June 6, 2008 

MEMORANDUM TO: Eric J. Leeds, Director 

 Office of Nuclear Reactor Regulation 

THRU: John A. Grobe, Associate Director /RA/

 for Engineering and Safety Systems 

 Office of Nuclear Reactor Regulation 

FROM: Michele G. Evans, Director /RA/

 Division of Component Integrity 

 Office of Nuclear Reactor Regulation 

SUBJECT: PRESSURIZER NOZZLE ALLOY 82/182 WELD FLAWS 

 IDENTIFIED IN A RETIRED PRESSURIZER IN 

 FEBRUARY 2008, LIC-504 ASSESSMENT, INTEGRATED 

 RISK-INFORMED DECISION MAKING PROCESS FOR 

 EMERGENT ISSUES

The purpose of this memorandum is to document the decision making process used to 

determine what regulatory action to recommend as the result of new information received 

regarding potential flaws in Alloy 82/182 welds in reactor coolant pressure boundary piping. 

The Office of Nuclear Reactor Regulation issued Office Instruction LIC-504, “Integrated 

Risk-Informed Decision Making Process for Emergent Issues,” Revision 2, on February 12, 

2007. The process described in this document formed the basis for the decision making 

process that the staff used to address the results of the preliminary inspections of dissimilar 

metal welds in the nozzles of a retired pressurizer. These results were obtained from EPRI in 

mid-February and early March 2008. The purpose of this memorandum is to document the 

integrated risk-informed assessment of the options considered in the regulatory decision 

reached. The LIC-504 assessment for this issue is provided in Enclosure 1. The decision 

authority and evaluation team for this issue are listed in Enclosure 2. The timeline of related 

events to support the LIC-504 Assessment is provided in Enclosure 3. 

In mid-February 2008 the NRC staff received the results of inspections of the nozzles of a 

retired pressurizer (ML080701017). This pressurizer was removed from service and was 

donated to NRC for research purposes and the inspections were performed to determine the 

research value of the nozzle welds. These inspections found indications using both dye 

penetrant testing (PT) and manual phased array ultrasonic examination (UT). The nozzle welds 

of most interest were the three safety nozzles. The inspection concluded that under normal field 

NDE conditions, these three welds would be reported as containing 360° 

circumferentially-oriented linear planar flaws of significant through-wall depth. 

CONTACT: Ted Sullivan, NRR/DCI 

 (301) 415-2796
E. Leeds - 2 - 

The deepest indications found were sized at 89% through-wall on the ‘A’ safety nozzle. The 

deepest indication found in the ‘B’ and ‘C’ safety nozzles were 75% and 69% through-wall, 

respectively. Based on this information NRC staff determined that the inspection results needed 

to be evaluated against the advanced finite element analysis (FEA) work the staff completed in 

September 2007, since the advanced FEA formed the basis for the continued operation of 9 

plants with pressurizer welds that had not yet been inspected as of the end of 2007, as 

mandated by industry guidelines. To help perform such an evaluation, the NRC staff requested 

that EPRI estimate the flaw profile for safety nozzle ‘A’ and provide some of the raw UT signals 

recorded during the inspection. EPRI provided this information to the NRC staff by letter dated 

March 4, 2008 (ML080701020). EPRI estimated that the ‘A’ safety nozzle weld contained a 

continuous deep indication 360° around the circumference. This reported flaw profile was more 

severe than any predicted flaws in the above-referenced advanced finite element analyses that 

would have led to leakage that would be detectable with sufficient time for plant shutdown prior 

to rupture. This information led the NRC staff to question whether the advanced finite element 

analyses would still support the spring 2008 pressurizer inspection schedules (ML072430836). 

In making a regulatory decision to address the retired pressurizer nozzle weld inspection 

results, the NRC staff considered three options. Option 1, the base case, would result in no 

change to existing regulatory and industry programs; i.e., the affected plants would operate until 

their scheduled spring 2008 outage and inspect or mitigate the affected welds at that time. 

Option 2 would allow continued operation of the plants for a short time period while NRC staff 

gathered additional information. Option 3 would require all affected plants to shutdown 

immediately and not restart until the basis for operation until the spring outages was 

reestablished or until inspection or mitigation activities were completed. 

The NRC staff based its regulatory decision on an assessment of the pros and cons of these 

options using the principles of risk-informed decision making. 

The NRC staff concluded that Option 2 was the appropriate decision; the staff judged that it had 

an appropriate basis to take a short period of time (within a week) to gather information to make 

a more informed decision. The initial inspection results were somewhat uncertain given the type 

of inspection that was performed. More refined inspection was judged to be prudent to reduce 

some of this initial uncertainty. Nevertheless, the NRC staff determined that the questions raised 

by the March 4, 2008, EPRI letter were safety significant questions and the staff put industry on 

notice that it was considering regulatory action. A discussion of the actions taken to gather 

additional information and the conclusions of the staff’s evaluation of the additional information 

is contained in a March 14, 2008 memorandum from Michele Evans to Catherine Haney 

(ML080740419). 

