POLICY ISSUE

INFORMATION

May 10, 2004 SECY-04-0078

FOR: The Commissioners

FROM: William D. Travers

Executive Director for Operations

SUBJECT: PROPOSED BULLETIN: “INSPECTION OF ALLOY 82/182/600 MATERIALS

USED IN THE FABRICATION OF PRESSURIZER PENETRATIONS AND

STEAM SPACE PIPING CONNECTIONS AT PRESSURIZED-WATER

REACTORS”

PURPOSE:

To inform the Commission of the staff’s intent to issue the attached bulletin to address

concerns about the inspection of pressurizer penetrations and steam space piping connections

at pressurized-water reactors (PWRs).

BACKGROUND:

The pressurizer and its penetrations and connections are an integral part of the reactor coolant

pressure boundary (RCPB), and their integrity is important to the safe operation of nuclear

power plants. Since the 1980s, facility operating experience has demonstrated that 

nickel-based alloys, including Alloys 82, 182, and 600, when exposed to the typical environment

found in PWR pressurizers, exhibit degradation due to primary water stress corrosion cracking

(PWSCC). In the pressurizer boundary, PWSCC has been observed at a variety of locations,

including (1) pressurizer heater sleeves at Combustion Engineering (CE)-designed facilities, 

(2) pressurizer heater bundles at Babcock and Wilcox-designed facilities, (3) instrument line

connections at several PWRs, and (4) in steam space piping connections at a foreign and a

domestic facility.

CONTACTS: Timothy G. Colburn, NRR/DIPM

301-415-1402

Matthew A. Mitchell, NRR/DE

301-415-3303
The Commissioners 2

Recent operational events which support the Nuclear Regulatory Commission staff’s conclusion

that issuance of the attached bulletin at this time is warranted include the observation of

circumferential PWSCC in the non-pressure-boundary portion of five heater sleeves at Palo

Verde, Unit 2, in the fall of 2003. Previously, evidence from finite element modeling studies and

limited nondestructive evaluation (NDE) of leaking heater sleeves at CE-designed facilities

suggested that PWSCC in these pressurizer penetrations could be assumed to be axially

oriented. The recent evidence of circumferential cracking of the non-pressure-boundary portion

of pressurizer heater sleeves at Palo Verde, Unit 2, challenges this assumption. Based on

experience with the evolution of PWSCC degradation in other locations (e.g., reactor pressure

vessel upper head penetrations), evidence of circumferential PWSCC at a given location has

usually been observed after axial PWSCC was discovered at that location. 

The Palo Verde, Unit 2 experience indicates that there is an increasing potential for the onset of

circumferential cracking of the RCPB portion of Alloy 82/182/600 materials used in the

fabrication of pressurizer penetrations and steam space piping connections. A bulletin is being

issued to determine if existing licensee degradation management programs are able to identify

in an effective and timely manner the onset of circumferential RCPB cracking at these

locations. The identification of the onset of circumferential PWSCC in Alloy 82/182/600

materials exposed to the high temperature pressurizer environment at a given facility is critical,

due to the potential for initiation and growth of PWSCC flaws and the potential for

circumferential cracking to result in abnormal leakage, rapidly propagating failure, and/or gross

rupture of the RCPB.

The NRC staff is currently working on a number of initiatives, including participating in American

Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code activities and

evaluating the potential for revising of Title 10, Section 50.55a, of the Code of Federal

Regulations (10 CFR 50.55a), which, if brought to successful completion, will address the

inspection of RCPB locations covered by the attached bulletin. However, the NRC staff expects

that the completion of the aforementioned activities may take an extended period of time to

complete. Therefore, the information requested by the attached bulletin is vital to support the

NRC staff’s ability to determine if any near-term regulatory action is required to ensure that

licensees are adequately protecting the integrity of the RCPB at their facilities.

DISCUSSION:

Under the provisions of Section 182a of the Atomic Energy Act of 1954, as amended, and

10 CFR 50.54(f), the attached bulletin transmits an information request, asking licensees to

verify compliance with existing applicable regulatory requirements (see the Applicable

