a

GAO

United States General Accounting Office

Report to Congressional Requesters

May 2004 NUCLEAR 

REGULATION

NRC Needs to More 

Aggressively and 

Comprehensively 

Resolve Issues Related 

to the Davis-Besse 

Nuclear Power Plant’s 

Shutdown

GAO-04-415
www.gao.gov/cgi-bin/getrpt?GAO-04-415.

To view the full product, including the scope 

and methodology, click on the link above. 

For more information, contact Jim Wells at 

(202) 512-3841 or wellsj@gao.gov. 

Highlights of GAO-04-415, a report to 

congressional requesters 

May 2004

NUCLEAR REGULATION 

NRC Needs to More Aggressively and 

Comprehensively Resolve Issues Related 

to the Davis-Besse Nuclear Power Plant’s 

Shutdown 

NRC should have but did not identify or prevent the corrosion at DavisBesse because its oversight did not generate accurate information on plant 

conditions. NRC inspectors were aware of indications of leaking tubes and 

corrosion; however, the inspectors did not recognize the indications’ 

importance and did not fully communicate information about them. NRC 

also considered FirstEnergy—Davis-Besse’s owner—a good performer, 

which resulted in fewer NRC inspections and questions about plant 

conditions. NRC was aware of the potential for cracked tubes and corrosion 

at plants like Davis-Besse but did not view them as an immediate concern. 

Thus, NRC did not modify its inspections to identify these conditions. 

NRC’s process for deciding to allow Davis-Besse to delay its shutdown lacks 

credibility. Because NRC had no guidance specifically for making a decision 

on whether a plant should shut down, it used guidance for deciding whether 

a plant should be allowed to modify its operating license. NRC did not 

always follow this guidance and generally did not document how it applied 

the guidance. The risk estimate NRC used to help decide whether the plant 

should shut down was also flawed and underestimated the amount of risk 

that Davis-Besse posed. Further, even though underestimated, the estimate 

still exceeded risk levels generally accepted by the agency. 

NRC has taken several significant actions to help prevent reactor vessel 

corrosion from recurring at nuclear power plants. For example, NRC has 

required more extensive vessel examinations and augmented inspector 

training. However, NRC has not yet completed all of its planned actions and, 

more importantly, has no plans to address three systemic weaknesses 

underscored by the incident. Specifically, NRC has proposed no actions to 

help it better (1) identify early indications of deteriorating safety conditions 

at plants, (2) decide whether to shut down a plant, or (3) monitor actions 

taken in response to incidents at plants. Both NRC and GAO had previously 

identified problems in NRC programs that contributed to the Davis-Besse 

incident, yet these problems continue to persist.

The Davis-Besse Nuclear Power Plant in Oak Harbor, Ohio

In March 2002, the most serious 

safety issue confronting the 

nation’s commercial nuclear power 

industry since Three Mile Island in 

1979 was identified at the DavisBesse plant in Ohio. After the 

Nuclear Regulatory Commission 

(NRC) allowed Davis-Besse to 

delay shutting down to inspect its 

reactor vessel for cracked tubing, 

the plant found that leakage from 

these tubes had caused extensive 

corrosion on the vessel head—a 

vital barrier preventing a 

radioactive release. GAO 

determined (1) why NRC did not 

identify and prevent the corrosion, 

(2) whether the process NRC used 

in deciding to delay the shutdown 

was credible, and (3) whether NRC 

is taking sufficient action in the 

wake of the incident to prevent 

similar problems from developing 

at other plants. 

Because the nation’s nuclear 

power plants are aging, GAO is 

recommending that NRC take more 

aggressive actions to mitigate the 

risk of serious safety problems 

occurring at Davis-Besse and other 

nuclear power plants. 