Relative to Option 1 the NRC staff concluded that for the plants to continue to operate until their 

spring 2008 outages, the NRC staff needed additional information to determine whether the 

advanced finite element analyses (FEA) continued to support continued operation. Without 

such information, the NRC staff would have lacked the requisite reasonable assurance that the 

health and safety of the public would be protected by continued operation of these plants until 

their spring 2008 outages. 

The proposed actions for Option 3 would provide the requisite reasonable assurance that the 

health and safety of the public would be protected upon restart of the affected plants. 
E. Leeds - 3 - 

However, the NRC staff judged that (1) additional information relevant to the decision could be 

obtained in a very short time period and (2) continued operation for a very short time frame 

would not be inimical to public health and safety. 

For a more complete understanding of the decision making process used to address the retired 

pressurizer nozzle weld inspection results, please refer to the enclosed LIC-504 assessment. 

Enclosures: 

1. LIC-504 Assessment on the Retired Pressurizer Nozzle Weld Flaws 

2. LIC-504 Decision Authority and Evaluation Team 

3. Timeline of Events Leading up to the LIC-504 Recommended Decision 
E. Leeds - 3 - 

However, the NRC staff judged that (1) additional information relevant to the decision could be 

obtained in a very short time period and (2) continued operation for a very short time frame 

would not be inimical to public health and safety. 

For a more complete understanding of the decision making process used to address the retired 

pressurizer nozzle weld inspection results, please refer to the enclosed LIC-504 assessment. 

Enclosures: 

1. LIC-504 Assessment on the Retired Pressurizer Nozzle Weld Flaws 

2. LIC-504 Decision Authority and Evaluation Team 

3. Timeline of Events Leading up to the LIC-504 Recommended Decision 

DISTRIBUTION: DCI r/f TChan DTerao JCollins JUhle 

 ACsontos SLaur MScott RHardies MMayfield 

ADAMS ACCESSION NUMBER: ML081580560

 OFFICE NRR/ADES/DCI NRR/ADES/DCI NRR/ADES/DRA 

 NAME E. Sullivan /RA/ J. Lubinski /RA/ MGE for M. Cunningham /RA/ 

DATE 6/3/08 6/6/08 6/3/08

OFFICE NRR/ADES/DCI NRR/ADES 

NAME M. Evans /RA/ J. Grobe /RA/ 

DATE 6/6/08 6/6/08 

 OFFICIAL RECORD COPY 
ENCLOSURE 

PRESSURIZER NOZZLE ALLOY 82/182 WELD FLAWS IDENTIFIED IN A RETIRED 

PRESSURIZER IN FEBRUARY 2008 

LIC-504 ASSESSMENT, INTEGRATED RISK-INFORMED DECISION MAKING PROCESS 

FOR EMERGENT ISSUES 

1. PURPOSE 

The context of this assessment is the safety issues related to nozzle weld indications identified 

in the dissimilar metal butt welds in a retired pressurizer. The indications were identified during 

inspections performed in February 2008. This assessment discusses the options considered in 

addressing the associated safety issues and the assessment of those options as part of the 

decision making process used to determine what regulatory actions to take to address the 

issues. 

2. BACKGROUND 

On October 13, 2006, the Wolf Creek Nuclear Operating Corporation performed pre-weld 

overlay inspections using ultrasonic testing (UT) techniques on the surge, spray, relief, and 

safety nozzle-to-safe end dissimilar metal and safe end-to-pipe stainless steel butt welds. The 

inspection identified five circumferential indications in the surge, relief, and safety nozzle-to-safe 

end dissimilar metal (DM) butt welds that the licensee attributed to primary water stress 

corrosion cracking (PWSCC) (ML063380456) and that were significantly larger and more 

extensive than previously seen in the industry. 

During public meetings with the industry on November 30, 2006, and December 20, 2006, the 

Nuclear Regulatory Commission (NRC) staff presented the results of a fracture mechanics 

based scoping study that assessed the safety significance of the UT indications found at Wolf 

Creek. As a result of these analyses, the staff concluded that there may be little or no time 

margin between the onset of leakage and rupture in pressurizer nozzle DM butt welds 

containing flaws similar to those found at Wolf Creek. 

In March 2007, the NRC issued Confirmatory Action Letters (CALs) to 40 nuclear power plant 

licensees with pressurized water reactors (PWR), confirming commitments from those licensees 

to resolve concerns regarding potential flaws in specific reactor coolant system (RCS) DM butt 

welds by the end of 2007. The remaining 29 PWR plants had either completed the requisite 

actions or do not have pressurizer welds susceptible to PWSCC. 

Nine of the plants receiving CALs did not have outages scheduled in 2007. These plants 

committed to accelerate outages into 2007 if the industry was not able to demonstrate an 

adequate level of safety to the NRC. The nine plants are Braidwood 2, Comanche Peak 2, 

Diablo Canyon 2, Palo Verde 2, Seabrook, South Texas Project 1, V. C. Summer, Vogtle 1, and 

Waterford 3. 