Regulatory Requirements section of the bulletin). Specifically, the attached bulletin requests

information to enable the staff to determine whether current inspection practices ensure the

integrity of pressurizer penetrations and steam space piping connections at PWRs, as required

by NRC regulations and plant-specific licenses. The requested information will also enable the

staff to determine whether the subject PWR addressees need to augment their inspection

practices to ensure that cracking of pressurizer penetrations and steam space piping

connections will not pose any undue risk to the public health and safety. The staff will use the

requested information in assessing the need for, and guiding the development of, additional

regulatory actions (e.g., generic communications, orders, or rulemaking) to address the integrity

of the RCPB.
The Commissioners 3

The degradation management program identified in the bulletin as being acceptable to the NRC

staff is similar to the recommendations made by the Westinghouse Owners Group and the

industry’s Materials Reliability Program. Nonetheless, the staff recognizes the need to have

continued communication with the industry and other stakeholders to effectively resolve the

concerns identified in the bulletin. Toward that end, the staff plans to hold a public meeting with

external stakeholders to explain the staff’s rationale for issuing the bulletin and to answer

questions about the information the licensees should provide in their responses. The staff

expects that this effort will help to ensure that complete, high quality responses will be received

from licensees, which will in turn expedite the staff’s review of the licensees’ responses. The

staff intends to issue the attached bulletin in May 2004 so that plants that have scheduled

refueling outages during the remainder of calendar year 2004 can plan their pressurizer

inspections accordingly.

COORDINATION:

The staff briefed the Committee To Review Generic Requirements (CRGR) on the proposed

bulletin on March 23, 2004, and has addressed the Committee’s comments. The CRGR has

endorsed this bulletin.

The staff briefed the Advisory Committee on Reactor Safeguards (ACRS) Full Committee on

the proposed bulletin on April 16, 2004. The ACRS members raised no objection to the

issuance of the proposed bulletin.

The Office of the General Counsel (OGC) has reviewed the proposed bulletin and has no legal

objections to its content. In addition, OGC has determined that the proposed bulletin does not

constitute a “rule” under the Small Business Regulatory Enforcement Fairness Act of 1996.

By memorandum dated March 29, 2004, the staff issued a communication plan associated with

this bulletin. This paper is being forwarded to the Commission in accordance with the

communication plan.

/RA/

William D. Travers

Executive Director

 for Operations

Attachment: Proposed NRC Bulletin, “Inspection of Alloy 82/182/600 Materials Used in the

 Fabrication of Pressurizer Penetrations and Steam Space Piping Connections

 at Pressurized-Water Reactors”
The Commissioners 3

The degradation management program identified in the bulletin as being acceptable to the NRC

staff is similar to the recommendations made by the Westinghouse Owners Group and the

industry’s Materials Reliability Program. Nonetheless, the staff recognizes the need to have

continued communication with the industry and other stakeholders to effectively resolve the

concerns identified in the bulletin. Toward that end, the staff plans to hold a public meeting with

external stakeholders to explain the staff’s rationale for issuing the bulletin and to answer

questions about the information the licensees should provide in their responses. The staff

expects that this effort will help to ensure that complete, high quality responses will be received

from licensees, which will in turn expedite the staff’s review of the licensees’ responses. The

staff intends to issue the attached bulletin in May 2004 so that plants that have scheduled

refueling outages during the remainder of calendar year 2004 can plan their pressurizer

inspections accordingly.

COORDINATION:

The staff briefed the Committee To Review Generic Requirements (CRGR) on the proposed

bulletin on March 23, 2004, and has addressed the Committee’s comments. The CRGR has

endorsed this bulletin.

The staff briefed the Advisory Committee on Reactor Safeguards (ACRS) Full Committee on

the proposed bulletin on April 16, 2004. The ACRS members raised no objection to the

issuance of the proposed bulletin.

The Office of the General Counsel (OGC) has reviewed the proposed bulletin and has no legal

objections to its content. In addition, OGC has determined that the proposed bulletin does not

constitute a “rule” under the Small Business Regulatory Enforcement Fairness Act of 1996.

By memorandum dated March 29, 2004, the staff issued a communication plan associated with

this bulletin. This paper is being forwarded to the Commission in accordance with the

communication plan.

 /RA/

William D. Travers

Executive Director

 for Operations

Attachment: Proposed NRC Bulletin, “Inspection of Alloy 82/182/600 Materials Used in the

 Fabrication of Pressurizer Penetrations and Steam Space Piping Connections

 at Pressurized-Water Reactors”

 ADAMS ACCESSION NUMBER.: ML041030130 

INDICATE IN BOX: “C”=COPY W/O ATTACHMENT/ENCLOSURE, “E”=COPY W/ATT/ENCL, “N”=NO COPY 

OFFICE PM/LA:DLPM EMCB:DE EMCB:DE DIPM:IROB

NAME TColburn MMitchell WBateman CPetrone

DATE 04/19/2004 04/19/2004 04/19/2004 04/19/2004

OFFICE TECHED D:DE PD:IROB OGC D:DIPM

NAME PKleene RBarrett(CGrimesfor) WBeckner RHoefling BBoger( CCarpenter

for)