NRC disagreed with two of the 

report’s five recommendations—

that it develop (1) additional means 

to better identify safety problems 

early and (2) guidance for making 

decisions whether to shut down a 

plant. GAO continues to believe 

these recommendations are 

appropriate and should be 

implemented. 
Page i GAO-04-415 Davis-Besse Nuclear Power Plant

Contents

Letter 1

Scope and Methodology 3

Results in Brief 5

Background 8

NRC’s Actions to Oversee Davis-Besse Did Not Provide an Accurate 

Assessment of Safety at the Plant 20

NRC’s Process for Deciding Whether to Allow a Delayed Davis-Besse 

Shutdown Lacked Credibility 33

NRC Has Made Progress in Implementing Recommended Changes, 

but Is Not Addressing Important Systemic Issues 45

Conclusions 57

Recommendations for Executive Action 59

Agency Comments and Our Evaluation 60

Appendixes

Appendix I: Time Line Relating Significant Events of Interest 64

Appendix II: Analysis of the Nuclear Regulatory Commission’s 

Probabilistic Risk Assessment for Davis-Besse 65

Appendix III: Davis-Besse Task Force Recommendations to NRC and Their 

Status, as of March 2004 89

Appendix IV: Comments from the Nuclear Regulatory Commission 94

GAO Comments 114

Appendix V: GAO Contacts and Staff Acknowledgments 129

GAO Contacts 129

Staff Acknowledgments 129

Related GAO Products 130

Table Table 1: Status of Davis-Besse Lessons-Learned Task Force 

Recommendations, as of March 2004 47

Figures Figure 1: Major Components of a Pressurized Water Reactor 12

Figure 2: Major Components of the Davis-Besse Reactor Vessel 

Head and Pressure Boundary 13

Figure 3: Diagram of the Cavity in Davis-Besse’s Reactor Vessel 

Head 17
Contents

Page ii GAO-04-415 Davis-Besse Nuclear Power Plant

Figure 4: The Cavity in Davis-Besse’s Reactor Vessel Head after 

Discovery 18

Figure 5: Rust and Boric Acid on Davis-Besse’s Vessel Head as 

Shown to Resident Inspector during the 2000 Refueling 

Outage 23

Figure 6: NRC’s Acceptance Guidelines for Core Damage 

Frequency 43

Abbreviations

NRC Nuclear Regulatory Commission

PRA Probabilistic risk assessment

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Page 1 GAO-04-415 Davis-Besse Nuclear Power Plant

United States General Accounting Office

Washington, D.C. 20548

A

May 17, 2004 Letr

Congressional Requesters

In 2002, the most serious safety issue confronting the nation’s commercial 

nuclear power industry since the accident at Three Mile Island in 1979 was 

identified at the Davis-Besse nuclear power plant in northwestern Ohio. On 

March 7, 2002, during shutdown for inspection and refueling, the owner of 

the Davis-Besse plant—FirstEnergy Nuclear Operating Company—

discovered a pineapple-sized cavity in the plant’s carbon steel reactor 

vessel head. The reactor vessel head is an 18-foot-diameter, 6-inch-thick, 

80-ton cap that is bolted to the reactor vessel. The vessel head is an integral 

part of the reactor coolant pressure boundary that serves as a vital barrier 

for protecting the environment from any release of radiation from the 

reactor core. In pressurized water reactors such as the one at Davis-Besse, 

the reactor vessel contains the nuclear fuel, as well as water with diluted 

boric acid that cools the fuel and helps control the nuclear reaction. At the 

Davis-Besse plant, vertical tubes had cracked that penetrate the reactor 

vessel head and that contain this water as well as drive mechanisms used to 

lower and raise the fuel, thus allowing leaked boric acid to corrode the 

reactor vessel head. The corrosion had extended through the vessel head to 

a thin stainless steel lining and had likely occurred over a period of several 

years. The lining, which is less than one-third of an inch thick and was not 

designed as a pressure barrier, was found to have a slight bulge with 

evidence of cracking. Had this lining given way, the water within the 

reactor vessel would have escaped, triggering a loss-of-coolant accident, 

which—if back-up safety systems had failed to operate—likely would have 

resulted in the melting of the radioactive core and a subsequent release of 

radioactive materials into the environment. In March 2004, after 2 years of 

increased NRC oversight and considerable repairs by FirstEnergy, NRC 

approved the restart of Davis-Besse’s operations. 

Under the Atomic Energy Act of 1954, as amended, and the Energy 

Reorganization Act of 1974, as amended, the Nuclear Regulatory 

Commission (NRC) and the operators of nuclear power plants share the 

responsibility for ensuring that nuclear reactors are operated safely. NRC is 

responsible for issuing regulations, licensing and inspecting plants, and 

requiring action, as necessary, to protect public health and safety; plant 

operators have the primary responsibility for safely operating the plants in 

accordance with their licenses. NRC has the authority to order plant 

operators to take actions, up to and including shutting down a plant, if 

licensing conditions are not being met and the plant poses an undue risk to 
Page 2 GAO-04-415 Davis-Besse Nuclear Power Plant

public health and safety. In carrying out its responsibilities, NRC relies on, 

among other things, on-site NRC resident inspectors to assess plant 

conditions and quality assurance programs, such as those for maintenance 

and operations, that operators establish to ensure safety at the plant. 

Before the discovery of the cavity in the Davis-Besse reactor vessel head, 

NRC had requested that operators of Davis-Besse and other similar 

pressurized water reactors (1) thoroughly inspect the vertical tubing on 

their reactor vessel heads by December 31, 2001, for possible cracking, or 

(2) justify why their tubing and reactor vessel heads were sufficiently safe 

without being inspected. This request was a reaction to cracked vertical 

tubing found on a pressurized water reactor vessel head at another plant. 

Such thorough inspections require that the reactor be shut down. 

FirstEnergy, believing that its reactor vessel head was safe, asked NRC if its 

shutdown could be delayed until the end of March 2002 to coincide with an 

already scheduled shutdown for refueling—during which time it would 

conduct the requested inspection. FirstEnergy provided evidence 

supporting its assertion that the reactor could continue operating safely. 

After considerable discussion, and after NRC developed a risk assessment 

estimate for deciding that Davis-Besse would not pose an unacceptable 

level of risk, NRC and FirstEnergy compromised, and FirstEnergy agreed 

to shut down the reactor in mid-February 2002 for inspection. Soon after 

Davis-Besse was shut down, the cracked tubes and the significant reactor 

vessel head corrosion were discovered. 

