any clear 

explanation of the discrepancy in the 

inspection results. This experience 

illustrates the observation in Chapter 2 

that it is very difficult to develop reliable 

models for predicting inservice 

performance of materials. These results 

also demonstrated the need for a

comprehensive, integrated approach to 

IGSCC inspection. 

In NUREG 1061, Volume 1, the staff 

reported that as a result of this inspection 

experience, U.S. utilities were 

developing programs to preclude or 

minimize IGSCC. Several utilities were 

in the process of replacing (or planning 

to replace) susceptible material. Weld 

overlay repairs (refer to Chapter 9.0) 

and induction heating stress 

improvement (refer to Chapter 6, Section 

6.2) had been implemented at several 

sites. 

Since the issuance of Generic Letter 

88-01, the industry has been performing 

IGSCC inspections on a regular basis. 

The inspection schedule for each weld 

was based on the weld's degree of 

susceptibility to IGSCC. In the earlier 

rounds of inspection, many large pipe 

welds were found to be cracked. 

Recently, there have been very few 

reported events pertaining to IGSCC in 

austenitic stainless steel piping. This 

reduction in reported cracking is most 

likely due to one of the following 

factors: 

"* The most highly susceptible piping 

was replaced with piping fabricated 

from more IGSCC-resistant 

materials. 

"* Cracks in the most highly susceptible 

locations occurred relatively early in 

life and were identified and overlay

4-3 NUREG-1719
repaired during the early rounds of improvement, improved water 

inspections, chemistry and hydrogen water 

chemistry has reduced the potential 

Implementation of effective for IGSCC in the remaining welds. 

mitigation measures such as stress 

Table 4.2: Summary of Inspection Findings on Large Piping in 

BWRs Inspected According to IEBs 82-03 and 83-02 (a) 

Extent of Inspection Inspection Results Number of 

(% of Welds Inspected) (Number of Cracked Welds) Welds 

Repaired by 

Plant Recirculation RHR Recirculation RHR Overlay 

Big Rock Point 20%(11/59) 0 --- 0 

Browns Ferry 1 98% (103/109) 90% (36/40) 33 14 42 

Browns Ferry 2 27% (25/91) 28% (9/31) 2 0 0 

Browns Ferry 3 • 28% (9/31) 0 0 0 

Brunswick 1 25% (29/115) 75% (3/4) 3 0 3 

Brunswick 2 100% (102/102) 100% (5/5) 15 1 8 

Cooper 100% (108/108) 100% (7/7) 20 0 13 

Dresden 2 47% (47/101) 10%(4/40) 10 0 7 

Dresden 3 100% (115/115) 90%(45/50) 53(b) I11 o) 61(b) 

Duane Arnold 42% (49/117) 40% (2/5) 0 0 0 

FitzPatrick 47% (49/106) 45% (5/11) 1 0 0 

Hatch 1 47% (47/100) 100% (1l/1l) 5 2 6 

Hatch 2 94% (97/103) 100% (11/11) 36 3 27 

Millstone 1 11% (11/100) 0%(0/46) 0 0 0 

Monticello 100% (106/106) 78%(18/23) 6 0 6 

Nine Mile Pt. 1 82% (62/76) --- 53 0 0 

Oyster Creek 39% (31/80) --- 0 0 0 

Peach Bottom 2 100% (91/91) 91%(32/35) 19 7 24 

Peach Bottom 3 91% (77/85) 92% (35/38) 10 5 15 

Quad Cities 1 8% (9/110) 20% (9/44) 0 0 0 

Quad Cities 2 100% (106/106) 90% (45/50) 20 2 9 

Vermont Yankee 56% (58/88) 7% 2/30) 1 33 1 22

Notes: (a) The Boston Edison Company (BECo) inspected seven (7) welds at the Pilgrim Atomic Power Plant and detected cracks in 4 of the 

welds. BECo decided to replace the piping with Type 316NG stainless steel, and therefore terminated further inspections of the 

systems. 

(b) Eighteen welds that were originally reported to be cracked and later re-evaluated and determined not to be cracked are not included 

in these totals.

NUREG-1719 4-4

I I
5 CHRONOLOGY OF REGULATORY ACTIONS

5.0 Overview 

This chapter presents a chronology of 

significant U.S. regulatory activities 

addressing IGSCC in BWR piping. 

Three special Pipe Crack Study Groups 

were established to assess the IGSCC 

issue in BWR piping. These groups and 

their conclusions and recommendations 

were driven largely by operating 

experience and represent the evolution of 

an integrated and mature regulatory 

approach to managing IGSCC in BWR 

piping. 

5.1 First Pipe Crack Study Group

October 1975 

The first Pipe Crack Study Group was 

formed to coordinate and accelerate the 

Atomic Energy Commission's (AEC's) 

investigation of pipe cracking. The 

group published the results of its study 

in NUREG-75/067, "Technical Report, 

Investigation and Evaluation of Cracking 

in Austenitic Stainless Steel Piping 

of Boiling Water Reactor Plants." 

The recommendations stated in 

NUREG-75/067 were implemented in 

NUREG-0313, which was issued in 

July 1977. 

5.2 NUREG-0313-July 1977 

In NUREG-0313, "Technical Report on 

Material Selection and Processing 

Guidelines for BWR Coolant Pressure

Boundary Piping," which was issued in 

July 1977, the staff concluded that the 

probability was extremely low that 

IGSCC would propagate far enough to 

create a significant safety hazard, but 

that actions should be taken to avoid or 

manage IGSCC in BWR piping. The 

report presented (1) the NRC technical 

positions on actions to take to avoid 

IGSCC and (2) augmented inspection 

and leak detection guidelines to be 

applied in the event that actions to avoid 

IGSCC were not practical. 

5.3 Second Pipe Crack Study Group 

and NUREG-053 1--February 1979 

The NRC established the second Pipe 

Crack Study Group in 1978 to further 

study the IGSCC issue. This group was 

formed after IGSCC was reported in 

large-diameter (> 20-inch-diameter) 

piping in a German BWR and after 

significant IGSCC was discovered in a 

recirculation inlet nozzle safe-end at the 

Duane Arnold plant in the U.S. The 

charter of the second study group 

included consideration of (1) the 

implications of IGSCC in large diameter 

piping, (2) the resolution of issues 

related to ultrasonic testing (UT), (3) the 

implications of IGSCC in large diameter 

safe-ends, (4) the potential for IGSCC in 

PWR piping, and (5) evaluation of the 

Duane Arnold safe-end cracking and 

development of specific 

recommendations based on that

5-1 NUREG-1719
experience. In NUREG-0531, 

"Investigation and Evaluation of Stress

Corrosion Cracking in Piping of Light 

Water Reactor Plants" the study group 

documented its conclusions and 

recommendations. The report 

recommended avoiding further use of 

IGSCC-sensitive materials, considering 

mitigating actions (e.g., induction 

heating stress improvement(IHSI)), 

minimizing oxygen levels in BWR 

coolant, codifying IGSCC ultrasonic 

inspection methods, further optimizing 

of inspection methods, implementing 

augmented inspection programs, and 

performing further studies of operating 

factors influencing IGSCC. These study 

group recommendations were 

implemented by issuing NUREG-0313, 

Revision 1. 

5.4 NUREG-0313, Revision 1

July 1979 

Following completion of the second Pipe 

Crack Study Group investigation and 

issuance of NUREG-0531, in October 

1979, the staff revised NUREG-0313 to 

present updated guidance and 

recommendations regarding materials 

and processes that could be used to 

minimize IGSCC, and to make 

recommendations regarding 

augmentation of the extent and 

frequency of inservice inspections (ISIs) 

on welds considered to be susceptible to 

IGSCC. Revision 1 also contained 

recommendations about upgrading leak

detection systems and leakage limits 

from plants with susceptible welds. 

NUREG-1719

The staff issued these recommendations 

to the industry in Generic Letter 81-03, 

"Implementation of NUREG-0313, 

Revision l,"on February 26, 1981. 

5.5 Third Pipe Crack Study Group and 

NUREG-069 1-September 1980 

During 1979, several instances of 

cracking in pressurized-water reactor 

(PWR) piping systems led the NRC to 

establish the third Pipe Crack Study 

Group. Since the third Pipe Crack Study 

Group focused on cracking in PWR 

rather than BWR piping, the work of this 

group is not discussed in this paper. It is 

listed here only for completeness. 