DATE 04/16/2004 04/20/2004 04/21/04 04/27/04 04/22/04

OFFICE ADPT D:NRR EDO

NAME BSheron JDyer WTravers

DATE 04/26/2004 04/29/2004 05/10/2004

OFFICIAL RECORD COPY 
Attachment

OMB Control No.: 3150-0012 

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, DC 20555

NRC BULLETIN 2004-xx: INSPECTION OF ALLOY 82/182/600 MATERIALS USED 

IN THE FABRICATION OF PRESSURIZER PENETRATIONS

AND STEAM SPACE PIPING CONNECTIONS AT

PRESSURIZED-WATER REACTORS

Addressees

All holders of operating licenses for pressurized-water nuclear power reactors (PWRs), except

those who have permanently ceased operations and have certified that fuel has been

permanently removed from the reactor pressure vessel (RPV).

All other holders of operating licenses for nuclear power plants will receive a copy of this bulletin

for information.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this bulletin to:

(1) advise PWR licensees that current methods of inspecting Alloy 82/182/600

materials used in the fabrication of pressurizer penetrations and steam space

piping connections may need to be supplemented with additional measures to

detect and adequately characterize flaws due to primary water stress corrosion

cracking (PWSCC),

(2) request PWR addressees to provide the NRC with information related to the

materials from which the pressurizer penetrations and steam space piping

connections at their facilities were fabricated,

(3) request PWR licensees to provide the NRC with information related to the

inspections that have been and those that will be performed to ensure that

degradation of Alloy 82/182/600 materials used in the fabrication of pressurizer

penetrations and steam space piping connections will be identified, adequately

characterized, and repaired, and

(4) require PWR addresses to provide a written response to the NRC in accordance

with the provisions of Section 50.54(f) of Title 10 of the Code of Federal

Regulations (10 CFR 50.54(f)).
BL 2004-xx

Page 2 of 11

Background

Typical PWR pressurizers may have multiple penetrations and/or steam space piping

connections which were fabricated using Alloy 600 base materials and/or Alloy 82/182 weld

materials. Such penetrations may include, but are not limited to, pressurizer heater

penetrations (including the “heater bundle” design used at Babcock and Wilcox-designed

(B&W-designed) PWRs) and instrument line penetrations. 

Steam space piping connections may include, but are not limited to, full penetration butt welds

connecting power-operated relief valves, safety valves, spray lines, and vent lines to the steam

space of the pressurizer. It should be noted that pressurizer surge line welds are not intended

to be within the scope of this bulletin. The NRC staff is considering separately whether any

additional action or information is required to address piping butt welds in the pressurizer surge

line and throughout the rest of the reactor coolant system. 

Operating experience has demonstrated that Alloy 82/182/600 materials exposed to primary

coolant water (or steam) at the normal operating conditions of PWR plants have cracked due to

PWSCC. The NRC has previously issued generic communications regarding the emergence of

this phenomena, and its consequential effects, in other areas of PWR primary systems. NRC

Bulletin 2001-01, “Circumferential Cracking of Reactor Pressure Vessel Head Penetration

Nozzles,” addressed PWSCC of control rod drive mechanism penetrations and other

penetrations in the RPV upper heads of PWRs. NRC Bulletin 2002-01, “Reactor Pressure

Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity,” addressed the

issue of boric acid corrosion of low alloy steel components as a result of leakage through

PWSCC-induced flaws in the reactor coolant pressure boundary (RCPB). NRC Bulletin

2002-02, “Reactor Pressure Vessel Head and Vessel Head Penetration Nozzle Inspection

Programs,” followed up on NRC staff concerns regarding the adequacy of visual examinations

as a primary inspection method for the RPV upper head and RPV upper head penetrations. 

Finally, NRC Bulletin 2003-02, “Leakage From Reactor Pressure Vessel Lower Head

Penetrations and Reactor Coolant Pressure Boundary Integrity,” addressed the potential for

PWSCC of RPV bottom mounted instrumentation nozzles.