By letter dated February 14, 2007, the Nuclear Energy Institute indicated that the Electric Power 

Research Institute (EPRI) Materials Reliability Program (MRP) would be undertaking a task to 

refine the crack growth analyses pertaining to the Wolf Creek pressurizer DM weld ultrasonic 

indications. These additional advanced finite element analyses (FEA) were performed to 

address the NRC staff’s concerns regarding the potential for rupture without prior evidence of 

leakage from circumferentially-oriented PWSCC in pressurizer nozzle welds. The goal of these 

studies was to demonstrate that PWSCC in pressurizer DM butt welds would progress 
- 2 - 

through-wall and exhibit detectable leakage prior to causing a possible rupture event. These 

studies reduced unnecessary conservatisms and some of the uncertainties in previous 

analyses. 

Industry completed these analyses and documented the results in MRP-216, Revision 1, 

“Advanced FEA Evaluation of Growth of Postulated Circumferential PWSCC Flaws in 

Pressurizer Nozzle Dissimilar Metal Welds: Evaluations Specific to Nine Subject Plants” 

(ML072410235). These results were provided to the NRC staff by letter dated August 13, 2007. 

The NRC staff completed independent analyses to enable the staff to perform an in-depth 

review and critique of the industry’s analyses and to extend the scope of the industry’s 

analyses. The NRC staff documented its safety assessment on the industry’s analyses in an 

enclosure to a memorandum from Michele Evans to Catherine Haney on September 7, 2007 

[ML072430836]. The staff’s assessment used advanced finite element analyses of the 

fabrication, loading, and postulated flaw growth in the pressurizer nozzle welds to assess crack 

growth rates and shapes based on an array of starting flaw sizes. It was concluded that 

PWSCC in pressurizer DM butt welds of the nine plants analyzed would progress through-wall 

and exhibit detectable leakage prior to causing a possible rupture event. Therefore, the 

conclusion of the NRC staff’s safety assessment was that there is reasonable assurance that 

the nine plants addressed by the evaluation can operate safely until their next scheduled 

refueling outages in the spring of 2008. 

In mid-February 2008, the NRC staff received the results of inspections of the nozzles of a 

retired pressurizer (ML080701017). This pressurizer was removed from service to eliminate the 

possibility of extended plant outages resulting from cracking associated with the heater sleeves. 

The pressurizer was donated to NRC for research purposes and the inspections were 

performed to determine the research value of the nozzle welds. These inspections found 

indications by dye penetrant (PT) and manual-phased array ultrasonic examination (UT). 

Circumferential and axial indications were found in five of six nozzles. The nozzle welds of most 

interest were the three safety nozzles. The inspection concluded that these nozzles had 360°, 

circumferentially-oriented indications with non-uniform depths around circumference. The 

deepest indications found were sized at 89% through-wall on the ‘A’ safety nozzle. The deepest 

indication found in the ‘B’ and ‘C’ safety nozzles were 75% and 69% through-wall, respectively. 

Based on this information NRC staff determined that the inspection results needed to be 

evaluated against the advanced FEA work the staff completed in September 2007 since the 

advanced FEA element analyses formed the basis for the continued operation of 9 plants with 

pressurizer welds that had not yet been inspected, as mandated by industry guidelines. To help 

perform such an evaluation, the NRC staff requested that EPRI estimate the flaw profile for 

safety nozzle ‘A’ and provide some of the raw UT signals recorded during the inspection. EPRI 

provided this information to the NRC staff by letter MRP 2008-014, dated March 4, 2008 

(ML080670004). EPRI estimated that the ‘A’ safety nozzle weld contained a continuous deep 

indication 360° around the circumference. This reported flaw profile was more severe than any 

of the predicted flaws in the above-referenced advanced finite element analyses that led to 

leakage that would be detected with sufficient time for plant shutdown prior to rupture. The flaw 

profile caused NRC staff to question whether the advanced finite element analyses would still 

support the spring 2008 pressurizer inspection schedules (ML072430836). 
- 3 - 

Degree of conservatism in the regulatory analysis of this issue

The industry program for inspecting and mitigating welds subject to this PWSCC was developed 

based on the information available prior to the Wolf Creek inspection findings. This program is 

described in “Materials Reliability Program (MRP): Primary System Piping Butt Weld Inspection 

and Evaluation Guidelines,” dated July 2005 and known as MRP-139. The information available 

at the time MRP-139 was issued indicated to the NRC staff that there was a serious safety issue 

with PWSCC in dissimilar metal butt welds. Based on operating experience, the NRC staff 

believed the industry baseline inspection schedule for pressurizer nozzle welds under MRP-139 

was generally adequate, if completed according to the schedule outlined in MRP-139. Based 

on the results of advanced finite element analyses of pressurizer nozzle welds performed after 

the discovery of the large circumferential indications at Wolf Creek, the NRC staff concluded 

that it was acceptable for nine plants to continue to operate until their spring 2008 outages to 

perform the pressurizer weld inspections rather than shutdown by December 31, 2007, 

consistent with MRP-139. 