5.6 Information Notice 82-39

September 1982 

After cracking was detected at Nine Mile 

Point Unit 1, the NRC Office of 

Inspection and Enforcement (IE) issued 

Information Notice 82-39, "Service 

Degradation of Thick Wall Stainless 

Steel Recirculation System Piping at a 

BWR Plant," on September 21, 1982, to 

alert all BWR licensees to the problem. 

5.7 Bulletin 82-03-October 1982 

Early in 1982, the NRC met with all 

BWR licensees to discuss plans for near

term inspections of welds in large

diameter recirculation piping, and on 

September 27, 1982, issued IE Bulletin 

82-03, "Stress Corrosion Cracking in 

Thick Wall, Large Diameter, Stainless 

Steel Recirculation system Piping at

5-2

I I
BWR Plants." The bulletin required that 

licensees of nine BWRs with outages 

scheduled through January 31, 1983, 

inspect a sample of the welds in the 

recirculation system during the next 

outage. In order to assure effective 

ultrasonic inspection, the bulletin 

required that procedures and competence 

of the examiners be demonstrated on 

samples of cracked piping removed from 

the Nine Mile Point recirculation system. 

5.8 Bulletin 83-02-March 1983 

After cracking was found in large pipes 

in a number of plants inspected in 

accordance with Bulletin 82-03, the staff 

issued another bulletin, Bulletin 

83-02, "Stress Corrosion Cracking in 

Large Diameter Recirculation System 

Piping at BWR Plants," in March 1983. 

Bulletin 83-02 extended the inspection 

requirement to all other BWRs with 

more than 2 years of operating service 

(14 units). It also required utilization of 

inspection personnel who had passed 

upgraded UT performance capability 

demonstrations. In addition, it required 

that a larger initial sample of welds be 

inspected than was required in 

accordance with IE Bulletin 82-03, and 

contained provisions for further increase 

in the inspection sample if cracks were 

found. 

5.9 NRC Orders-August 1983 

Because the inspections performed in 

accordance with Bulletin 83-02 

continued to indicate that cracking in

most BWRs was extensive and that many 

cracked welds required weld overlay 

repair for further operation, the NRC 

issued orders on August 26, 1983, to five 

plants to conduct as soon as practical the 

augmented inspection of IGSCC in 

large-diameter pipes in recirculation and 

residual heat removal (RHR) systems. 

5.10 Generic Letter 84-11-April 1984 

The results of the IGSCC inspections 

performed in accordance with the orders 

issued in August 1993 indicated that it 

was necessary to perform similar 

reinspections at all operating BWRs. In 

Generic Letter 84-11, that was issued on 

April 19, 1984, the staff presented 

criteria for conducting short-term 

reinspections, for detecting leaks and 

determining leakage limits, for 

evaluating flaw indications, and for 

making repairs. 

5.11 NDE Coordination Plan on 

Inspection Qualification

July 1984 

To gain confidence in the ability of 

ultrasonic inspection personnel to detect 

IGSCC cracks and determine their size, a 

program was established at the Electric 

Power Research Institute (EPRI) Non

destructive Examination (NDE) Center 

in Charlotte, North Carolina. This 

program was developed in accordance 

with an NDE Coordination Plan 

designed to provide training and to 

qualify the examination procedures, 

equipment, and inspection personnel

5-3 NUREG-1719
through demonstration on flawed 

samples. The NRC, EPRI and the 

BWR Owners Group (BWROG) agreed 

on the NDE Coordination Plan on 

July 3, 1984, and upgraded it in 

September 1985 (the plan is sometimes 

referred to as the tripartite agreement). 

5.12 Fourth Pipe Crack Study Group 

and NUREG-1061, Volume 1

August 1984 

The NRC established the fourth Pipe 

Crack Study Group in August 1984. The 

charter of the group called for 

developing an integrated plan to deal 

with the entirety of the stress corrosion 

cracking problem in BWR piping. The 

results of inspections conducted 

according to IE Bulletins 82-03 and 83

02 showed a wide variation in the extent 

of cracking reported. The scope of the 

fourth Pipe Crack Study Group was very 

similar to that of the second group. The 

main difference was that the fourth 

group related its recommendations to the 

regulatory documents that required 

revision in order to implement the 

recommendations. In NUREG-1061, 

Volume 1, the group documented its 

conclusions and recommendations. The 

report updated the discussions of the 

IGSCC mechanism, nondestructive 

testing for IGSCC, short-term and long

term approaches to mitigating IGSCC, 

and flaw evaluation, repair, and 

replacement options. The report 

contains specific recommendations in 

each of these areas with the intent of 

providing an integrated regulatory 

NUREG-1719

approach to managing IGSCC. It also 

summarized the status of domestic and 

foreign BWR pipe cracking. 

5.13 NUREG-0313, Revision 2

June 1986 and Generic 

Letter 88-01- January 1988 

Subjects covered by Revision 2 to 

NUREG-0313, which was published in 

June 1986, included recommendations 

regarding piping and weld material 

selection, special processing to minimize 

IGSCC susceptibility, improvements in 

BWR coolant chemistry and control, 

inspection requirements, repair methods, 

and leak detection. The conclusions and 

recommendations in the report are 

largely consistent with those made in 

NUREG-1061, Volume 1, prepared by 

the fourth Pipe Crack Study Group. The 

NRC positions stated in the report were 

implemented by Generic Letter 88-01, 

"NRC Position on IGSCC in BWR 

Austenitic Stainless Steel Piping," dated 

January 25, 1988. 

5.14 Generic Letter 88-01, 

Supplement 1-February 1992 

The staff issued Generic Letter 88-01, 

Supplement 1 on February 4, 1992. This 

supplement presented acceptable 

alternatives to some of the staff positions 

delineated in Generic Letter 88-01. The 

alternatives presented concerned 

requirements for inspecting of reactor 

water cleanup (RWCU) system piping 

outside the containment isolation valves

5-4

I 1
and requirements pertaining to the 

operability of leakage measurement 

instruments and the frequency of 

monitoring leakage rates. The 

supplement also clarified the staff's 

positions regarding the sample 

expansion of Category D welds, the 

effect of shrinkages resulting from weld 

overlay repairs and the need for technical 

amendments to incorporate IGSCC ISI 

and leak detection requirements, as 

delineated in GL 88-01, into the 

technical specifications. 

5.15 Appendix VIII and the 

Performance Demonstration 

Initiative-1 996 

The 1989 edition of Section XI of the 

ASME Code incorporated a new 

Appendix VII, "Qualification of 

Nondestructive Examination Personnel 

for Ultrasonic Examination," and the 

1989 addenda to the 1989 edition 

incorporated a new Appendix VIII, 

"Performance Demonstration for 

Ultrasonic Examination". Together,

these two appendices present code rules 

for qualifying personnel, techniques, and 

procedures. The qualification approach 

is intended to be performance based. In 

this regard, Appendix VIII does not 

present prescriptive code rules for 

inspection methods; rather, it relies on 

the demonstrated ability to detect and 

size defects in a prescribed number of 

test specimens. In order to implement 

the new-performance based approach, 

the industry developed the Performance 

Demonstration Initiative (PDI). This is a 

program, that is managed for the industry 

by EPRI, and which provides the 

necessary facilities, test samples, 

training, test procedures, and test 

administration to implement Appendix 

VIII. In March 1996, the NRC accepted 

the PDI qualification program for 

IGSCC in lieu of the program previously 

established under the NDE Coordination 

Plan (tripartite agreement), the NRC, 

EPRI, and the BWROG agreed to on 

July 3, 1984. The NRC is currently 

finalizing a revision to 1OCFR50.55a 

that will endorse Appendix VIII of the 

ASME Code.

5-5 NUREG-1719
6 MITIGATION METHODS

6.0 Overview 

As discussed in Chapter 2, IGSCC is 

controlled by three factors; material 

susceptibility, environment, and stress. 

Efforts to avoid or mitigate IGSCC focus 

on control of these three factors. 

Eliminating material susceptibility 

generally requires replacement of the 

piping. This subject is discussed in 

Chapter 10. Actions to mitigate IGSCC 

in operating plants, without piping 

replacement, focus on controlling 

environment or stresses. 

6.1 Water Chemistry Control 

Two approaches to modifying water 

chemistry have been used to mitigate 

IGSCC. The first approach has been to 

reduce the levels of impurities in the 

reactor coolant that can accelerate the 

initiation and propagation of IGSCC. 

The second approach has been to reduce 

the level of oxygen in the reactor coolant 

through the addition of hydrogen. 

Controlling the level of impurities in the 

bulk reactor coolant is important for a 

variety of reasons including 

minimization of IGSCC, fuel 

performance degradation, and radiation 

buildup. In 1973 the NRC issued 

Regulatory Guide (RG) 1.56, 

"Maintenance of Water Purity in Boiling 

Water Reactors," and revised it in 1974. 