Operating experience, both domestic and foreign, has demonstrated that Alloy 82/182/600

materials connected to a PWR’s pressurizer may be particularly susceptible to PWSCC. Since

the late 1980's, approximately 50 Alloy 600 pressurizer heater sleeves at Combustion

Engineering-designed (CE-designed) facilities in the United States have shown evidence of

RCPB leakage which has been attributed to PWSCC. The most recent events of this type

occurred at Millstone, Unit 2, and Waterford, Unit 3, in October 2003, and at Palo Verde, Unit 3,

in February 2004. All available evidence from finite element modeling studies and limited

nondestructive evaluation (NDE) has suggested that these leakage events were the result of

axially-oriented PWSCC of the pressure boundary portion of these heater sleeves. However, 

NDE results from Palo Verde, Unit 2's fall 2003 refueling outage, on heater sleeves which had

not shown evidence of leakage have demonstrated that circumferentially-oriented

PWSCC can occur in the non-pressure boundary portion (i.e., above the J-groove attachment

weld) of these components. Cracking in a TMI-1 diaphragm plate was attributed to PWSCC in

the heat affected zone of the seal weld. Boric acid corrosion of the low alloy steel strongback

was also observed to have resulted from the leakage.
BL 2004-xx

Page 3 of 11 

Small diameter Alloy 82/182 instrument line penetrations have also shown evidence of PWSCC

at many PWR facilities since the 1980’s. For example, in October 2003, the Crystal River,

Unit 3, licensee reported RCPB leakage from three pressurizer upper level instrument tap

nozzles, which are exposed to the steam space in the pressurizer. The leakage was attributed

to PWSCC of Alloy 82/182/600 material from which the connections were constructed.

Finally, inspection results from September 2003 at Tsuruga, Unit 2, in Japan are relevant with

respect to PWSCC in larger diameter, butt welded lines connected to the steam space of the

pressurizer. Evidence of boron deposits on the surface of a pressurizer relief valve nozzle

(inside diameter 130 mm, or approximately 5 inches) lead to the discovery of five

axially-oriented flaws in the nickel-based alloy weld material used in the fabrication of the

nozzle-to-safe end weld. Subsequent NDE performed on a safety valve nozzle of similar

diameter resulted in the discovery of two additional flaws in its nozzle-to-safe end weld. 

Fractographic analysis of the flaw surfaces confirmed PWSCC as the mechanism for flaw

initiation and growth. This event at Tsuruga, Unit 2 was similar to an event at Palisades in 1993

where leakage was observed and attributed to a circumferentially-oriented PWSCC flaw in a

line leading to the unit’s power-operated relief valves.

The occurrence of PWSCC in Alloy 82/182/600 materials used in the fabrication of pressurizer

penetrations and steam space piping connections is not surprising. The initiation and growth of

PWSCC flaws is known to be strongly dependent on the temperature of the primary system

water to which the Alloy 82/182/600 materials are exposed. Given the fact that at the

pressurizer the reactor coolant system environment attains a temperature of about 650F

(343 C), PWSCC should be expected to occur in these materials and an effective degradation

management program is warranted.

Discussion

The pressurizer and its penetrations and connections are an integral part of the RCPB, and

their integrity is important to the safe operation of the plant. Since the onset of PWSCC in Alloy

82/182/600 materials used in the fabrication of pressurizer penetrations in the late 1980s,

inspection programs have been based on the identification of RCPB leakage and the repair of

leaking penetrations. The acceptability of such a degradation management program has been,

in large part, predicated on the assumption that PWSCC in the RCPB at these locations will

always be axially-oriented, limited in extent, and readily detectable through visual inspections

and/or on-line leakage monitoring systems inside containment.

The recent evidence of circumferential cracking of the non-pressure boundary portion of

pressurizer heater sleeves at Palo Verde, Unit 2, challenges this assumption. 

Based upon NRC experience with the evolution of PWSCC degradation in other parts of the

RCPB (e.g., RPV upper head penetrations), evidence of circumferential PWSCC at a given

RCPB location has usually been observed after axial PWSCC at that location was discovered. 

The experience with Alloy 82/182/600 pressurizer heater sleeves in CE-designed facilities 

through the fall 2003 inspections at Palo Verde, Unit 2, appears to substantiate that a similar

trend may be evident at these locations.
BL 2004-xx

Page 4 of 11

This experience indicates that there is an increasing potential for the onset of circumferential

cracking of the RCPB portion of Alloy 82/182/600 materials used in the fabrication of

pressurizer penetrations and steam space piping connections. The NRC is also concerned

about the ability of existing licensee degradation management programs to identify the onset of

circumferential RCPB cracking at these locations in an effective and timely manner. The

identification of the onset of circumferential PWSCC in Alloy 82/182/600 materials exposed to

the high temperature pressurizer environment at a given facility is critical, due to the potential

for initiation and growth of PWSCC flaws rates and the potential for circumferential cracking to

result in abnormal leakage, rapidly propagating failure, and/or gross rupture of the RCPB.