You asked us to determine (1) why NRC did not identify and prevent the 

vessel head corrosion at Davis-Besse, (2) whether the process NRC used 

when deciding to allow FirstEnergy to delay its shutdown was credible, 

and (3) whether NRC is taking sufficient action in the wake of the DavisBesse incident to prevent similar problems from developing in the future at 

Davis-Besse and other nuclear power plants. As agreed with your offices, 

our review focused on NRC’s role in the events leading up to Davis-Besse’s 

shutdown, NRC’s response to the problems discovered, and NRC’s 

management controls over programs and processes that may have 

contributed to the Davis-Besse incident. We did not evaluate the role of 

FirstEnergy because, at the time of our review, NRC’s Office of 

Investigations and the Department of Justice were conducting separate 

inquiries into the potential liability of FirstEnergy concerning its 

knowledge of conditions at Davis-Besse, including the condition of the 

reactor vessel head. We also did not review NRC’s March 2004 decision to 

allow the plant to restart. 
Page 3 GAO-04-415 Davis-Besse Nuclear Power Plant

Scope and 

Methodology

To determine why NRC did not identify and prevent the vessel head 

corrosion at the Davis-Besse nuclear power plant, we reviewed NRC’s 

lessons-learned task force report;1

 FirstEnergy’s root cause analysis 

reports;2 NRC’s Office of the Inspector General reports on Davis-Besse;3

NRC’s augmented inspection team report;4 and NRC’s inspection reports 

and licensee assessments from 1998 through 2001. We also reviewed NRC 

generic communications issued on boric acid corrosion and on nozzle 

cracking. In addition, we interviewed NRC regional officials who were 

involved in overseeing Davis-Besse at the time corrosion was occurring, 

and when the reactor vessel head cavity was found, to learn what 

information they had, their knowledge of plant activities, and how they 

communicated information to headquarters. We also held discussions with 

the resident inspector who was at Davis-Besse at the time that corrosion 

was occurring to determine what information he had and how this 

information was communicated to the regional office. Further, we met with 

FirstEnergy and NRC officials at Davis-Besse and walked through the 

facility, including the containment building, to understand the nature and 

extent of NRC’s oversight of licensees. Additionally, we met with NRC 

headquarters officials to discuss the oversight process as it related to 

Davis-Besse, and the extent of their knowledge of conditions at DavisBesse. We also met with county officials from Ottawa County, Ohio, to 

discuss their views on NRC and Davis-Besse plant safety. Further, we met 

with representatives from a variety of public interest groups to obtain their 

thoughts on NRC’s oversight and the agency’s proposed changes in the 

wake of Davis-Besse. 

1

NRC, Degradation of Davis-Besse Nuclear Power Station Reactor Pressure Vessel Head 

Lessons-Learned Report (Washington, D.C.; Sept. 30, 2002).

2

FirstEnergy, Davis-Besse Nuclear Power Station, Root Cause Analysis Report: Significant 

Degradation of the Reactor Pressure Vessel Head, CR 2002-089 (Oak Harbor, Ohio; Aug. 27, 

2002) and Root Cause Analysis Report: Failure to Identify Significant Degradation of the 

Reactor Pressure Vessel Head, CR-02-0685, 02-0846, 02-0891, 02-1053, 02-1128, 02-

1583, 02-1850, 02-2584, and 02-2585 (Oak Harbor, Ohio; Aug. 13, 2002).

3

NRC, Office of the Inspector General, NRC’s Regulation of Davis-Besse Regarding 

Damage to the Reactor Vessel Head (Washington, D.C.; Dec. 30, 2002) and NRC’s Oversight 

of Davis-Besse Boric Acid Leakage and Corrosion during the April 2000 Refueling Outage 

(Washington, D.C.: Oct. 17, 2003).

4

NRC, Davis-Besse Nuclear Power Station NRC Augmented Inspection Team—

Degradation of the Reactor Pressure Vessel Head (Washington, D.C.; May 3, 2002). 
Page 4 GAO-04-415 Davis-Besse Nuclear Power Plant

To determine whether the process NRC used was credible when deciding 

to allow Davis-Besse to delay its shutdown, we evaluated NRC guidelines 

for reviewing licensee requests for temporary and permanent license 

changes, or amendments to their licenses. We also reviewed NRC guidance 

for making and documenting agency decisions, such as those on whether to 

accept licensee responses to generic communications, as well as NRC’s 

policies and procedures for taking enforcement action. We supplemented 

these reviews with an analysis of internal NRC correspondence related to 

the decision-making process, including e-mail correspondence, notes, and 

briefing slides. We also reviewed NRC’s request for additional information 

to FirstEnergy following the issuance of NRC’s generic bulletin for 

conducting reactor vessel head and nozzle inspections, as well as 

responses provided by FirstEnergy. In addition, we reviewed the draft 

shutdown order that NRC prepared before accepting FirstEnergy’s 

proposal to conduct its inspection in mid-February 2002. We reviewed 

these documents to determine whether the basis for NRC’s decision was 

clearly laid out, persuasive, and defensible to a party outside of NRC. 

As part of our analysis for determining whether NRC’s process was 

credible, we also obtained and reviewed NRC’s probabilistic risk 

assessment (PRA) calculations that it developed to guide its decision 

making. To conduct this analysis, we relied on the advice of consultants 

who, collectively, have an extensive background in nuclear engineering, 

PRA, and metallurgy. These consultants included Dr. John C. Lee, Professor 

and Chair, Nuclear Engineering and Radiological Sciences at the University 

of Michigan’s College of Engineering; Dr. Thomas H. Pigford, Professor 

Emeritus, at the University of California-Berkeley’s College of Engineering; 

and Dr. Gary S. Was, Associate Dean for Research in the College of 

Engineering, and Professor, Nuclear Engineering and Radiological Sciences 

at the University of Michigan’s College of Engineering. These consultants 

reviewed internal NRC correspondence relating to NRC’s PRA estimate, 

NRC’s calculations, and the basis for these calculations. These consultants 

also discussed the basis for NRC’s estimates with NRC officials and outside 

contractors who provided information to NRC as it developed its estimates. 