This decision was based on extensive flaw growth analyses for the pressurizer nozzle welds 

and the enhanced leakage monitoring and shutdown criteria adopted by these plants. 

Performing these analyses required nozzle-specific information of the weld configuration and 

fabrication history, the applied loading, the material properties, crack growth rates for the 

specific weld materials for each of the 9 plants, and consideration of possible initial flaw shapes, 

sizes, and locations. As discussed in the NRC staff safety assessment (ML072400199), there 

are potential conservatisms and non-conservatisms in the analysis that are difficult to quantify. 

Therefore, the NRC staff extended the industry’s analyses by considering additional sensitivity 

cases and applying higher safety factors to evaluate the acceptability of the results. The NRC 

staff concluded, based on these analyses and the enhanced leakage monitoring, that if a 

PWSCC flaw were to initiate and grow in one of the pressurizer nozzle DM welds before the 

Spring 2008 refueling outages, there is reasonable assurance that leakage would occur and be 

detected such that adequate time would be available to safely shutdown the plant prior to 

rupture at the nozzle weld. 

3. EVALUATION AND ASSESSMENT OF REGULATORY OPTIONS CONSIDERED BY THE 

STAFF AGAINST THE FIVE KEY PRINCIPLES OF RISK INFORMED REGULATION 

The following three options were developed based on the estimated severity of the retired 

pressurizer flaw in safety nozzle ‘A’ and the associated concerns discussed previously. 

Option Description of Regulatory Option 

1 Base Case – No change to existing regulatory and industry program 

2 Continued operation of the plants for a short time period while NRC staff 

gathers additional information 

3 Issue orders for immediate plant shutdown 

As discussed in the following sections, these options were assessed so that a recommendation 

could be reached based on the information available on March 7, 2008. 

3.1 Option 1 - Base Case: No changes in the existing regulatory and industry programs or 

inspection/shutdown schedule. The eight plants1

 still operating with uninspected pressurizer 

 

1 One of the nine plants had already shutdown to perform weld overlays by the time the March 4, 2008 EPRI report 

was received, leaving eight plants involved in this decision. 
- 4 - 

nozzle welds proceed on their present schedule to shutdown and mitigate/inspect these welds 

during their planned spring 2008 outages based on the results of the advanced finite element 

analyses completed in August 2007 

3.1.1 Compliance with Regulations

This issue involves the potential reduction in the capability in the reactor coolant system 

pressure boundary at the cited plants. The reactor coolant system pressure boundary is one of 

the principal fission product barriers in the general design criteria and is integral to defense-indepth of the plant. 

The indications discovered in the nozzle welds of the retired pressurizer were considered to be 

potentially due to PWSCC. If the flaw profile in the March 4, 2008 EPRI report was assumed to 

be due to PWSCC, the weld may have had sufficient ASME Code, Section XI, margin on 

structural integrity under design basis loadings at the time the retired pressurizer was taken out 

of service. Regardless, if this flaw was in an operating pressurizer and was due to PWSCC, it 

could not have been left in service because of the potential high growth rate of PWSCC, and the 

repair/replacement criteria specified in ASME Code, Section XI. Therefore, such a degraded 

condition would have required corrective actions in accordance with the ASME Code, which is 

incorporated by reference into NRC regulations by 10 CFR 50.55a. 

Additionally, as discussed above, the previous decision to allow continued operation until Spring 

2008 was based, in part, on the advanced FEA analyses. However, the flaw profile reported on 

March 4th was more severe than any predicted flaws in the above-referenced advanced FEA 

that led to leakage that would be detected with sufficient time for plant shutdown prior to rupture. 

Therefore, based on the information available on March 7, 2008, the NRC staff questioned the 

continued applicability of the advanced FEA analyses if any of the plants with uninspected 

pressurizer welds had PWSCC of a similar, or greater, extent than the reported flaw profile . 

3.1.2 Defense-in-Depth

This issue potentially impacts the integrity of the reactor coolant system pressure boundary, 

which is one of the principal fission product barriers. The effect is a potential increase in the 

probability of failure for that boundary, not an assured loss of the boundary. This issue may 

increase the frequency of a loss of coolant accident (LOCA) in the size range requiring 

emergency core cooling system (ECCS) operation in the recirculation mode.2

 The other two 

principal fission product barriers, the fuel cladding and containment, are not impacted. 

The decision in September 2007 that it was acceptable for nine plants to continue to operate 

until their spring 2008 was based, in part, upon reliance on plant leakage monitoring programs. 

This option would continue to require a high degree of reliance on plant leakage monitoring 

programs. The questions raised by the flaw profile in the March 4, 2008 EPRI report potentially 

impacted the ability to rely on leakage monitoring as a suitable compensatory action. 