The small amount of leakage from the cracks that have been discovered to date in Alloy

82/182/600 pressurizer penetrations and steam space piping connections does not represent

an immediate safety problem. Safety systems included in plant designs and required to be

available during plant operation would be able to mitigate the effects of even more significant

leaks, up to and including the gross rupture of the largest piping connection to the pressurizer

shell. However, to maintain the overall defense-in-depth philosophy incorporated into the

design and operation of nuclear power plants, licensees should take appropriate actions to

ensure that the integrity of pressurizer penetrations and steam space piping connections is

maintained and the potential for challenging facility safety systems is minimized.

The NRC staff believes that it is appropriate for licensees to assess their current inspection

practices to ensure that the integrity of pressurizer penetrations and steam space piping

connections is being maintained. Inspections capable of detecting through-wall leakage from

any pressurizer penetration or steam space piping connection, beginning at the next refueling

outage, would provide confidence in the integrity of the locations. If visual inspections are

performed to detect evidence of possible leakage, a bare metal examination of 100% of the

circumference of each Alloy 82/182/600 pressurizer penetration and steam space piping

connection would be effective at finding leakage due to PWSCC flaws and acceptable to the

NRC staff.

The industry’s Materials Reliability Program (MRP) has recommended that PWR licensees 

perform bare metal visual inspections of all Alloy 82/182/600 primary system pressure boundary

locations normally operated at greater than or equal to 350F within a facility’s next two

refueling outages, unless an equivalent inspection was performed during the facility’s most

recent refueling outage. This recommendation would be applicable to all Alloy 82/182/600

pressurizer penetrations and steam space piping connections. This recommendation was 

included in a letter from Leslie Hartz, Chair, MRP Senior Representatives, dated January 20,

2004, [ADAMS Accession# ML040360483] to all PWR licensees. 

The MRP is an industry program, coordinated by the Electric Power Research Institute (EPRI),

to address materials-related issues associated with PWRs. In addition, recommendations

regarding the inspection of the CE-designed fleet’s pressurizer heater sleeves were provided by 
BL 2004-xx

Page 5 of 11

the Westinghouse Owners Group (WOG) [ADAMS Accession# ML040480309]. The WOG

recommended that licensees of CE-designed facilities should, at each refueling outage:

(1) perform a bare metal visual inspection of 100% of all pressurizer heater sleeve

locations in such a way that visual access to the bare metal 360 degrees around

each sleeve can be attained,

(2) perform NDE capable of characterizing crack orientation of all sleeves for which

visual inspection shows evidence of leakage. The NDE of each leaking sleeve

will be performed prior to repair of the sleeve, and

(3) if the NDE defines the flaw as potential circumferential cracking below the sleeve

attachment weld, the NRC will be notified immediately and an appropriate

inspection plan developed. The plan will define additional sleeves to be

inspected by NDE sufficient to determine the extent of condition commensurate

with the characterization of the flaw.

The NRC staff has reviewed the recommendations made by the WOG in its January 30, 2004,

letter and finds that, with minor modifications, it would constitute an effective degradation

management program for all Alloy 82/182/600 covered under the scope of this bulletin based on

our current state of knowledge. The use of bare metal visual examinations is warranted given

that significant PWSCC flaws may only produce small amounts of boric acid deposits that may

not be visible unless a 100% bare metal visual inspection is conducted. As a modification of

item (3) above, the NRC staff believes that the topic of NDE scope expansion should be

discussed with the NRC staff if circumferential PWSCC is observed in either the pressure

boundary or non-pressure boundary portions of any locations covered under the scope of this

bulletin to ensure that the licensee has performed an adequate extent-of-condition evaluation. 

A facility which has seen evidence of non-pressure boundary circumferential PWSCC may be

more susceptible to circumferential PWSCC of the pressure boundary in “like locations” (those

similar in design, function, fabrication, etc.) and thus may need to perform additional NDE to

verify that like locations within the pressurizer boundary retain adequate structural integrity

before returning the unit to power operation. A licensee who does not intend to perform

additional NDE upon the identification of non-pressure boundary circumferential PWSCC

should be prepared to provide a technical basis to explain why the observation of non-pressure

boundary circumferential PWSCC is not a relevant indication with respect to the potential for

pressure boundary circumferential PWSCC in other like locations within the boundary of its

pressurizer.