These consultants were selected on the basis of recommendations made by 

other nuclear engineering experts, their résumés, their collective 

experience, lack of a conflict of interest, and previous experience with 

assessing incidents at nuclear power plants such as Three Mile Island. 

To determine whether NRC is taking sufficient action in the wake of the 

Davis-Besse incident to prevent similar problems from developing in the 

future, we reviewed NRC’s lessons-learned task force recommendations, 
Page 5 GAO-04-415 Davis-Besse Nuclear Power Plant

NRC’s analysis of the underlying causes for failing to identify the corrosion 

of the reactor vessel head, and NRC’s action plan developed in response to 

the task force recommendations. We also reviewed other NRC lessonslearned task force reports and their recommendations, our prior reports to 

identify issues related to those at Davis-Besse, and NRC’s Office of the 

Inspector General reports. We met with NRC officials responsible for 

implementing task force recommendations to obtain a clear understanding 

of the actions they were taking and the status of their efforts, and discussed 

NRC’s recommendations with NRC regional officials, on-site inspectors, 

and representatives from public interest groups. We conducted our review 

from November 2002 through May 2004 in accordance with generally 

accepted government auditing standards.

Results in Brief NRC should have but did not identify or prevent the vessel head corrosion 

at Davis-Besse because both its inspections at the plant and its assessments 

of the operator’s performance yielded inaccurate and incomplete 

information on plant safety conditions. With respect to inspections, NRC 

resident inspectors had information revealing potential problems, such as 

boric acid deposits on the vessel head and air monitors clogged with boric 

acid deposits, but this information did not raise alarms about the plant’s 

safety. NRC inspectors did not know that these indications could signal a 

potentially significant problem and therefore did not fully communicate 

their observations to other NRC staff, some of whom might have 

recognized the significance of the problem. However, even if these staff had 

been informed, according to NRC officials, the agency would have taken 

action only if these indications were considered significant safety 

concerns. Furthermore, NRC’s assessments of Davis-Besse, which include 

inspection results as well as other data, did not provide complete and 

accurate information on FirstEnergy’s performance. For example, NRC 

consistently assessed Davis-Besse’s operator as a “good performer” during 

those years when the corrosion was likely occurring, and the operator was 

not correctly identifying the source of boric acid deposits. NRC had been 

aware for several years that corrosion and cracking were issues that could 

possibly affect safety, but did not view them as immediate safety concerns 

and therefore had not fully incorporated them into its oversight process. 

NRC’s process for deciding whether Davis-Besse could delay its shutdown 

to inspect for nozzle cracking lacks credibility because the guidance NRC 

used was not intended for making such a decision and the basis for the 

decision was not fully documented. In the absence of written guidance 

specifically intended to direct the decision-making process for a shutdown, 
Page 6 GAO-04-415 Davis-Besse Nuclear Power Plant

NRC used guidance designed for considering operator requests for license 

amendments. This guidance describes safety factors that NRC should 

consider in deciding whether to approve a license amendment, as well as a 

process for considering the relative risk the amendment could pose. 

However, the guidance does not specify how NRC should use the safety 

factors, and we could not determine if NRC appropriately followed this 

guidance because it did not clearly document the basis for its decision. For 

example, NRC initially decided that several safety factors were not met and 

considered issuing a shutdown order. Regardless, the agency allowed 

FirstEnergy to delay its shutdown, even though it is not clear whether—

and if so, how—the safety factors were subsequently met. Further, NRC did 

not provide a rationale for its decision for more than a year. NRC also did 

not follow other aspects of its guidance. In the absence of specific 

guidance, and with little documentation of the decision-making process, we 

could not judge whether the agency’s decision was reasonable. Our 

consultants identified substantial problems with how NRC developed and 

used its risk estimate when making the decision. For example, NRC did not 

perform an analysis of the uncertainty associated with the risk estimate; if 

it had, our consultants believe the uncertainty would have been so large as 

to render NRC’s risk estimate of questionable value. Further, the risk 

estimate indicated that the likelihood of an accident occurring at DavisBesse was greater than the level of risk generally accepted as being 

reasonable by NRC.

Responding to the Davis-Besse incident, NRC has taken several significant 

actions to help prevent boric acid from corroding reactor vessel heads at 

nuclear power plants. NRC issued requirements that licensees more 

extensively examine their reactor vessel heads, revised NRC inspection 

guidance used to identify and resolve licensee problems before they affect 

operations, augmented training to keep its inspectors better informed 

about boric acid and cracking issues, and revised guidance to better ensure 

that licensees implement commitments to change their operations. 

However, NRC has not yet implemented more than half of its planned 

actions, and resource constraints could affect the agency’s ability to fully 

and effectively implement the actions. More importantly, NRC is not 

addressing three systemic problems underscored by the Davis-Besse 

incident. First, its process for assessing safety at nuclear power plants is 

not adequate for detecting early indications of deteriorating safety. In this 

respect, the process does not effectively identify changes in the operator’s 

performance or approach to safety before a more serious safety problem 

can develop. Second, NRC’s decision-making guidance does not specifically 

address shutdown decisions or explain how different safety 
Page 7 GAO-04-415 Davis-Besse Nuclear Power Plant

considerations, such as quantitative estimates of risk, should be weighed. 