These regulatory guides recommended 

that licensees establish maximum

conductivity levels that would (1) 

indicate breakthrough of one or more 

demineralizer units, (2) require orderly 

shutdown of the reactor, and (3) require 

immediate shutdown of the reactor. The 

regulatory guidance included 

representative conductivity levels for 

each of these conditions and 

representative chloride limits. However, 

research by the General Electric 

Company (GE) in the late 1970s showed 

that within the conductivity levels 

recommended in RG 1.56, dissolved 

anionic impurities could significantly 

affect IGSCC initiation and propagation. 

The two most common anions that were 

found to affect IGSCC were (1) sulfates 

introduced as a result of condenser 

leakage, chemical in-leakage, and resin 

ingress from the demineralizers; and 

(2) chlorides introduced as a result of 

condenser leakage. Chromates 

introduced through corrosion products 

can significantly contribute to room

temperature conductivity. This research 

resulted in the recognition that more 

stringent controls than those 

recommended in RG 1.56 were 

necessary to effect any change in the 

potential for IGSCC. Subsequently, in 

1986 EPRI published recommended 

water chemistry control guidelines, and 

in 1987 EPRI issued hydrogen water 

chemistry guidelines. EPRI has 

conducted additional, extensive research 

into BWR water chemistry and in 1996 

issued a revision to the BWR water 

chemistry guidelines that contains both

6-1 NUREG-1719
the water chemistry control guidelines 

and the hydrogen water chemistry 

guidelines in one document. The EPRI 

guidelines establish three operational 

conditions: cold shutdown, startup/hot 

standby, and power operation (at power 

levels above 10-percent power). For 

each of these operational conditions, the 

guidelines establish action levels that 

should not be exceeded. Action Level 1 

represents those impurity or chemistry 

parameter level which if exceeded, 

would threaten long-term system 

reliability of the plant system. The 

guideline suggests that the plant impurity 

or parameter conditions should be 

brought to levels below Action Level 1 

within 96 hours, or else a review should 

be performed and a program and 

schedule for implementing corrective 

actions should be developed and 

presented to management. Action Level 

2 represents the impurity or chemistry 

parameter level which, if exceeded, 

could cause damage to the plant system 

in the short-term, and if conditions are 

not reduced below the levels indicated 

by Action Level 2 within 24 hours, the 

plant should be brought to cold 

shutdown in an orderly fashion. Action 

Level 3 represents the impurity or 

chemistry parameter level which, if 

exceeded, indicates significant damage 

could occur to the plant system in the 

short-term and an orderly shutdown of 

the reactor should be initiated 

immediately. Action levels also are 

suggested in the guidelines for 

conductivity, chloride level, and sulfate 

level during cold shutdown and for

NUREG-1719

dissolved oxygen during startup and hot 

standby. For power operation, action 

levels are suggested for zinc and 

electrochemical potential in addition to 

the other action levels. Implementation 

of these guidelines can be beneficial in 

extending IGSCC initiation times and in 

reducing IGSCC growth rates. 

However, research and operating 

experience show that improving the 

water chemistry purity in an operating 

plant cannot, alone, eliminate IGSCC. 

U.S. BWRs are currently operating with 

much better water chemistries 

(conductivity < 0.3 pS/cm), which have 

reduced the potential for IGSCC. 

However, as noted in Chapter 3, once 

IGSCC has initiated, the microchemistry 

at the crack cannot be effectively altered 

through control of the bulk water 

chemistry. 

Hydrogen water chemistry (HWC) is a 

more aggressive approach for modifying 

the reactor coolant chemistry and has 

significant potential benefits. HWC 

involves the addition of hydrogen to the 

reactor coolant to reduce the level of 

dissolved oxygen in the coolant. By 

reducing the level of dissolved oxygen in 

the coolant, the electrochemical potential 

of the stainless steel will be shifted to a 

region in which IGSCC should not 

initiate or propagate. During the late 

1970s, laboratory testing and field 

testing using special test loops installed 

at Oskarshamn and Ringhals in Sweden, 

and at Dresden Unit 2 in the U.S., 

showed that reduction in typical BWR

6-2

I I
dissolved oxygen levels from 200 ppb to 

about 20 ppb could halt IGSCC 

propagation. 

Currently, about 15 BWRs in the U.S. 

operate with HWC. The operating 

experience for these plants has shown 

various levels of arrest of IGSCC, 

depending on how much hydrogen is 

added. 

It should be noted that use of HWC 

increases the radiation levels in the 

steam system because of increased levels 

of N- 16. Operating experience indicates 

an increase in radiation by a factor of up 

to 8, which can require additional 

radiation protection measures. Also, 

radiation fields from Co-60 have 

increased by more than 30 percent at 

some plants that have changed to HWC. 

As of November, 1993, 13 plants were 

injecting zinc into the feedwater to 

control Co-60 radiation levels. 

6.2 Stress Improvement 

Experimental and analytical studies 

show that significant variability can exist 

in the through-wall axial stress 

distributions of both small- and large

diameter pipe. But generally, the as

welded residual stress distributions in 

large-diameter pipes have a sinusoidal 

shape beginning with a high (on the 

order of 25 to 40 ksi) inside surface 

tensile stress, decreasing to a 

compressive stress field in the inside half 

of the pipe wall, and returning to a 

tensile stress field in the outer half of the

pipe wall. Several methods of stress 

improvement have been developed. 

These include induction heating stress 

improvement (IHSI), last pass heat sink 

welding (LPHSW), heat sink welding 

(HSW), and mechanical stress 

improvement. The objective of all these 

methods is to create a compressive, or at 

least lower, tensile stress field at the 

inside surface of the pipe. 

The IHSI process has been used 

extensively in the U.S.. In the IHSI 

process, induction coils are used to heat 

the pipe at the weld location and create a 

temperature gradient across the wall so 

that the material at the inside surface of 

the pipe yields in tension and the 

material at the outside surface of the pipe 

yields in compression. This results in a 

compressive residual stress field at the 

inside surface of the pipe. When 

combined with operating stresses, the 

resultant stress field may be tensile, but 

of much lower magnitude than the 

original as-welded stress field. 

Experimental work in the late 1970s and 

early 1980s showed improved life for 

defect-free specimens that received 

IHSI. The benefits of IHSI for 

specimens with pre-existing cracks were 

less clear, and concerns were expressed 

regarding the possible extension of 

preexisting cracks during the IHSI 

process, stress relaxation over time, and 

limited data with regard to pipe-to

component welds. It was also noted that 

the process parameters need to be 

carefully controlled to achieve the 

desired results. In NUREG 0313,

6-3 NUREG-1719
Revision 2, the staff allowed a reduction 

in the frequency of IGSCC inspections 

for susceptible welds treated with IHSI. 

For IGSCC-susceptible welds treated 

with IHSI in the first 2 years of 

operation, the staff reduced the 

inspection scope and frequency in 

NUREG 0313, Revision 2,the inspection 

scope and frequency from all welds 

every two refueling cycles to 50 percent 

of the welds every 10 years. All welds 

that received IHSI after 2 years of 

operation had to be inspected once 

within the two refueling outages 

following the IHSI application and then 

every 10 years following that inspection. 

A large number of susceptible welds 

have been mitigated with IHSI to 

improve their resistance to IGSCC. The 

operating experience of IHSI-treated 

welds has been reasonably good. 

However, the effectiveness of IHSI 

treatment has been questioned because 

small circumferential or axial cracks 

have been found on a few IHSI-treated 

welds. The probable root cause of this 

cracking has been attributed either to the 

failure of NDE examiners to detect the 

indications during the inspections that 

were performed preceding the 

application of IHSI or the failure of IHSI 

to achieve the desired compressive 

residual stress field on account of the 

complex weld joint geometry. IHSI has 

not been incorporated in the ASME 

Code; however, EPRI has presented 

industry guidelines regarding the proper 

procedures for effective application of 

IHSI.