The NRC staff is working with the industry and other stakeholders to revise the American

Society of Mechanical Engineers (ASME) Code and NRC regulations to address inspection of

RCPB locations susceptible to cracking, including pressurizer penetrations and steam space

piping connections. The NRC has posted and will continue to post information about this

subject on its Web site (www.nrc.gov).
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Page 6 of 11

Applicable Regulatory Requirements

Several provisions of the NRC regulations and plant operating licenses (addressed in plant

technical specifications (TSs)) pertain to RCPB integrity and the issues addressed by this

bulletin. The general design criteria (GDC) for nuclear power plants (Appendix A to 10 CFR

Part 50), or, as appropriate, similar requirements in the licensing basis for a reactor facility, the

requirements of 10 CFR 50.55a, and the quality assurance criteria of Appendix B to 10 CFR 

Part 50, provide the bases and requirements for NRC staff assessment of the potential for, and

consequences of, degradation of the RCPB.

The applicable GDCs include GDC 14 (Reactor coolant pressure boundary), GDC 31 (Fracture

prevention of reactor coolant pressure boundary), and GDC 32 (Inspection of reactor coolant

pressure boundary). GDC 14 specifies that the RCPB be designed, fabricated, erected, and

tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating

failure, and of gross rupture. GDC 31 specifies that the probability of rapidly propagating

fracture of the RCPB be minimized. GDC 32 specifies that components which are part of the

RCPB have the capability of being periodically inspected to assess their structural and leaktight

integrity.

NRC regulations at 10 CFR 50.55a state that ASME Code Class 1 components (which include

the RCPB) must meet the requirements of Section XI of the ASME Code. Various portions of

the ASME Code address RCPB inspection. For example, Table IWB-2500-1 of Section XI of

the ASME Code provides examination requirements during system leakage testing of all

pressure-retaining components of the RCPB and references IWB-3522 for acceptance

standards. IWB-3522.1(c) and (e) specify that conditions requiring correction include the

detection of leakage from insulated components and discoloration or accumulated residues on

the surfaces of components, insulation, or floor areas that may be evidence of borated water 

leakage, with leakage defined as the through-wall leakage that penetrates the pressure

retaining membrane. Therefore, 10 CFR 50.55a, by reference to the ASME Code, does not

permit through-wall degradation of pressurizer penetrations and steam space piping

connections. For through-wall leakage identified by visual examinations in accordance with the

ASME Code, acceptance standards for the identified degradation are provided in IWB-3142. 

Specifically, supplemental examination (by surface or volumetric examination), corrective

measures or repairs, analytical evaluation, and replacement provide methods for determining

the acceptability of degraded components.

Criterion V (Instructions, Procedures, and Drawings) of Appendix B to 10 CFR Part 50 states

that activities affecting quality shall be prescribed by documented instructions, procedures, or

drawings of a type appropriate to the circumstances and shall be accomplished in accordance

with these instructions, procedures, or drawings. Criterion V further states that instructions,

procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria

for determining that important activities have been satisfactorily accomplished. Visual and

volumetric examinations of the RCPB are activities that should be documented in accordance

with these requirements.

Criterion IX (Control of Special Processes) of Appendix B to 10 CFR Part 50 states that special

processes, including nondestructive testing, shall be controlled and accomplished by qualified 
BL 2004-XX

Page 7 of 11

personnel using qualified procedures in accordance with applicable codes, standards,

specifications, criteria, and other special requirements.

Criterion XVI (Corrective Action) of Appendix B to 10 CFR Part 50 states that measures shall be

established to assure that conditions adverse to quality are promptly identified and corrected.

For significant conditions adverse to quality, the measures taken shall include root cause

determination and corrective action to preclude repetition of the adverse conditions.

For degradation of the RCPB, the root cause determination is important for understanding the

nature of the degradation present and the required actions to mitigate future degradation. 

These actions could include proactive inspections and repair of potentially degraded portions of

the RCPB. 

Plant TSs pertain to this issue insofar as they do not allow operation with known through-wall

reactor coolant system pressure boundary leakage.

Requested Information

(1) All subject PWR licensees are requested to provide the following information

within 60 days of the date of this bulletin. 

(a) A description of the pressurizer penetrations and steam space piping

connections at your plant. At a minimum, this description should include

materials of construction (e.g., stainless steel piping and/or weld metal,

Alloy 600 piping/sleeves, Alloy 82/182 weld metal or buttering, etc.), joint

design (e.g., partial penetration welds, full penetration welds, bolted

connections, etc.), and, in the case of welded joints, whether or not the

weld was stress-relieved prior to being put into service. Additional

information relevant with respect to determining the susceptibility of your

plant’s pressurizer penetrations and steam space piping connections to

PWSCC should also be included.