Third, NRC does not have adequate management controls for 

systematically tracking actions that it has taken in response to incidents at 

plants to determine if the actions were sufficient to resolve underlying 

problems and thereby prevent future incidents. Analyses of earlier 

incidents at other plants identified several issues, such as inadequate 

communication, that contributed to the Davis-Besse incident. Such 

management controls may have helped to resolve these issues before the 

Davis-Besse incident occurred. While NRC is monitoring how it 

implements actions taken as a result of the Davis-Besse incident, the 

agency has not yet committed to a process for assessing the effectiveness 

of actions taken. 

Given NRC’s actions in response to Davis-Besse, severe vessel head 

corrosion is unlikely to occur at a plant any time soon. However, in part 

because of unresolved systemic problems, another incident unrelated to 

vessel head corrosion could occur in the future. As a result, we are 

recommending that NRC take more aggressive and specific actions in 

several areas, such as revising how it assesses plant performance, 

establishing a more specific methodology for deciding to shut down a 

plant, and establishing management controls for monitoring and assessing 

the effectiveness of changes made in response to task force findings. 

In commenting on a draft of this report, NRC generally addressed only 

those findings and recommendations with which it disagreed. While 

commenting that it agreed with many of our findings, the agency said that 

the report overall does not appropriately characterize or provide a 

balanced perspective on NRC’s actions surrounding the discovery of the 

reactor vessel head condition at Davis-Besse or its efforts to incorporate 

the lessons learned from that experience into its processes. More 

specifically, NRC stated that the report does not acknowledge that NRC 

must rely heavily on its licensees to provide complete and accurate 

information. NRC also expressed concern about the report’s 

characterization of its use of risk estimates. We believe that the report 

fairly and accurately describes NRC’s actions regarding the Davis-Besse 

incident. Nonetheless, we expanded our discussion of NRC’s roles and 

responsibilities to point out that licensees are required to provide NRC with 

complete and accurate information.

NRC disagreed with our recommendations to develop (1) specific guidance 

and a well-defined process for deciding when to shut down a plant and (2) 

a methodology to assess early indications of deteriorating safety at nuclear 
Page 8 GAO-04-415 Davis-Besse Nuclear Power Plant

power plants. NRC stated that it has sufficient guidance to make plant 

shutdown decisions. NRC also stated that, as regulators, the agency is not 

charged with managing licensees’ facilities and that direct involvement 

with those aspects of licensees’ operations that could provide it with 

information on early indications of deteriorating safety crosses over to a 

management function. We continue to believe that NRC should develop 

specific guidance and a well-defined process to decide when to shut down 

a plant. In absence of such guidance for making the Davis-Besse shutdown 

decision, NRC used its guidance for considering operators’ requests for 

amendments to their licenses. This guidance describes safety factors that 

NRC should consider in deciding whether to approve license changes, as 

well as a process for considering the relative risk the amendment would 

pose. This guidance does not specify how NRC should use the safety 

factors. We also continue to believe that NRC should develop a 

methodology to assess aspects of licensees’ operations as a means to have 

an early warning of developing safety problems. In implementing this 

recommendation, we envision that NRC would be analyzing data for 

changes in operators’ performance or approach to safety, not prescribing 

how the plants are managed. 

Background

NRC’s Role and 

Responsibilities

NRC, as an independent federal agency, regulates the commercial uses of 

nuclear material to ensure adequate protection of public health and safety 

and the environment. NRC is headed by a five-member commission 

appointed by the President and confirmed by the Senate; one 

commissioner is appointed as chairman.5

 NRC has about 2,900 employees 

who work in its headquarters office in Rockville, Maryland, and its four 

regional offices. NRC is financed primarily by fees that it imposes on 

commercial users of the nuclear material that it regulates. For fiscal year 

2004, NRC’s appropriated budget of $626 million includes about $546 

million financed by these fees.

NRC regulates the nation’s commercial nuclear power plants by 

establishing requirements for plant owners and operators to follow in the 

design, construction, and operation of the nuclear reactors. NRC also 

5

Two commissioner positions are currently vacant.
Page 9 GAO-04-415 Davis-Besse Nuclear Power Plant

licenses the reactors and individuals who operate them. Currently, 104 

commercial nuclear reactors at 65 locations are licensed to operate.6

 Many 

of these reactors have been in service since the early to mid-1970s. NRC 

initially licensed the reactors to operate for 40 years, but as these licenses 

approach their expiration dates, NRC has been granting 20-year extensions. 

To ensure the reactors are operated within their licensing requirements and 

technical specifications, NRC oversees them by both inspecting activities 

at the plants and assessing plant performance.7

 NRC’s inspections consist 

of both routine, or baseline, inspections and supplemental inspections to 

assess particular licensee programs or issues that arise at a power plant. 

Inspections may also occur in response to a specific operational problem 

or event that has occurred at a plant. NRC maintains inspectors at every 

operating nuclear power plant in the United States and supplements the 

inspections conducted by these resident inspectors with inspections 

conducted by staff from its regional offices and from headquarters. 