NUREG-1719

In addition to IHSI, the mechanical 

stress improvement process (MSIP) has 

been developed as an alternate stress 

improvement process. Instead of using a 

large temperature gradient through the 

wall of the pipe to achieve the desired 

plastic strain pattern, MS]P uses 

mechanical methods stated to be less 

expensive and time-consuming, thereby 

exposing the technicians performing the 

operation to less radiation. In the MSIP, 

the pipe is "squeezed" concentrically at 

a location about 2 inches to one side of 

the weld being treated. The force is 

provided hydraulically, working through 

split rings and flexible metallic pads 

between the rings and the pipe. Through 

this process a small permanent reduction 

in diameter is achieved. After the 

equipment is removed, the elastic 

recovery of the pipe results in residual 

tensile stresses in the squeezed area, 

balanced by compressive residual 

stresses in the weld and heat-affected 

zone (HAZ) area at the inside surface of 

the pipe. The residual stress pattern 

produced by MSIP has been confirmed 

by finite element analyses. Some surface 

residual stresses were also 

experimentally evaluated and confirmed 

the analytical results. Through-wall 

residual stresses were measured on 

MSLP-treated 12-inch and 28-inch 

weldments. The axial compressive 

residual stresses were found to extend to 

almost 50 percent of the wall thickness. 

MSIP was accepted in NUREG 0313, 

Revision 2, as a stress improvement (SI) 

process for mitigation of IGSCC in 

BWR plants. This process has not been

6-4

I I
incorporated in the ASME Code. On the 

basis of a 1990 report, MSIP had been 

applied to 532 welds in 12 BWR plants 

in the world, including 157 nozzle welds 

ranging from 4-inch to 28-inch diameter. 

There is no reported cracking in MSIP

treated welds. 

Heat sink welding (HSW) is a process of 

butt welding pipes or fittings in which 

the major portion of the weld is 

produced while cooling water is flowing 

inside the pipe. The cooling effect of the 

water minimizes the sensitization caused 

by the welding process and, in addition, 

produces a steep temperature and stress 

gradient through the pipe wall during 

welding. After the weldment is cooled, 

the inner portion of the weld is under 

high compressive residual stress. The 

combination of reduced sensitization and 

high beneficial residual stresses provides

significant resistance to IGSCC. Last 

pass heat sink welding (LPHSW) is a 

welding process similar to HSW, except 

that only the last welding passes are 

performed when there is cooling water 

inside the pipe. Initial assessment of 

these processes in the early 1980s 

indicated that they were more sensitive 

to the weldment geometry than IHSI, and 

data on the effectiveness of the LPHSW 

process in mitigating IGSCC was 

limited. Thus, although NUREG-0313 

gave credit only for the HSW process in 

reducing inspections, these processes 

have been determined to be good 

practice in making repairs or 

replacements. There is very little 

operating experience reported on the 

effectiveness of these processes. The 

processes have not been incorporated 

into the ASME Code.

6-5 NUREG-1719
7 INSPECTION

7.0 Overview 

Extensive inspections have been a 

critical element of the NRC and U.S. 

industry's efforts to manage IGSCC in 

BWR piping. Inspection is key to 

identifying ongoing degradation, 

assessing the significance of the 

degradation and assuring the 

effectiveness of repairs. For an 

inspection program to be effective 

several key areas have to be addressed. 

These are 1) the qualification of 

inspection personnel and methods i.e., 

techniques and procedures; 2) 

definition of the inspection scope and 

frequency; and 3) criteria for 

dispositioning inspection results. The 

NRC and U.S. industry expended 

significant resources during the early 80s 

to improve the reliability of IGSCC 

inspections and continues to expend 

significant resources to perform effective 

inspections. 

7.1 Qualification of Inspection 

Personnel and Methods 

IGSCC inspection capabilities in the late 

1970s and early 1980s were in need of 

significant improvement. Plant 

examinations and laboratory studies 

showed that the inspection methods and 

personnel qualification requirements in 

place at the time were resulting in poor 

reliability of detection and poor flaw 

sizing capabilities. It was clear that a 

more rigorous approach to qualifying

inspection methods and personnel was 

needed. As discussed in Section 5.0, 

pipe crack study groups formed in 1975, 

1979, and 1984 assessed and made 

recommendations for improvements in 

ultrasonic inspection of IGSCC. This 

subject was addressed in the NUREG 

reports published by the pipe crack study 

groups and in NUREG-0313, Revision 

2, 'Technical Report on Material 

Selection and Processing Guidelines for 

BWR Coolant Pressure Boundary 

Piping." However, the first significant 

change in qualification requirements for 

NDE personnel and procedures came in 

1982, when the NRC issued IE 

Bulletin 82-03. Bulletin 82-03 

requested that operating BWR plants 

planning outages in 1982 must 

demonstrate the capability of their 

ultrasonic testing to detect IGSCC in 

large-diameter piping. 

In response to Bulletin 82-03, the 

industry established a limited ultrasonic 

performance demonstration at Battelle 

Columbus Laboratories. Using 

specimens removed from Nine Mile 

Point, the demonstration consisted of 

blind examinations of five pipe sections 

containing a total of five cracks. In 

order to assure ultrasonic capability 

sufficient to detect IGSCC, a passing 

criterion was established requiring 

detection of four of the five cracks. In 

response to the results of this 

demonstration and the inspections that 

followed, the NRC issued IE

7-1 NUREG-1719
Bulletin 83-02 requiring the continued 

inspection of piping using qualified 

personnel. This bulletin also established 

criteria for ultrasonic false calls and the 

number of inspection personnel 

representing a facility. 

During the same time period the industry 

had established the Boiling Water 

Reactor Owners Group (BWROG) 

IGSCC Research Program. One of the 

objectives of the program was the 

development of ultrasonic techniques 

that would minimize the outage time 

required for inspection and repair. To 

further this objective, the Electric Power 

Research Institute (EPRI) established the 

Nondestructive Evaluation (NDE) 

Center in Charlotte, North Carolina. The 

center was responsible for establishing a 

formal qualification program for the 

ultrasonic inspection of IGSCC. As 

discussed in Chapter 5, the BWROG, 

EPRI and the NRC agreed to this 

program on July 3, 1984 in a 

memorandum of understanding that 

became known as the "tripartite 

agreement." As a consequence, no 

ultrasonic inspector could perform 

augmented IGSCC inspections at power 

plants unless he or she was qualified by 

the EPRI NDE Center. 

Research continued at the NRC and 

EPRI into the accuracy of UT. The NRC 

and several U.S. corporations 

participated in an international series of 

round robin testing as well. The body of 

research led ultimately to the conclusion 

that UT, as performed at nuclear power 

NUREG-1719

stations in the U.S., had significant room 

for improvement. This issue was taken 

up by the American Society for 

Mechanical Engineers, Subcommittee 

Section XI. At the Subgroup for NDE, 

the need for a separate and 

comprehensive qualification program 

was established. This specification was 

established as mandatory Appendix VIII 

to the ASME Code and published in 

1989. The industry established a 

program to comply with Appendix VIII 

as the "Performance Demonstration 

Initiative" (PDI). This program for 

comprehensive qualifications was 

established at the NDE Center under the 

administration of EPRI personnel. 

In March 1996, the NRC agreed with the 

BWROG and EPRI that the 

performance qualification for IGSCC 

examiners could be administered through 

the PDI in lieu of the previous tripartite 

agreement. PDI currently provides the 

test specimens, training, procedures, and 

other services necessary to qualify 

IGSCC inspection personnel and 

methods, and the current PDI 

qualification carries a separate 

certification for the examination of 

piping for IGSCC. 

Details of the most effective inspection 

methods are available in the literature. 

A few key aspects of IGSCC inspection 

methodology are presented here. UT is 

the most common and effective method 

of inspection. Isotopic radiography 

(RT), using Ir-192 or Co-60 has not be 

shown to be very effective for the

7-2

I I
detection or sizing of IGSCC. A field

deployable high-energy X-ray was 

developed and used successfully at one 

plant. It was demonstrated as an 

effective technique for detecting IGSCC; 

however, the equipment was too 

cumbersome for regular use and the 

approach was abandoned. The UT 

methods demonstrated most effective for 

IGSCC detection are 60-degree and 

70-degree shear waves at 1.5 MHz and 

60-degree and 70-degree refracted 

longitudinal waves in the range of 

2-2.25 MHz. These configurations are 

used to generate comer trap and tip 

diffraction signals with a confirmatory 

creeping wave used to determine the 

degree of surface ligament to the tip. 

This method is effective in generating 

ultrasonic signals that are less 

susceptible to beam redirection caused 

by anisotropic wave velocities in the 

large weld grain size. A side benefit is 

the ability to direct the beam under 

interfering crown geometries, thus 

avoiding the need for a full "V" 

ultrasonic interrogation and its additional 

susceptibility to beam misdirection and 

complicated geometric calculations for 

flaw location. For simple length 

determination, amplitude drop is a 

conservative and readily applied method. 