(b) A description of the inspection program for Alloy 82/182/600 pressurizer

penetrations and steam space piping connections that has been

implemented at your plant. The description should include when the

inspections were performed; the areas, penetrations and steam space

piping connections inspected; the extent (percentage) of coverage

achieved for each location which was inspected; the inspection methods

used; the process used to resolve any inspection findings; the quality of

the documentation of the inspections (e.g., written report, video record,

photographs); and, the basis for concluding that your plant satisfies

applicable regulatory requirements related to the integrity of pressurizer

penetrations and steam space piping connections. If leaking pressurizer

penetrations or steam space piping connections were found, indicate

what followup NDE was performed to characterize flaws in the leaking

penetrations.

(c) A description of the Alloy 82/182/600 pressurizer penetration and steam

space piping connection inspection program that will be implemented at 
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your plant during the next and subsequent refueling outages. The

description should include the areas, penetrations and steam space 

piping connections to be inspected; the extent (percentage) of coverage

to be achieved for each location; inspection methods to be used;

qualification standards for the inspection methods and personnel; the

process used to resolve any inspection indications; the inspection

documentation to be generated; and the basis for concluding that your

plant will satisfy applicable regulatory requirements related to the

structural and leakage integrity of pressurizer penetrations and steam

space piping connections. If leaking pressurizer penetrations or steam

space piping connections are found, indicate what followup NDE will be

performed to characterize flaws in the leaking penetrations. Provide your

plans for expansion of the scope of NDE to be performed if

circumferential flaws are found in any portion of the leaking pressurizer

penetrations or steam space piping connections.

(d) In light of the information discussed in this bulletin and your

understanding of the relevance of recent industry operating experience to

your facility, explain why the inspection program identified in your

response to item (1)(c) above is adequate for the purpose of maintaining

the integrity of your facility’s RCPB and for meeting all applicable

regulatory requirements which pertain to your facility.

(2) Within 60 days of plant restart following the next inspection of the Alloy

82/182/600 pressurizer penetrations and steam space piping connections, the

subject PWR licensees should either:

(a) submit to the NRC a statement indicating that the inspections described

in the licensee’s response to item (1)(c) of this bulletin were completed

and a description of the as-found condition of the pressurizer shell, any

findings of relevant indications of through-wall leakage, followup NDE

performed to characterize flaws in leaking penetrations or steam space

piping connections, a summary of all relevant indications found by NDE,

a summary of the disposition of any findings of boric acid, and any

corrective actions taken and/or repairs made as a result of the indications

found,

or

(b) if the licensee was unable to complete the inspections described in

response to item (1)(c) of this bulletin, submit to the NRC a summary of

the inspections performed, the extent of the inspections, the methods

used, a description of the as-found condition of the pressurizer shell, any

findings of relevant indications of through-wall leakage, followup NDE

performed to characterize flaws in leaking penetrations or steam space

piping connections, a summary of all relevant indications found by NDE,

a summary of the disposition of any findings of boric acid, and any

corrective actions taken and/or repairs made as a result of the indications
BL 2004-xx

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found. In addition, supplement the answer which you provided to item

(1)(d) above to explain why the inspections that you completed were

adequate for the purpose of maintaining the integrity of your facility’s

RCPB and for meeting all applicable regulatory requirements which

pertain to your facility.

For lines attached directly to the pressurizer, with the exception of the

surge line, the information requested in (1) and (2) above should be

provided for any locations, including those remote from the pressurizer

shell, which contain Alloy 82/182/600 materials which are exposed to

conditions similar to those of the pressurizer environment.

Required Response

In accordance with 10 CFR 50.54(f), the subject PWR addressees are required to submit a

written response to this bulletin. This information is sought to verify licensees’ compliance with

the current licensing basis for the subject PWR addressees. 

The addressees have two options: 

(1) addressees may choose to submit written responses providing the information

requested above within the requested time periods, or

(2) addressees who choose not to provide the information requested or cannot meet

the requested completion dates are required to submit written responses within

15 days of the date of this bulletin. The responses must address any alternative

course of action proposed, including the basis for the acceptability of the

proposed alternative course of action.

The required written responses should be addressed to the U.S. Nuclear Regulatory

Commission, ATTN: Document Control Desk, 11555 Rockville Pike, Rockville, Maryland

20852, under oath or affirmation under the provisions of Section 182a of the Atomic Energy Act

of 1954, as amended, and 10 CFR 50.54(f). In addition, a copy of the response should be

submitted to the appropriate Regional Administrator.

Reasons for Information Request

NRC regulatory requirements and plant TS requirements prohibit operation with known

through-wall leakage from the RCPB. NRC regulations, plant TSs, and ASME Code

requirements are intended to make licensees perform inspections to maintain an extremely low

probability of abnormal leakage, or rapidly propagating failure, and of gross rupture. The

current inspection program implemented at some PWRs may not permit timely identification of

degradation of Alloy 82/182/600 materials exposed to the pressurizer environment in a manner

consistent with NRC requirements.