Generally, inspectors verify that the plant’s operator qualifications and 

operations, engineering, maintenance, fuel handling, emergency 

preparedness, and environmental and radiation protection programs are 

adequate and comply with NRC safety requirements. NRC also oversees 

licensees by requesting information on their activities. NRC requires that 

information provided by licensees be complete and accurate and, 

according to NRC officials, this is an important aspect of the agency’s 

oversight.8

 While we have added information to this report on the 

requirement that licensees provide NRC with complete and accurate 

information, we believe that NRC’s oversight program should not place 

undue reliance on this requirement.

Nuclear power plants have many physical structures, systems, and 

components, and licensees have numerous activities under way, 24-hours a 

6

These licensed reactors include Browns Ferry Unit 1—one of three reactors owned by the 

Tennessee Valley Authority in Alabama—which was shut down in 1985. The Tennessee 

Valley Authority plans to restart the reactor in 2007, which will require NRC approval. 

7

NRC’s oversight program has changed significantly since the beginning of 1998. The third 

and most recent change occurred in mid-2000, when the agency adopted its Reactor 

Oversight Process. Under this process, NRC continues to rely on inspection results to assess 

licensee performance. However, it supplements this information with other indicators of 

self-reported licensee performance, such as how frequently unscheduled shutdowns occur. 

8

10 C.F.R. § 50.9 requires that information provided by licensees be complete and accurate in 

all material respects. 
Page 10 GAO-04-415 Davis-Besse Nuclear Power Plant

day, to ensure the plants operate safely. Programs to ensure quality 

assurance and safe operations include monitoring, maintenance, and 

inspection. To carry out these programs, licensees typically prepare several 

thousand reports per year describing conditions at the plant that need to be 

addressed to ensure continued safe operations. Because of the large 

number of activities and physical structures, systems, and components, 

NRC focuses its inspections on those activities and pieces of equipment or 

systems that are considered to be most significant for protecting public 

health and safety. NRC terms this a “risk-informed” approach for regulating 

nuclear power plants. Under this risk-informed approach, some systems 

and activities that NRC considers to have relatively less safety significance 

receive little NRC oversight. NRC has adopted a risk-informed approach 

because it believes it can focus its regulatory resources on those areas of 

the plant that the agency considers to be most important to safety. In 

addition, it was able to adopt this approach because, according to NRC, 

safety performance at nuclear power plants has improved as a result of 

more than 25 years of operating experience.

To decide whether inspection findings are minor or major, NRC uses a 

process it began in 2000 to determine the extent to which violations 

compromise plant safety. Under this process, NRC characterizes the 

significance of its inspection findings by using a significance determination 

process to evaluate how an inspection finding impacts the margin of safety 

at a power plant. NRC has a range of enforcement actions it can take, 

depending on how much the safety of the plant had been compromised. For 

findings that have low safety significance, NRC can choose to take no 

formal enforcement action. In these instances, nonetheless, licensees 

remain responsible for addressing the identified problems. For more 

serious findings, NRC may take more formal action, such as issuing 

enforcement orders. Orders can be used to modify, suspend, or even revoke 

an operating license. NRC has issued one enforcement order to shut down 

an operating power plant in its 28-year history—in 1987, after NRC 

discovered control room personnel sleeping while on duty at the Peach 

Bottom nuclear power plant in Pennsylvania. In addition to enforcement 

orders, NRC can issue civil penalties of up to $120,000 per violation per day. 

Although NRC does not normally use civil penalties for violations 

associated with its Reactor Oversight Process, NRC will consider using 

them for issues that are willful, have the potential for impacting the 

agency’s regulatory process, or have actual public health and safety 

consequences. In fiscal year 2003, NRC proposed imposing civil penalties 

totaling $120,000 against two power plant licensees for the failure to 

provide complete and accurate information to the agency. 
Page 11 GAO-04-415 Davis-Besse Nuclear Power Plant

NRC uses generic communications—such as bulletins, generic letters, and 

information notices—to provide information to and request information 

from the nuclear industry at large or specific groups of licensees. Bulletins 

and generic letters both usually request information from licensees 

regarding their compliance with specific regulations. They do not require 

licensees to take any specific actions, but do require licensees to provide 

responses to the information requests. In general, NRC uses bulletins, as 

opposed to generic letters, to address significant issues of greater urgency. 

NRC uses information notices to transmit significant recently identified 

information about safety, safeguards, or environmental issues. Licensees 

are expected to review the information to determine whether it is 

applicable to their operations and consider action to avoid similar 

problems. 

Operation of Pressurized 

Water Nuclear Power Plants 

and Events Leading to the 

March 2002 Discovery of 

Serious Corrosion

The Davis-Besse Nuclear Power Station, owned and operated by 

FirstEnergy Nuclear Operating Company, is an 882-megawatt electric 

pressurized water reactor located on Lake Erie in Oak Harbor, Ohio, about 

20 miles east of Toledo. The power plant is under NRC’s Region III 

oversight, which is located in Lisle, Illinois. Like other pressurized water 

reactors, Davis-Besse is designed with multiple barriers between the 

radioactive heat-producing core and the outside environment—a design 

concept called “defense-in-depth.” Three main design components provide 

defense-in-depth. First, the reactor core is designed to retain radioactive 

material within the uranium oxide fuel, which is also covered with a layer 

of metal tubing. Second, a 6-inch-thick carbon steel vessel, lined with threesixteenth-inch-thick stainless steel, surrounds the reactor core. Third, a 