7.2 Scope and Frequency of 

Inspections 

Extensive and frequent inspections are a 

major part of the NRC and U.S. industry 

approach to managing IGSCC in BWR 

piping. IE Bulletin 82-03, issued in

October 1982, and IE Bulletin 83-02, 

issued in March 1983, requested that 

licensees owing BWR-design reactors 

(nine with scheduled outages through 

January 31, 1983, and the remaining 

licensees owning BWRs with more than 

two years of amassed operating service) 

inspect the large diameter recirculation 

piping for their units. In addition, the 

NRC issued orders on August 26, 1983, 

to five plants requiring inspections to be 

performed as soon as practical. 

Inspections conducted in response to 

these bulletins and orders identified 

IGSCC in large-diameter recirculation 

and RHR systems. Subsequently, on 

April 19, 1984, the NRC issued Generic 

Letter 84-11 requesting all affected 

BWR licensees to implement an 

augmented inspection program for 

IGSCC reinspection. The reinspection 

had to be performed within 2 years from 

the previous inspection performed in 

accordance with the IE Bulletins and 

Orders. On the basis of the industry's 

inspection experience, the scope of the 

inspections requested in GL 84-11 was 

expanded to include all stainless steel 

welds, susceptible to IGSCC, in piping 

equal to or greater than 4 inches in 

diameter, or in systems operating over 

93.3 'C (200TF) that are part of or 

connected to the reactor coolant pressure 

boundary, out to the second isolation 

valve. 

The NRC positions stated in NUREG

0313, Revision 2, and implemented by 

Generic Letter 88-01 and its supplement 

constitute the primary basis for licensee

7-3 NUREG-1719
IGSCC inspection programs still being 

implemented today. All IGSCC

susceptible welds made of austenitic 

stainless steel and nickel-based materials 

are included, irrespective of ASME 

Code classification. The extent and 

frequency of the inspections were 

dependent on the resistance of the 

materials and the effectiveness of any 

processes used to prevent cracking. 

Welds were categorized as "A" through 

"G," in accordance with their 

susceptiblity to IGSCC based on material 

type, processing history, and applied 

mitigating actions. A substantial initial 

inspection scope and aggressive 

expansion criteria (capable of driving to 

a 100 percent inspection) were required. 

The inspection sample was to be biased

NUREG-1719

toward those welds with the highest 

propensity for cracking. It was further 

noted that the inspection sample should 

be based on operating experience and 

that other factors that could increase the 

susceptibility to IGSCC, such as weld 

preparation, excessive grinding, 

extensive repairs, and high stress 

locations, be considered in developing 

the inspection program. As a result of 

implementing mitigating measures or of 

replacing piping, the scopes and 

frequencies of inspection may have 

changed (within the bounds of the 

guidance in GL 88-01 and its 

supplement); however, ultrasonic 

inspections remain a key aspect of 

managing IGSCC.

7-4

I I
8 FLAW EVALUATION METHODS

Early versions of Section XI of the 

ASME Code included flaw evaluation 

criteria in Table IWB 3514-3. These 

criteria allowed only small flaws (e.g., 

essentially 10 percent of the wall 

thickness or less) to remain in service. 

In the Winter Addenda to the 1983 

edition of the ASME Code, a new 

section, IWB 3640, was added to the 

code whereby larger flaws could be 

analyzed and accepted for continued 

service in austenitic piping. The fracture 

mechanics analyses in IWB 3640 for 

austenitic stainless steels were based on 

limit load analyses. 1WB 3640 also 

requires that the potential flaw growth 

during the period of operation be 

included in the analysis. In NUREG

0313, Revision 2, the NRC staff 

presented fairly detailed guidance on 

how to analyze IGSCC cracks for 

continued service. Two of the more 

important analysis areas addressed in 

Revision 2 of NUREG-0313 and in GL 

84-11 were (1) uncertainties in weld 

residual stresses and crack growth rates 

and (2) uncertainties in UT sizing. 

To address the uncertainties in crack 

growth rate, NUREG-0313, Revision 2, 

established a residual stress distribution 

and stress-intensity crack-growth 

relationship that were expected to 

provide a conservative estimate of 

IGSCC growth rate during the analyzed 

operating period.

The staff also established an acceptance 

criterion in GL 84-11 for continued 

operation of the plant without having to 

perform a repair of the piping system. In 

order to address UT measurement 

uncertainties, this criterion limited the 

calculated flaw dimensions at the end of 

the operating period to be within two

thirds of the acceptable flaw dimensions 

listed in IWB 3640. This assumption 

allows for UT sizing uncertainty up to 

100 percent of the crack depth for cracks 

up to 25 percent of the wall thickness. 

In NUREG-0313, Revision 2, 

uncertainties for UT sizing were 

considered only if the ultrasonic 

examination was performed by personnel 

not fully qualified in accordance with the 

NRC, EPRI, and BWROG NDE 

Coordination Plan, or if limitations to the 

examination (e.g., wide weld crowns), 

obstructions, or other adverse geometric 

configurations existed. For such 

examinations, the flaw was analyzed 

assuming the depth was at least 75 

percent of the pipe wall thickness. 

For low-toughness materials such as 

fluxed welds fabricated using shielded 

metal arc welding (SAW) or submerged 

metal arc welding (SMAW), additional 

criteria to evaluate flaws in such 

materials are given in IWB 3640 of the 

1989 edition of the ASME Code. This is 

because for these less tough materials, 

crack extension and pipe failure may 

occur at load levels below fully plastic

8-1 NUREG-1719
limit load. IWB 3640 includes factors to 

account for the lower fracture toughness 

of these materials. It also includes the 

NUREG-1719

effect of pipe size and secondary stresses 

in the flaw evaluation of such materials.

8-2

I III
9 REPAIR METHODS

IGSCC in BWR piping initiates at the 

inner surface of the pipe near the weld 

root. The cracks commonly propagate 

along the heat-affected zone (HAZ) 

adjacent to the weld toward the outside 

diameter surface of the pipe. Therefore, 

it is not practical to perform code repair 

of cracked welds because the normal 

code repair requires grinding out the 

defective area and backfilling the area 

with weld metal. 

In NUREG-0313, Revision 2, the staff 

identifies the following repair methods: 

"* reinforcement by weld overlay 

"* partial replacement 

"* SI (for minor circumferential cracks 

with crack depths less than 30 

percent of the wall thickness, crack 

lengths less than 10 percent of the 

piping circumference, and service 

stress less than the allowable Sm 

value cited in the ASME Code) 

"* approved clamping devices 

Reinforcement using weld overlays 

consists of applying weld metal on the 

outside diameter of the pipe over the 

weld and beyond the weld on both sides. 

This is done completely around the 

circumference of the pipe. IGSCC

resistant low-carbon, high-ferrite, type 

308L weld metal is used, and the process 

is usually performed with an automatic 

welding machine using the gas tungsten 

arc weld (GTAW) or gas metal arc weld

(GMAW) process. Weld overlaying can 

be performed with water in the pipe 

during welding, without the need for 

draining the pipe during repair. 

A limit load analysis is used to determine 

the minimum acceptable thickness of the 

weld overlay design. The criteria for the 

limit load analysis is given in IWB 3640 

of Section XI of the ASME Boiler and 

Pressure Vessel Code. NRC accepts two 

approaches for weld overlay designs: 

(1) the standard overlay design and 

(2) specially designed overlay. The 

standard overlay design provides a 

nominal margin of 2.77 against limit 

load failure, assuming that the crack 

goes completely (360 degrees) around 

the circumference and through the wall 

of the pipe. For minor cracking 

extending less than about 10 percent of 

the pipe's circumference, with no more 

than four axial cracks, a specially 

designed overlay can be used for repair. 

Credit for part of the uncracked area of 

the original pipe is taken in the special 

overlay design. The special overlay 

design also extends 360 degrees around 

the pipe circumference but may be 

thinner than the standard overlay. One 

application of the special overlay design 

is to prevent leakage from minor axial 

cracking. 

The implementation of weld overlay 

repair has been delineated in Code Case 

N-504 as an acceptable alternative for

9-1 NUREG-1719
repair of Class 1, 2, and 3 austenitic 

stainless steel piping. 

In accordance with GL 88-01, the 

cracked welds reinforced by weld 

NUREG-1719

overlays or mitigated by SI are required 

to be inspected every two refueling 

cycles to ensure that the structural 

integrity of the repaired or mitigated 

weld joint is maintained.

9-2

I I II
10 PIPING REPLACEMENT

As discussed in Chapter 2, IGSCC can 

be avoided through a proper selection of 

materials and fabrication processes. In 

NUREG-0313, Revision 2, the staff 

made specific recommendations 

regarding the selection of materials and 

processes to be used when replacing the 

original IGSCC-susceptible piping 

materials. 