 

2 While the decision makers were aware that some of the eight plants may still have been working to resolve the 

containment slump clogging issue of Generic Safety Issue (GSI)-191, the compounding affect of GSI-191 on LOCA 

frequency was not explicitly considered in the decision making because the decision involved allowing the plants to 

operate only for one additional week to gather information.
- 5 - 

3.1.3 Safety Margins

Assuming the flaws in the retired pressurizer were due to PWSCC, the safety margin impacted 

is the predicted time between leakage and failure from the advanced FEA. For any weld 

affected by PWSCC of the size shown on the March 4th EPRI flaw profile, the potential reduction 

in safety margin could be significant, because a pipe failure would result in a loss of coolant 

accident (LOCA), up to and including a large break LOCA. Failures of the subject butt welds 

due to large circumferential cracks would produce medium or large LOCAs at the pressurizer. 

PRAs typically indicate conditional core damage probabilities (CCDPs) for these types of 

LOCAs are in the 10-3 to 10-2 range. 

This option would allow the industry to continue operating until the spring 2008 outages based 

on the conclusions of the advanced finite element analyses. During the spring 2008 outages 

the pressurizer welds are scheduled to be inspected and mitigated. As noted above, the NRC 

staff had questions concerning the continued applicability of the advanced finite element 

analyses if any of the plants with uninspected pressurizer welds had PWSCC of a similar, or 

greater, extent than the reported flaw profile. This option would have allowed for a potential 

reduction in safety margin for up to approximately seven weeks after the staff received the letter 

from EPRI dated March 4, 2008. 

3.1.4 Risk Assessment

EPRI provided a letter report, “Implications Of Wolf Creek Pressurizer Butt Weld Indications 

Relative to Safety Assessment and Inspection Requirements,” MRP 2007-003, dated January 

2007 [ML070240159]. The NRC staff reviewed this report and documented its conclusions in 

Section 3.1.4 of the Wolf Creek LIC-504 assessment [ML070990071]. The staff conclusion in 

this review that the available data and current lack of understanding of the underlying physics of 

the degradation mechanism made any estimates of the risk insufficient to support a regulatory 

decision. 

EPRI subsequently updated its probabilistic study with the report, “Evaluation of Pressurizer 

Alloy 82/182 Nozzle Failure Probability,” provided in Appendix E to MRP-216, Revision 1, 

“Advanced FEA Evaluation of Growth of Postulated Circumferential PWSCC Flaws in 

Pressurizer Nozzle Dissimilar Metal: Evaluation Specific to Nine Subject Plants,” EPRI, Palo 

Alto, CA, August 2007 [ML072410235]. The staff reviewed the updated probabilistic study and, 

for reasons stated in its advanced FEA safety assessment [ML072430836], found it difficult to 

develop a basis for accepting the probability of failure estimates in Appendix E and, therefore, 

based its conclusions on its evaluation of the deterministic analyses and results.3

3.1.5 Performance Measurement

For Option 1, the performance measurement strategies or compensatory measures are the 

enhanced leakage monitoring programs and shutdown criteria in effect at the subject plants. 

However, the results of the inspection of the nozzle welds in the retired pressurizer led staff to 

question the continued applicability of the advanced finite element analyses. These analyses 

demonstrated that PWSCC in pressurizer DM butt welds of the nine plants analyzed would 

 

3 While not explicitly considered in the decision making, this issue does potentially impact the risk from seismic 

events, because a degraded pipe will be more likely to fail during an earthquake than a non-degraded pipe. 

However, given the low probability of a seismic event in the timeframe for Option 1, the risk contribution from seismic 

events is judged to be very low.
- 6 - 

progress through-wall and exhibit detectable leakage prior to causing a possible rupture event. 

Therefore, the staff questions about the applicability of these analyses also calls into question 

the viability of the enhanced leakage monitoring to serve as a compensatory measure. 

3.2 Option 2 - Continued operation of the plants for a short time period while NRC staff gathers 

additional information 

3.2.1 Compliance with Regulations

The NRC staff reached the judgment that the questions raised by the March 4, 2008 EPRI letter 

were safety-significant and the staff communicated to the industry that regulatory action was 

being considered. Based on the information available to the staff from the March 4th letter, the 

staff considered that it would be appropriate to take up to one week to gather additional 

information before taking regulatory actions that may have led to shutdown of the plants. While 

considering the significance of the data provided by the EPRI letters, one week is a relatively 

short time to allow continued operation, given typical crack propagation rates and prior 

operating history. Therefore, it was judged that this small delay in taking action would not be 

inimical to public health and safety. In addition, the staff believed that conclusive information 

could be developed over this short period of time. The following paragraphs discuss why the 

NRC staff did not have sufficient information to determine whether the advanced FEA continued 

to be applicable and why the NRC staff judged that information from the EPRI letters was not 

sufficiently conclusive as to warrant immediate plant shutdown. 