steel containment structure, surrounded by a thick reinforced concrete 

building, encloses the reactor vessel and other systems and components 

important for maintaining safety. The containment structure and concrete 

building are intended to help not only prevent a release of radioactivity to 

the environment, but also shield the reactor from external hazards like 

tornados and missiles. The reactor vessel, in addition to housing the 

reactor core, contains highly pressurized water to cool the radioactive 

heat-producing core and transfer heat to a steam generator. Consequently, 

the vessel is referred to as the reactor pressure vessel. From the vessel, hot 

pressurized water is piped to the steam generator, where a separate supply 

of water is turned to steam to drive turbines that generate electricity. (See 

fig. 1.) 
Page 12 GAO-04-415 Davis-Besse Nuclear Power Plant

Figure 1: Major Components of a Pressurized Water Reactor

The top portion of the Davis-Besse reactor pressure vessel consisted of an 

18-foot-diameter vessel head that was bolted to the lower portion of the 

pressure vessel. At Davis-Besse, 69 vertical tubes penetrated and were 

welded to the vessel head. These tubes, called vessel head penetration 

nozzles, contained control rods that, when raised or lowered, were used to 

moderate or shut down the nuclear reaction in the reactor.9

 Because 

control rods attach to control rod drive mechanisms, these types of nozzles 

are referred to as control rod drive mechanism nozzles. A platform, known 

as the service structure, sat above the reactor vessel head and the control 

rod drive mechanism nozzles. Inside the service structure and above the 

pressure vessel head was a layer of insulation to help contain the heat 

emanating from the reactor. The sides of the lower portion of the service 

Containment structure

Reactor vessel

Steam

generator

Pressurizer

Turbine

Condenser

Generator

Control rods

Source: NRC.

9

While Davis-Besse had 69 nozzles, 7 were spare and 1 was used for head vent piping. 
Page 13 GAO-04-415 Davis-Besse Nuclear Power Plant

structure were perforated with 18 5- by 7-inch rectangular openings, 

termed “mouse-holes,” that were used for vessel head inspections. In 

pressurized water reactors such as Davis-Besse, the reactor vessel, the 

vessel head, the nozzles, and other equipment used to ensure a continuous 

supply of pressurized water in the reactor vessel are collectively referred to 

as the reactor coolant pressure boundary. (See fig. 2.)

Figure 2: Major Components of the Davis-Besse Reactor Vessel Head and Pressure 

Boundary

To better control the nuclear reaction at nuclear power plants, boron in the 

form of boric acid crystals is dissolved in the cooling water contained 

within the reactor vessel and pressure boundary. Boric acid, under certain 

2-inch minimum

gap between insulation

and top of vessel head

18 access openings

(mouse-holes) at

Davis Besse

Support

steel

Service structure

Nozzles Vessel head insulation

Source: FirstEnergy.
Page 14 GAO-04-415 Davis-Besse Nuclear Power Plant

conditions, can cause corrosion of carbon steel. For about 3 decades, NRC 

and the nuclear power industry have known that boric acid had the 

potential to corrode reactor components. In general, if leakage occurs from 

the reactor coolant system, the escaping coolant will flash to steam and 

leave behind a concentration of impurities, including noncorrosive dry 

boric acid crystals. However, under certain conditions, the coolant will not 

flash to steam, and the boric acid will remain in a liquid state where it can 

cause extensive and rapid degradation of any carbon steel components it 

contacts. Such extensive degradation, in both domestic and foreign 

pressurized water reactor plants, has been well documented and led NRC 

to issue a generic letter in 1988 requesting information from pressurized 

water reactor licensees to ensure they had implemented programs to 

control boric acid corrosion. NRC was primarily concerned that boric acid 

corrosion could compromise the reactor coolant pressure boundary. This 

concern also led NRC to develop a procedure for inspecting licensees’ 

boric acid corrosion control programs and led the Electric Power Research 

Institute to issue guidance on boric acid corrosion control.10

NRC and the nuclear power industry have also known that nozzles made of 

alloy 600,11 used in several areas within nuclear power plants, were prone 

to cracking. Cracking had become an increasingly topical issue as the 

nuclear power plant fleet has aged. In 1986, operators at domestic and 

foreign pressurized water reactors began reporting leaks in various types of 

alloy 600 nozzles. In 1989, after leakage was detected at a domestic plant, 

NRC identified the cause of the leakage as cracking due to primary water 

stress corrosion.12 However, NRC concluded that the cracking was not an 

immediate safety concern for a few reasons. For example, the cracks had a 

low growth rate, were in a material with an extremely high flaw tolerance 

and, accordingly, were unlikely to spread. Also, the cracks were axial—that 

is, they ran the length of the nozzle rather than its circumference. NRC and 

10The Electric Power Research Institute is a nonprofit energy research consortium whose 

members include utilities. It provides science and technology-based solutions to members 

through its scientific research, technology development, and product implementation 

program. 

11Alloy 600 is an alloy of nickel, chromium, iron, and minor amounts of other elements. The 

alloy is highly resistant to general corrosion but can be susceptible to cracking at high 

temperatures.