In NUREG-0313, the staff provided the 

following guidance for selecting 

materials that would be considered 

resistant to sensitization, and therefore to 

IGSCC: 

" Base metals of low carbon wrought 

austenitic stainless steels with 

designations 304L, 304NG, 316NG, 

and similar low carbon grade steels 

with a maximum carbon content of 

0.035 percent were considered to be 

IGSCC resistant. 

" Type 347 austenitic stainless steel, as 

modified for nuclear use, would be 

considered to be resistant with a 

somewhat higher carbon content, 

the usual maximum of 0.04 percent 

being an adequate limit for carbon 

content. 

" Austenitic stainless steel materials 

not meeting these criteria were 

considered to be resistant if given a 

solution heat treatment after welding.

" Low carbon weld metals, including 

types 308L, 316L, 309L, and similar 

grades, with a maximum carbon 

content of 0.035 percent and a 

minimum of 7.5 percent (or FN) 

ferrite, as deposited, were considered 

resistant. Low carbon weld filler 

material especially developed for 

joining modified type 347 was 

considered resistant. 

" Welds joining resistant materials that 

meet the ASME Code requirement of 

5 percent (or FN) ferrite, but that are 

below 7.5 percent, were to be 

evaluated on a case by-case-basis 

considering carbon content and other 

factors affecting their sensitivity. 

" Piping weldments were considered 

resistant to IGSCC if the weld heat

affected zone on the inside of the 

pipe is protected by a cladding of 

resistant weld metal, often referred to 

as "corrosion-resistant" cladding. 

" Cast austenitic stainless steels with a 

maximum of 0.035 percent carbon 

and a minimum of 7.5 percent (or 

FN) ferrite were considered resistant. 

Weld joints between resistant piping 

and cast valve or pump bodies that do 

not meet these requirements were 

subject to more inspection.

10-1 NUREG-1719
Other austenitic materials, including 

nickel-based alloys such as Inconel 

600, were to be evaluated on a case

by-case basis. Inconel 82 was the 

only commonly used nickel-based 

weld material considered to be 

resistant. 

The NRC staff also recommended in 

NUREG-0313, Revision 2, that no 

austenitic material be considered 

resistant to IGSCC in the presence of a 

crevice, such as is formed by a partial 

penetration weld, where the crevice is 

exposed to reactor coolant. 

NUREG-1719

In NUREG-0313, Revision 2, the staff 

discussed several process controls that 

may be used to mitigate a material's 

susceptibility to IGSCC. The following 

processes were considered to be 

qualified for providing resistance to 

IGSCC in BWR pipe welds: 

(1) solution heat treatment (SHT) 

(2) heat sink welding (HSW) 

(3) induction heating stress 

improvement (IHSI) 

(1) mechanical stress improvement 

process (MSIP)

10-2

I I
1 LEAK DETECTION AND LEAK-BEFORE-BREAK

As noted in the introduction to this 

report, the NRC has not accepted leak

before-break (LBB) as a regulatory basis 

for assuring the integrity of piping 

systems susceptible to IGSCC. The 

NRC's position on LBB is established in 

Criterion 4 to 10 CFR Part 50 Appendix 

A, in the associated statements of 

consideration (SOCs) for this Criterion, 

and in draft Standard Review Plan (SRP) 

Section 3.6.3. In total, these documents 

establish the position that for piping 

systems that are demonstrated to satisfy 

LBB criteria, the dynamic effects (e.g., 

blowdown loads, pipe whip reactions, 

etc.) associated with high energy pipe 

breaks can be eliminated from the design 

basis. However, even when LBB criteria 

are satisfied, the design basis for 

emergency core cooling systems and 

containment are not changed. 

Furthermore, the NRC's position is that 

piping systems that are subject to active 

degradation modes (e.g., IGSCC, 

fatigue, or erosion-corrosion) do not 

satisfy the criteria for application of 

LBB. Therefore, LBB does not provide 

a regulatory basis for assuring the 

integrity of BWR piping. NRC's 

approach has been one of defense-in

depth but with particular emphasis on 

inspection in order to avoid challenges to 

operators and safety systems. 

Nonetheless, NRC does recognize that 

LBB can be an important element in a 

defense-in-depth approach to assuring

piping integrity. Operating experience 

shows that careful monitoring for 

leakage and prompt corrective actions 

can identify conditions that, if left 

unaddressed, could result in pipe failure. 

Therefore, the NRC has established 

stringent requirements for monitoring 

leakage in BWRs. 

In NUREG- 1061, Volume 1, the fourth 

Pipe Crack Study Group recommended 

a decrease in allowable unidentified 

leakage in BWRs from 1140 L/hr 

(5 gpm) to 684 L/hr (3 gpm). However, 

upon further evaluation, the staff 

concluded that the existing leakage 

limits, if effectively implemented along 

with other recommendations in the areas 

of inspection, and so forth, would 

provide an acceptable level of 

protection, and that the reduction in the 

leakage limit could not be supported by a 

regulatory analysis performed in 

accordance with the "Backfit Rule" (e.g., 

10 CFR 50.109.c). 

In NUREG-0313, Revision 2, and in 

Generic Letter 88-01, the NRC staff 

established the position that leakage 

detection systems should be in 

conformance with Position C of 

Regulatory Guide 1.45, "Reactor 

Coolant Pressure Boundary Leakage 

Detection Systems," or else as otherwise 

approved by the NRC. The following 

items summarize this position in more 

detail:

11-1 NUREG-1719
" Plant shutdown should be initiated 

for inspections and corrective action 

when, within any period of 24 hours 

or less, any leakage detection system 

indicates an increase in rate of 

unidentified leakage in excess of 

456 L/hr (2 gpm) or its equivalent, or 

when the total unidentified leakage 

attains a rate of 1140 L/hr (5 gpm) or 

equivalent, whichever comes first. 

For sump level monitoring systems 

with fixed measurement intervals, the 

level should be monitored at 

approximately 4-hour intervals or 

less. 

" Unidentified leakage should include 

all leakage other than the following: 

(a) leakage into closed systems, 

such as pump seal or valve 

packing leaks that are captured, 

flow metered, and conducted to 

a sump or collecting tank, or 

(b) leakage into the containment 

atmosphere from sources that 

are both specifically located and 

known either not to interfere 

with the operations of 

unidentified leakage monitoring 

systems or not to be from a 

through wall crack in the piping 

within the reactor coolant 

pressure boundary. 

" For plants operating with any IGSCC 

Category D, E, F, or G welds (i.e., 

welds with a high susceptibility to 

IGSCC, at least one of the leakage 

NUREG-1719

measurement instruments associated 

with each sump shall be 

operable, and the outage time for 

inoperable instruments shall be 

limited to 24 hours, or immediately 

initiate an orderly shutdown. 

The preceding position emphasizes 

several important aspects that are 

necessary in order to identify and take 

corrective action for defective piping 

systems that do exhibit LBB. These are 

(1) timely leakage measurements, (2) 

criteria related to rate of increase in 

leakage, and (3) prompt corrective 

action. In order for LBB to be effective, 

leakage must be measured on a timely 

basis (i.e., as close to real time as 

possible). Many systems that exhibit 

LBB show large (e.g., exponential) 

increases in leakage as the defect 

becomes more significant. Delays in 

measurements or monitoring of leakage 

could result in loss of opportunity for 

timely corrective action. Similarly, 

criteria that are based solely on 

magnitude of leakage may not recognize 

a rapidly developing situation. 

Monitoring the trend in leakage makes 

better use of available data and provides 

the opportunity for more timely action. 

Finally, operators must have clear 

procedures for responding to increasing 

unidentified leakage. Rapid, decisive 

response is critical in taking advantage 

of LBB to avoid gross component 

failure.

11-2

I I
12 CURRENT STATUS OF BWR PIPING

In Table 12.1, the staff presents the 

status of piping in operating U.S. BWRs. 

The table indicates whether full or partial 

replacements of recirculation and RHR 

system piping has occurred and whether 

or not the unit has implemented 

hydrogen water chemistry. 

As indicated in Table 12.1, 12 BWR 

plants in the U.S. have replaced either all 

or part of the main recirculation system 

stainless steel piping with low carbon 

stainless steel materials. The materials 

used in the piping replacement are either 

316NG or 316SS with low carbon 

content. Eight plants have replaced all 

the piping in the recirculation and RHR 

systems. One plant has replaced all the 

piping in the recirculation system and a 

portion of the piping in the RHR system. 