This information request is necessary to permit the NRC staff to verify compliance with existing

regulations and plant-specific licensing bases. The information being requested by this bulletin

focuses on Alloy 82/182/600 materials exposed to the pressurizer environment and is

information that is not currently available to the NRC staff. 
BL 2004-xx

Page 10 of 11

The NRC staff will use the information to assess the acceptability of current licensee inspection

programs, and to determine the need for, and guide the development of, any additional

regulatory actions (e.g., generic communications, orders, or rulemaking) to address the integrity

of these RCPB locations. Such regulatory actions could include regulatory requirements for

augmented inspection programs under 10 CFR 50.55a(g)(6)(ii). The NRC staff will review the

responses to this bulletin to determine whether the PWR licensees’ inspections provide

reasonable assurance that existing applicable regulations are met. If concerns are identified,

the NRC staff will contact each affected licensee.

Related Generic Communications

Bulletin 2003-02, “Leakage From Reactor Pressure Vessel Lower Head Penetrations and

Reactor Coolant Pressure Boundary Integrity,” August 21, 2003 (ADAMS Accession No.

ML032320153)

Bulletin 2002-02, “Reactor Pressure Vessel Head and Vessel Head Penetration Nozzle

Inspection Programs,” August 9, 2002 (ADAMS Accession No. ML022200494)

Bulletin 2002-01, "Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure

Boundary Integrity,” March 18, 2002 (ADAMS Accession No. ML020770497)

Bulletin 2001-01, “Circumferential Cracking of Reactor Pressure Vessel Head Penetration

Nozzles,” August 3, 2001 (ADAMS Accession No. ML012080284)

Backfit Discussion

Under the provisions of Section 182a of the Atomic Energy Act of 1954, as amended, and

10 CFR 50.54(f), this bulletin transmits an information request for the purpose of verifying

compliance with existing applicable regulatory requirements (see the Applicable Regulatory

Requirements section of this bulletin). Specifically, the required information will enable the 

NRC staff to determine whether current inspection and maintenance practices for the detection

of degradation of the RCPB at reactor facilities (similar to the degradation observed at

Millstone, Unit 2, Crystal River, Unit 3, Tsuruga, Unit 2, and TMI-1) provide reasonable

assurance that RCPB integrity is being maintained. No backfit is either intended or approved

by the issuance of this bulletin, and the NRC staff has not performed a backfit analysis.

Federal Register Notification

A notice of opportunity for public comment on this bulletin was not published in the

Federal Register because the NRC staff is requesting information from power reactor licensees

on an expedited basis for the purpose of assessing compliance with existing applicable

regulatory requirements and the need for subsequent regulatory action. This bulletin was

prompted by the discovery of leaks at Tsuruga, Unit 2, Millstone, Unit 2, Waterford, Unit 3, and

non-RCPB circumferential cracking at Palo Verde, Unit 2. As the resolution of this matter

progresses, the opportunity for public involvement will be provided. Nevertheless, comments

on the actions requested and the technical issues addressed by this bulletin may be sent to the
2

BL2004-xx

Page 11 of 11

U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC

20555-0001.

Small Business Regulatory Enforcement Fairness Act

The NRC has determined that this action is not subject to the Small Business Regulatory

Enforcement Fairness Act of 1996.

Paperwork Reduction Act Statement

This Bulletin contains information collections that are covered by the Office of Management and

Budget clearance number 3150-0012, which expires August 31, 2006. The burden to the public

for these mandatory information collections is estimated to average 200 hours per response,

including the time for reviewing instructions, searching existing data sources, gathering and

maintaining the data needed, and completing and reviewing the information collection. Send

comments regarding this burden estimate or any other aspect of these information collections,

including suggestions for reducing the burden, to the Records and FOIA/Privacy Services

Branch (T-5 F52), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by

Internet electronic mail to INFOCOLLECTS@NRC.GOV; and to the Desk Officer, Office of

Information and Regulatory Affairs, NEOB-10202, (3150-0012), Office of Management and

Budget, Washington, DC 20503. 

Public Protection Notification

The NRC may not conduct or sponsor, and a person is not required to respond to, an

information collection unless the requesting document displays a currently valid OMB control

number.

If you have any questions about this matter, please contact one of the persons listed below or

the appropriate Office of Nuclear Reactor Regulation project manager. 

Bruce A. Boger, Director

Division of Inspection Program Management

Office of Nuclear Reactor Regulation