(1) The February 15, 2008 EPRI report was not conclusive regarding whether or not the UT 

indications recorded on the three safety nozzles were indicative of PWSCC or preexisting 

fabrication defects. The report noted that performing additional automated UT on these three 

nozzle welds would allow for more accurate flaw characterization. One area of uncertainty was 

that the report indicated that UT sound beams that were projected normal to the surface 

(straight beams) did not pick up the indications. This would lead to the conclusion that the 

indications were service-induced. However, the staff was aware that while UT techniques can 

be used to detect construction flaws, no information was provided on the capability or 

qualifications of the straight beam inspection to detect construction flaws. Another area of 

uncertainty was that the results of the PT examinations of the inside surface, where PWSCC 

would initiate, showed a small number of short linear indications, but the PT results did not 

approach the circumferential extent of the UT indications. Although PT of PWSCC may have 

been affected by surface grinding, the PT results were inconsistent with the flaw profile shown in 

the March 4, 2008 EPRI report. 

(2) The NRC staff and industry had performed ten of thousands of hours of advanced analyses 

to understand flaw growth in the pressurizer nozzles of the plants still operating. These 

analyses were based on actual fabrication histories that covered a range of practices. A flaw 

similar to the profile shown in the March 4, 2008 EPRI report was contrary to the results of the 

NRC staff’s analytical flaw growth studies and would not have been expected. Based on the 

information available to the staff about Combustion Engineering pressurizer units, the staff could 

not postulate a set of conditions that could develop the reported flaw profile in the retired 

pressurizer safety nozzle weld from typical configurations, fabrication processes, fabrication 

flaws, fabrication repairs, and operating loads. 

For example, the staff was aware that the Combustion Engineering (CE) pressurizer units 

studied had high margins between leak and rupture based on the known fabrication processes. 

The CE processes resulted in low residual stresses on the inside surface of the pipe and 
- 7 - 

postulated flaws either arrested or led to large margins between leakage and rupture. St. Lucie 

is a Combustion Engineering design but was not one of the plants that were specifically 

analyzed. Therefore, the staff did not have plant specific information on St. Lucie’s pressurizer 

nozzle configuration, fabrication, or nozzle loadings or any other information that would provide 

insights on why the advanced FEA would no longer be applicable. Option 2 would allow time for 

industry to provide information on why the original FEA would or would not continue to be 

applicable in light of the reported, retired pressurizer nozzle weld flaw profile. 

(3) The flaw profile in the March 4, 2008 EPRI report was not consistent with prior operating 

experience. The staff is aware of the PWSCC identified in butt welds in domestic as well as 

foreign pressurized water reactors. Long flaws have been previously identified but no flaws 

have been reported approaching the depth of the circumferential 360° inside surface connected 

flaws found in the retired pressurizer safety nozzle welds. 

3.2.2 Defense-in-Depth

The assessment of defense-in-depth for Option 1 in Section 3.1.2 is unchanged for Option 2. 

3.2.3 Safety Margins

This option would allow the affected licensees to continue operating for up to one week while 

additional information was gathered. The NRC staff needed additional information to determine 

whether the advanced FEA continued to support the original decision for allowing the plants to 

operate until Spring of 2008. Therefore, this option would have allowed for a potential reduction 

in safety margin for up to one week after March 7, 2008. As discussed in Section 3.2.1, the staff 

believed that it would be appropriate to take this time to gather additional information before 

taking regulatory actions in order to address questions and reduce uncertainties. 

3.2.4 Risk Assessment

As discussed in Section 3.1.4, the available risk estimates associated with PWSCC in DMBWs 

are insufficient to support a regulatory decision. 

3.2.5 Performance Measurement

The assessment of performance measurement for Option 1 in Section 3.1.5 is unchanged for 

Option 2. 

3.3 Option 3 - Issue orders for immediate plant shutdown 

3.3.1 Compliance with Regulations

The proposed actions for Option 3 would provide the requisite reasonable assurance that the 

health and safety of the public would be protected upon restart of the affected plants. However, 

the staff needed additional information to determine whether the advanced FEA continued to 

support continued operation of the eight remaining plants with spring 2008 outages. Further, 

the NRC staff judged that (1) additional information relevant to the decision could be obtained in 

a short time period and (2) continued operation for a short time period would not be inimical to 

public health and safety. The NRC staff judged that information from the March 4th EPRI letter 

raised safety questions but was not sufficiently conclusive as to warrant immediate plant 

shutdown. 
- 8 - 

3.3.2 Defense-in-Depth

This option would remove the contribution of potentially degraded welds to concerns about 

defense-in-depth. 

3.3.3 Safety Margins

Option 3 would promptly provide assurance that the intended safety margins are restored to at 

least the same level as that prior to discovery of the potentially adverse condition of the safety 

nozzle welds in the retired pressurizer. These eight plants would not have restarted until the 

questions about the continued applicability of the advanced FEA were answered or until the 

welds were either inspected or mitigated. For plants that would inspect their nozzles but not 

mitigate the welds, safety margin would be adequate because the required inspections and 

reinspection frequency would provide reasonable assurance that PWSCC would not 

compromise the reactor coolant pressure boundary. For plants that would mitigate the welds, 

the safety margin would be restored to a level similar to that of plants when first placed into 

operation. 

3.3.4 Risk Assessment

As discussed in Section 3.1.4, the available risk estimates associated with PWSCC in DMBWs 

are insufficient to support a regulatory decision. 