Three plants have replaced only the riser 

portion of the recirculation system 

piping. 

Hydrogen water chemistry (HWC) has 

been applied at 16 BWR plants. 

At least 20 domestic BWRs have found 

IGSCC in large-diameter stainless steel

piping. Except for very minor cracking, 

almost all cracked welds inside the 

containment were repaired with weld 

overlays. However, in general, operating 

BWRs have not operated with cracked 

welds for more than one or two fuel 

cycles, because repair or mitigation has 

the benefit of requiring less inspection. 

Those plants that have not replaced all 

susceptible piping or that did not startup 

with resistant piping are currently 

operating with numerous overlay 

repaired welds. 

Since the staff issued GL 88-01, the U.S. 

industry has been performing inspections 

for IGSCC on a regular basis. The 

inspection schedule for each weld is 

based on the degree of susceptibility to 

IGSCC, as recommended in the guidance 

of GL 88-01 and its supplement. In the 

early rounds of inspection, many large 

pipe welds were found to be cracked. 

However, as discussed in Section 4.0, as 

a result of piping replacements, repairs, 

and implementation of other measures to 

reduce the sensitivity to IGSCC, there 

have been very few recently reported 

events of IGSCC in U.S. BWRs.

12-1 NUREG-1719
Table 12.1 Status of U.S. BWR Piping

Plant Date of Original Replacements Replacements Hydrogen 

Operating Design for for Water 

Design Name License Material Recirculation RHR Pipingcb) Chemistry I Piping Implemented? 

Nine Mile 12/26/74 304SS or Full, 316SS Part, 316SS Yes 

Point 1 316SS (low carbon) (low carbon) 

BWR2 

Oyster Creek 08/01/69 304SS or None None Yes 

316SS 

Dresden 2 02/21/70 304SS or None None Yes 

316SS 

Dresden 3 03/02/70 304SS or Full, 316NG Full, 316NG No 

316SS 

Millstone 1 10/31/86 304SS or None None Yes 

316SS 

BWR3 Monticello 01/09/81 304SS or Full, 316NG Full, 316NG Yes 

316SS 

Pilgrim 09/15/72 304SS or Full, 316NG Full, 316NG Yes 

316SS 

Quad Cities 1 12/14/72 304SS or None None Yes 

316SS 

Quad Cities 2 12/14/72 304SS or None None Yes 

316SS 

Browns Ferry 1 12/20/73 304SS or None None No 

316SS 

Browns Ferry 2 08/02/74 304SS or Part (a) (riser) None No 

316SS 316NG 

Browns Ferry 3 08/18/76 304SS or None None No 

BWR4 316SS 

Brunswick 1 11/12/76 304SS or Part (a) (riser) None Yes 

316SS 316NG 

Brunswick 2 12/27/74 304SS or Part (a) (riser) None Yes 

316SS 316NG

NUREG-1719 12-2

I II
Date of Original Replacements Replacements Hydrogen 

Plant Operating Design for for Water 

Design 1 License Material Recirculation RHR Pipinge Chemistry Namegn Piping Implemented?

Cooper 01/18/74 304SS or 

316SS

Full, 316NG Full, 316NG No

Duane Arnold 02/20/74 304SS or None None Yes 

316SS 

Fermi 2 07/15/85 304SS or None None Yes 

316SS 

FitzPatrick 10/17/74 304SS or None None Yes 

316SS 

Hatch 1 10/13/74 304SS or None None Yes 

316SS 

Hatch 2 06/13/78 304SS or Full Full Yes 

316SS 316NG 316NG 

Hope Creek 07/25/86 316NG N/A N/A Yes 

REC, RHR 

RWCU 

Limerick 1 08/08/85 316NG N/A N/A Yes 

REC, RHR, 

Core Spray, 

RWCU 

Limerick 2 08/25/89 316NG N/A N/A Yes 

REC, RHR, 

Core Spray, 

RWCU 

Peach Bottom 2 12/14/73 304SS or Full Full Yes 

316SS 316NG 316NG 

Peach Bottom 3 07/02/74 304SS or Full Full Yes 

316SS 316NG 316NG 

Susquehanna 11/12/82 304SS or None None Yes 

Unit 1 316SS 

Susquehanna 06/27/84 304SS or None None No 

Unit 2 316SS

Vermont 

Yankee

02/28/73 304SS or 

316SS

Full 

316NG

Full 

316NG

No

NUREG-1719

BWR4 

cont.

12-3
Date of Original Replacements Replacements Hydrogen 

Plant Operating Design for for Water 

License Material Recirculation RHR Pipinge) Chemistry Design Name Piping Implemented? 

La Salle 1 08/13/82 304SS or None None No 

316SSL(c) 

La Salle 2 03/23/84 304SS or None None No 

BWR5 316SSL(o) 

Nine Mile 07/02/87 316NG for N/A N/A No 

Point 2 All Piping 

Systems 

WNP 2 04/13/84 304SS or None None No 

316SS 

Clinton 1 04/17/87 316NG for N/A None No 

REC, RWCU 

Grand Gulf 1 11/01/84 304SS or None None Yes 

316SS 

BWR6 Perry 1 11/13/86 304SS or None None No 

316SS 

River Bend 1 11/20/85 316NG for N/A None No 

I___ _ I___ _REC I I _I 

Notes: 

(a) Recirculation system riser piping only. 

(b) Residual Heat Removal piping inside containment that is classified as ASME Code Class 1 pipe. 

(c) 12 inch inlet safe-ends 

Abbreviation Descriptions: 

Full - full replacement of the piping 

Part - partial replacement of the piping 

304SS - Type 304 austenitic stainless steel 

316SS - Type 316 austenitic stainless steel 

316NG - Type 316 austenitic stainless steel, nuclear grade quality 

None - no replacement of the piping performed to date 

N/A - initial material of the piping is already Type 316NG steel; replacement is not applicable in this case 

REC - Recirculation SystemPiping 

RWCU - Reactor Water Cleanup System Piping 

RHR - Residual Heat Removal System Piping

NUREG-1719 12-4

I I
13 CONCLUSIONS

Development of the U.S. regulatory 

framework for managing IGSCC in 

BWR piping was an evolutionary 

process driven by operating experience, 

research, and technological 

developments. IGSCC was initially 

observed in only small-diameter piping; 

however, in time it affected both large

and small-diameter piping systems. 

A key element in the management of 

IGSCC was an aggressive inspection 

program utilizing qualified inspection

techniques and personnel. In parallel, 

the development of qualified repair and 

mitigation methods, along with 

replacement of piping with IGSCC

resistant materials resulted in an 

effective and economic solution to the 

problem. Today, a combination of 

regulatory documents, code rules, and 

industry guidelines are available to 

support effective programs for managing 

IGSCC in the piping of U.S. BWRs.

13-1 NUREG-1719
14 REFERENCES

NRC NUREG Reports 

1. NUREG-75-067, "Technical Report, Investigation and Evaluation of Cracking in 

Austenitic Stainless Steel Piping of Boiling Water Reactor Plants," U.S. Nuclear 

Regulatory Commission, 1975. 

2. NUREG-0313, "Technical Report on Material Selection and Processing Guidelines 

for BWR Coolant Pressure Boundary Piping," U.S. Nuclear Regulatory 

Commission, July 1977. 

3. NUREG-0313, Revision 1, "Technical Report on Material Selection and Processing 

Guidelines for BWR Coolant Pressure Boundary Piping," U.S. Nuclear Regulatory 

Commission, October 1979. 

4. NUREG-0313, Revision 2, 'Technical Report on Material Selection and Processing 

Guidelines for BWR Coolant Pressure Boundary Piping," U.S. Nuclear Regulatory 

Commission, January 1988. 

5. NUREG-0376, "Residual Stresses at Girth-Butt Welds in Pipes and Pressure 

Vessels," U.S. Nuclear Regulatory Commission, November 1977. 

6. NUREG-0531, "Investigation and Evaluation of Stress-Corrosion Cracking in 

Piping of Light Water Reactor Plants," U.S. Nuclear Regulatory Commission, 

February 1979. 

7. NUREG-1061, Volume 1, "Report of the U.S. Nuclear Regulatory Commission 

Piping Review Committee, Investigation and Evaluation of Stress Corrosion 

Cracking in Piping of Boiling Water Reactor Plants," U.S. Nuclear Regulatory 

Commission, August 1984. 

8. NUREG/CP-0077, "LEAK-BEFORE-BREAK: International Policies and 

Supporting Research," U.S. Nuclear Regulatory Commission, June 1986. 

9. NUREG/CR-4464, PNL-5705, "Performance Demonstration Tests for Detection of 

Intergranular Stress Corrosion Cracking," U.S. Nuclear Regulatory Commission, 

February 1986.

14-1 NUREG-1719
10. NUREG/CR-4697, BMI-2142, "An Assessment of Circumferentially Complex

Cracked Pipe Subjected to Bending," U.S. Nuclear Regulatory Commission, 

October 1986. 

11. NUREG/CR-4813, ANL-86-52, "Assessment of Leak Detection Systems for 

LWRs," U.S. Nuclear Regulatory Commission, October 1988. 

12. NUREG/CR-4853, "Approximate Methods for Fracture Analyses of Through-Wall 

Cracked Pipes," U.S. Nuclear Regulatory Commission, February 1987. 

NRC Bulletins, Generic Letters and Information Notices 

13. Bulletin 82-03, "Stress Corrosion Cracking in Thick Wall, Large Diameter, 

Stainless Steel Recirculation System Piping at BWR Plants," U.S. Nuclear 

Regulatory Commission, September 27, 1982. 

14. Bulletin 83-02, "Stress Corrosion Cracking in Large Diameter Recirculation System 

Piping at BWR Plants," U.S. Nuclear Regulatory Commission, March 1983. 

15. Generic Letter 81-03, "Implementation of NUREG-0313, Revision 1," U.S. Nuclear 

Regulatory Commission, February 26, 1981 

16. Generic Letter 84-11, "Inspections of BWR Stainless Steel Piping," U.S. Nuclear 

Regulatory Commission, April 19, 1984. 

17. Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel 

Piping," U.S. Nuclear Regulatory Commission, January 25, 1988. 

18. Generic Letter 88-01, Supplement 1, "NRC Position on IGSCC in BWR Austenitic 

Stainless Steel Piping," U.S. Nuclear Regulatory Commission, February 4, 1992. 

19. IE Information Notice 82-39, "Service Degradation of Thick Wall Stainless Steel 

Recirculation System Piping at a BWR Plant," U.S. Nuclear Regulatory 

Commission, September 21, 1982. 

EPRI Reports 

20. EPRI NP-81-4LD, "Residual Stress Improvement by Means of Induction Heating," 

Electric Power Research Institute, March 1981. 

NUREG-1719 14-2
21. EPRI NP-2347, "Instability Predictions for Circumferentially Cracked Type-304 

Stainless Steel Pipes Under Dynamic Loading," Electric Power Research Institute, 

April, 1982. 

22. EPRI NP-2472-SY, "The Growth and Stability of Stress Corrosion Cracks in Large

Diameter BWR Piping," Electric Power Research Institute, July 1982. 

23. EPRI WS-79-174, Volumes 1-4, "Countermeasures for Pipe Cracking in BWRs," 

Electric Power Research Institute, May 1980. 

Miscellaneous References 

24. J.S. Porowski, W.J. O'Donnell, M.L. Badlani, and E.J. Hampton, "Use of the 

Mechanical Stress Improvement Process to Mitigate Stress Corrosion Cracking in 

BWR Piping System," Nuclear Engineering and Design, Volume 124, pp. 91-100, 

(1990)

14-3 NUREG-1719
APPENDIX A

ABBREVIATIONS 

Chemistry Abbreviations 

Cr Chromium 

Ir- 192 Irridium Isotope 192 

Co-60 Cobalt Isotope 60 

Terminology Abbreviations 

ASME American Society of Mechanical Engineers 

BWR boiling water reactor 

BWROG Boiling Water Reactor Owners Group 

CRD control rod drive 

ECCS emergency core cooling system 

EPRI Electrical Power Research Institute 

FN ferrite number 

GE General Electric Company 

GL Generic Letter 

GMAW gas metal arc welding 

GTAW gas tungsten arc welding 

HAZ heat-affected zone 

HPCS high pressure coolant spray 

HSW heat sink weld 

HWC hydrogen water chemistry 

IE Office of Inspection and Enforcement 

IGSCC intergranular stress corrosion cracking 

IHSI induction heating stress improvement 

ISI inservice inspection

A-1 NUREG-1719
LBB leak-before-break 

LPCI low pressure coolant injection 

LPHSW last pass heat sink welding 

LPCS low pressure coolant spray 

MSIP mechanical stress improvement process 

NDE nondestructive examination 

PDI performance demonstration initiative 

PWR pressurized-water reactor 

RCIC reactor core isolation cooling 

RG Regulatory Guide 

RHR residual heat removal 

RT Radiography Testing 

RWCU reactor water cleanup 

SAW submerged arc welding 

SHT solution heat treatment 

SLC standby liquid control 

SMAW shielded metal arc welding 

SOC statement of consideration 

SRP standard review plan 

UT ultrasonic testing 

Unit Abbreviations 

gpm gallons per minute, English System unit of flow 

L/hr liters per hour, SI Metric System unit of flow 

MPa megapascal, SI Metric System unit of stress or pressure 

ppb parts per billion, a unit of concentration 

psi pounds per square inch, English System unit of stress or pressure °C degrees Celsius, Metric System unit of temperature 

OF degrees Fahrenheit, English System unit of temperature 

pS/cm micro-Siemens per centimeter, a metric unit for the measurement of 

conductivity 

MHz megahertz, a unit for frequency equal to one million cycles per second 

NUREG-1719 A-2
Pipe Cracking in U.S. BWRs: A Regulatory History

- I NRC FORM 335 

(2-N9) 

NRCM 1102, 

3201,.3202

5. AUTHOR(S) 6. TYPE OF REPOHT 

J.R. Strosnider, Jr., W.H. Koo, J.A. Davis, and M.C. Modes Technical 

7. PERIOD COVERED pncdusie Datf) 

Dec. 1969 - Nov. 1999 

.8. PERFORMING ORGANIZATION -NAME AND ADDRESS (If NRC, provide Diison, Oftfe or Region, U.& NuCarR.gutoryCommission, and maaVdua; ontrator 

provide name and mailig addres ) 

Division of Engineering 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, D.C. 20555-0001

4. FIN OR GRANT NUMBER

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type Same asao , ifoontractor, provide NRC Diision, Oft@ orRegion. U.S. Nuclear Regulatory Commission, and mat aedisss.)

"Same as above"

10. SUPPLEMENTARY NOTES 

In support of the IAEA Regional Workshop, Slavutych, Ukraine. June 1988.

11. ABSTRACT Ow was of iss) 

This paper was prepared, in part, to support discussions at an IAEA Regional Workshop on Environmentally Assisted Cracking of Nuclear Power Plant Austenitic 

Piping that was held in Slavutych, Ukraine from 22 to 26 June 1988. This paper presents a regulatory history and offers a perspective on intergranular stress 

corrosion cracking (IGSCC) in U.S. boiling-water reactor piping. The paper focuses 

on regulatory and industry actions taken to assure that U.S. licensees manage IGSCC in a manner that provides safe and reliable plant operation. Although the 

paper does not offer extensive theoretical details on the IGSCC phenomenon, it 

does discuss some of the key technical issues that influenced regulatory positions 

and industry actions.

12. KEY WORDS/DESCRIPTORS (List Wods orphrss etiat wWasslat rwsa mhers in ioc.:ng me repor) 1& AVAILABIUTY STATEMENT 

boiling-water reactor (BWR); intergranular stress corrosion unlimited 

cracking (IGSCC); recirculation system piping; residual heat 14. SECURJTYCLASSIFICATION 

removal (RHR) system piping; reactor water cleanup (RWCU) system M*P,•,• piping; leak-before-break (LBB); austenitic stainless steel; unclassified 

fracture toughness; ultrasonic testing (UT); non-destructive Mear!) 

examination (NDE); weld overlay; stress improvement (SI); unclassified 

hydrogen water chemistry (HWC) 15. NUMBER OF PAGES 

16. PRICE

UJS. NUCLEAR REGULATORY COMMIbS!0N 

BIBLIOGRAPHIC DATA SHEET 

(See instructions on fte nverse)

2. TITLE AND SUBTITLE

1. REPORT NUMBER 

(Assigned by NRC, Add Vol, Supp., Rev., 

and Addendum Numbers, It any.) 

NUREG-1719 

3. DATE REPORT PUBLISHED 

MOH 2YEAR 

May 2000

NRC• FORM 33 (2-NI)
Federal Recycling Program 

I I I I
PIPE CRACKING IN U.S. BWRs: A REGULATORY HISTORY

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001

MAY 2000

FIRST CLASS MAIL 

POSTAGE AND FEES PAID 

USNRC 

PERMIT NO. G-67

NUREG-1719