3.3.5 Performance Measurement

For Option 3 no performance measurement strategies or compensatory measures are proposed 

or necessary since the option requires prompt shutdown of the affected plants to either inspect 

or mitigate the affected welds. 

4. CONCLUSIONS

The indications discovered in the nozzle welds of the retired pressurizer were considered to be 

potentially due to PWSCC. If the flaw profile in the March 4, 2008 EPRI report was assumed to 

be due to PWSCC, the weld may have had sufficient ASME Code, Section XI, margin on 

structural integrity under design basis loadings at the time the retired pressurizer was taken out 

of service. Regardless, if this flaw were in an operating pressurizer and were due to PWSCC, it 

could not have been left in service because of the potential high growth rate of PWSCC, and the 

repair/replacement criteria specified in ASME Code, Section XI. Therefore, such a degraded 

condition would have required corrective actions in accordance with the ASME Code, which is 

incorporated by reference into NRC regulations by 10 CFR 50.55a. 

The previous decision to allow continued operation until Spring 2008 was based, in part, on the 

advanced FEA analyses. However, the flaw profile reported on March 4, 2008 was more severe 

than any predicted flaws in the above-referenced advanced FEA that led to leakage with 

sufficient time for plant shutdown prior to rupture. Therefore, based on the information available 

on March 7, 2008, the NRC staff questioned the continued applicability of the advanced FEA 

analyses if any of the plants with uninspected pressurizer welds had PWSCC of a similar, or 

greater, extent than the reported flaw profile. 

On March 7, 2008 when a regulatory decision was made to allow the eight plants to continue to 

operate, at least for a short period of time, staff involved in the decision making considered 
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three options. The NRC staff performed an assessment of the three regulatory options against 

the principles of risk-informed decision making. The NRC staff recommended Option 2 - 

continued operation of the plants for a short time period while NRC staff gathers additional 

information. 

At that time the NRC staff was aware of the risk assessment previously performed in connection 

with the Wolf Creek LIC-504 assessment. The NRC staff had prepared an evaluation of a risk 

assessment performed by industry and concluded that available data and current lack of 

understanding of the underlying physics of the degradation mechanism made any quantitative 

estimates of the risk insufficient to support a regulatory decision. Therefore, the NRC staff’s 

assessment of the options in this case was based primarily on the other principles. 

This decision was based on the judgment that taking up to one week to gather additional 

information gave appropriate consideration to each of the following factors: 

(1) the need to quickly restore confidence in the safety margins of the eight 

 remaining plants potentially impacted by the inspection results. 

(2) for the plants to continue to operate until their spring 2008 outages, the NRC staff 

needed additional information to determine whether the advanced FEA continued to 

support operation until the outage. 

(3) the NRC staff believed that conclusive information could be developed in a short 

period of time. 

(4) the NRC staff judged that information from the March 4, 2008 EPRI letter raised 

safety questions but was not sufficiently conclusive as to warrant immediate plant 

shutdown. 

Factors (1) and (2) led the staff to reject Option 1. The staff judged that a one week information 

gathering period is short enough to maintain public health and safety, given typical crack 

propagation rates and prior operating history. Option 2 was supported by factor (3). Factor (4) 

led the staff to reject Option 3. 

 
 

ENCLOSURE 2 

LIC-504 Evaluation Team

Decision Authority 

 James Wiggins, Acting Director, Office of Nuclear Reactor Regulation (NRR) 

Evaluation Team Members 

NRR 

 Jack Grobe, Associate Director for Engineering and Safety Systems 

 Michele Evans, Director, Division of Component Integrity (DCI), NRR 

 Ted Sullivan, Senior Level Advisor, DCI, NRR 

 Jay Collins, Acting Chief, Piping and Non-Destructive Evaluation Branch (CPNB), 

RES 

 Jennifer Uhle, Director, Division of Engineering (DE), RES 

 Robert Tregoning, Senior Level Advisor, DE, RES 
 

ENCLOSURE 3 

Timeline of Related Events to Support LIC-504 Assessment of Pressurizer Nozzle 

Alloy 82/182 Weld Flaws Identified in a Retired Pressurizer in February 2008

October 19, 2005: 

• St. Lucie Unit 1 pressurizer removed from service and shipped to a decontamination and 

reduction facility. 

October 13, 2006: 

• Wolf Creek discovered 5 circumferential indications in the pressurizer relief, safety, and 

surge nozzles. 

November 30, 2006: 

• NRC staff presented the results of a fracture mechanics based scoping study at a public 

meeting that assessed the safety significance of the Wolf Creek indications. 

December 20, 2006: 

• As a result of the NRC analyses, the staff concluded that there may be little or no time 

margin between the onset of leakage and rupture in pressurizer nozzle DM butt welds 

containing flaws similar to those found at Wolf Creek.

February 14, 2007: 

• Nuclear Energy Institute (NEI) indicates that the Electric Power Research Institute 

(EPRI) Materials Reliability Program (MRP) would refine the crack growth analyses 

pertaining to the Wolf Creek pressurizer dissimilar metal butt weld (DMBW) indications 

to address NRC staff concerns.

March 2007: 

• NRC issued Confirmatory Action Letters to 40 PWRs, confirming commitments from 

those licensees to address concerns regarding potential flaws in pressurizer nozzle 

DMBWs by 12/31/07. Of the 40 PWRs receiving CALs, 9 plants did not have scheduled 

outages in 2007, but, committed to enhanced leakage monitoring and to accelerate their 

outage schedule into 2007 if the industry was not able to demonstrate an adequate level 

of safety to the NRC. The nine plants were Braidwood 2, Comanche Peak 2, Diablo 

Canyon 2, Palo Verde 2, Seabrook, South Texas Project 1, V. C. Summer, Vogtle 1, and 

Waterford 3. 

March-August 2007: 

• Advanced Finite Element Analysis (AFEA) Project undertaken to address NRC staff’s 

concerns regarding the potential for rupture without prior evidence of leakage from 

circumferentially oriented primary water stress corrosion cracks in pressurizer nozzle 

welds.
- 2 - 

Spring-Summer 2007: 

• NRC Office of Research (RES) learns of the availability of the retired St. Lucie Unit 1 

pressurizer and takes the opportunity to acquire the DMBW nozzles to potentially 

address several Office of Nuclear Reactor Regulations (NRR) user need requests. RES 

staff negotiates ownership and disposal issues with Florida Power and Light Company. 

August 13, 2007: 

• Industry completed their AFEA analyses and documented the results in MRP-216, 

Revision 1, “Advanced FEA Evaluation of Growth of Postulated Circumferential PWSCC 

Flaws in Pressurizer Nozzle Dissimilar Metal Welds: Evaluations Specific to Nine 

Subject Plants.”

September 7, 2007: 

• NRC staff’s confirmatory AFEA analyses completed and decision to allow operation of 

the 9 plants until June 2008 is documented in the AFEA Safety Assessment. 

September 28, 2007: 

• RES initiated a commercial contract to section the St. Lucie pressurizer top (spray, 

safety, and relief) and bottom head (surge) nozzles from the main pressurizer housing. 

February 4-8, 2008: 

• Final surface preparation performed and initial surface and volumetric NDE exams 

performed. Exams included liquid dye penetrant inspection of both the outside and 

inside surfaces of each DMW and manual phased-array ultrasonic examination of the 

volume of each DMW. 

February 15, 2008: 

• NRC staff received the Materials Reliability Program (MRP) Letter 2008-012 with 

attached report entitled “St. Lucie Pressurizer Nozzle DM Weld Examination Project.” 

Report documents the results of the initial PDI-qualified manual ultrasonic tests. Staff 

informs management of the initial findings. 

February 19, 2008: 

• Staff sets up a conference call with cognizant EPRI staff members for Wednesday 

February 27, 2008 to discuss results and path forward. 

• Staff informs management that results of Letter 2008-012 need to be evaluated against 

the AFEA work the staff completed in September 2007. 
- 3 - 

February 27: 

• Staff discusses the preliminary St. Lucie NDE results and path forward with EPRI staff 

during the scheduled teleconference. EPRI representatives indicate that the NDE 

indications are deeper on one side and shallower on the opposite side. Staff requests a 

profile of the indication be provided as soon as possible and compares the qualitative 

flaw description to the previous AFEA results. 

March 4: 

• EPRI Report (Letter 2008-014) provided to NRC showing manual UT’s estimated flaw 

profile. The profile in this report depicts a single indication with significant through-wall 

depth at 19 points spaced somewhat evenly around the circumference of the “A” safety 

nozzle. This profile depicts substantially more cracked area than originally anticipated. 

March 5: 

• Briefings and meetings 

o Staff notes that flaw profile from manual UT may not be bound by AFEA results 

used to support the decision to allow the 9 plants to operate beyond 12/31/07. 

o Staff discusses whether industry should conduct an accelerated, more accurate 

NDE inspection of St. Lucie nozzles. 

March 6: 

• Management share views regarding concerns that NRC may not have an adequate 

basis for allowing the 9 plants to operate without additional assurances given the 

available information concerning the St. Lucie reported indication. 

• Management agrees to promptly revisit the decision to allow continued operation of 9 

plants and brief OEDO of the conclusions of those discussions. 

March 7: 

• Meeting involving NRR and RES management and staffs to determine actions going 

forward 

o Decision made to escalate the issue on the basis that a question was raised 

regarding continued applicability of the Advanced Finite Element Analysis 

(AFEA) results. 

o Decision made to recommend continued operation of the plants for a short time 

period while NRC staff requests that licensees provide information to address the 

basis for continued operation of the 9 plants. 

• Briefing of DEDO on recommendation. DEDO agrees with recommendation. 

• Chief Nuclear Operations officers of all 8 plants were informed of NRC staff concerns 

and the need for information to address the basis for continued operation, and were put 

on notice that the staff was considering regulatory action.