12Primary water stress corrosion cracking refers to cracking under stress and in primary 

coolant water. The primary water coolant system is that portion of a nuclear power plant’s 

coolant system that cools the reactor core in the reactor pressure vessel and deposits heat 

to the steam generator. 
Page 15 GAO-04-415 Davis-Besse Nuclear Power Plant

the nuclear power industry were more concerned that circumferential 

cracks could result in broken or snapped nozzles. NRC did, however, issue 

a generic information notice in 1990 to inform the industry of alloy 600 

cracking. Through the early 1990s, NRC, the Nuclear Energy Institute,13 and 

others continued to monitor alloy 600 cracking. In 1997, continued concern 

over cracking led NRC to issue a generic letter to pressurized water reactor 

licensees requesting information on their plans to monitor and manage 

cracking in vessel head penetration nozzles as well as to examine these 

nozzles. 

In the spring of 2001, licensee inspections led to the discovery of large 

circumferential cracking in several vessel head penetration nozzles at the 

Oconee Nuclear Station, in South Carolina. As a result of the discovery, the 

nuclear power industry and NRC categorized the 69 operating pressurized 

water reactors in the United States into different groups on the basis of (1) 

whether cracking had already been found and (2) how similar they were to 

Oconee in terms of the amount of time and the temperature at which the 

reactors had operated. The industry had developed information indicating 

that greater operating time and temperature were related to cracking. In 

total, five reactors at three locations were categorized as having already 

identified cracking, while seven reactors at five locations were categorized 

as being highly susceptible, given their similarity to Oconee.14

In August 2001, NRC issued a bulletin requesting that licensees of these 

reactors provide, within 30 days, information on their plans for conducting 

nozzle inspections before December 31, 2001.15 In lieu of this information, 

NRC stated that licensees could provide the agency with a reasoned basis 

for their conclusions that their reactor vessel pressure boundaries would 

continue to meet regulatory requirements for ensuring the structural 

integrity of the reactor coolant pressure boundary until the licensees 

13The Nuclear Energy Institute comprises companies that operate commercial power plants 

and supports the commercial nuclear industry; and universities, research laboratories, and 

labor unions affiliated with the nuclear industry. Among other things, it provides a forum to 

resolve technical and business issues and offers information to its members and 

policymakers on nuclear issues.

14Reactors that were categorized as having already identified cracking or were highly 

susceptible included Arkansas Nuclear reactor unit 1; D.C. Cook reactor unit 2; Davis-Besse; 

North Anna reactor units 1 and 2; Oconee reactor units 1, 2 and 3; Robinson reactor unit 2; 

Surry reactor units 1 and 2; and Three Mile Island reactor unit 1.

15NRC, “Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles” 

(Bulletin 2001-01, Aug. 8, 2001).
Page 16 GAO-04-415 Davis-Besse Nuclear Power Plant

conducted their inspections. NRC used a bulletin, as opposed to a generic 

letter, to request this information because cracking was considered a 

significant and urgent issue. All of the licensees of the highly susceptible 

reactors, except Davis-Besse and D.C. Cook reactor unit 2, provided NRC 

with plans for conducting inspections by December 31, 2001.16

In September 2001, FirstEnergy proposed conducting the requested 

inspection in April 2002, following its planned March 31, 2002, shutdown to 

replace fuel. In making this proposal, FirstEnergy contended that the 

reactor coolant pressure boundary at Davis-Besse met and would continue 

to meet regulatory requirements until its inspection. NRC and FirstEnergy 

exchanged information throughout the fall of 2001 regarding when 

FirstEnergy would conduct the inspection at Davis-Besse. NRC drafted an 

enforcement order that would have shut down Davis-Besse by December 

2001 for the requested inspection in the event that FirstEnergy could not 

provide an adequate justification for safe operation beyond December 31, 

2001, but ultimately compromised on a mid-February 2002 shutdown date. 

NRC, in deciding when FirstEnergy had to shut down Davis-Besse for the 

inspection, used a risk-informed decision-making process, including 

probabilistic risk assessment (PRA), to conclude that the risk that DavisBesse would have an accident in the interim was relatively low. PRA is an 

analytical tool for estimating the probability that a potential accident might 

occur by examining how physical structures, systems, and components, 

along with employees, work together to ensure plant safety. 

Following the mid-February 2002 shutdown and in the course of its 

inspection in March 2002, FirstEnergy removed about 900 pounds of boric 

acid crystals and powder from the reactor vessel head, and subsequently 

discovered three cracked nozzles. The number of nozzles that had cracked, 

as well as the extent of cracking, was consistent with analyses that NRC 

staff had conducted prior to the shutdown. However, in examining the 

extent of cracking, FirstEnergy also discovered that corrosion had caused 

a pineapple-sized cavity in the reactor vessel head. (See figs. 3 and 4.) 

16The licensee for D.C. Cook reactor unit 2 proposed to shut down in mid-January 2002 for 

its inspection. NRC agreed to the delay after crediting D.C. Cook for having been shut down 

for about a month during the fall of 2001, thus reducing the reactor’s operating time.
Page 17 GAO-04-415 Davis-Besse Nuclear Power Plant

Figure 3: Diagram of the Cavity in Davis-Besse’s Reactor Vessel Head
Page 18 GAO-04-415 Davis-Besse Nuclear Power Plant

Figure 4: The Cavity in Davis-Besse’s Reactor Vessel Head after Discovery

After this discovery, NRC directed FirstEnergy to, among other things, 

determine the root cause of the corrosion and obtain NRC approval before 

restarting Davis-Besse. NRC also dispatched an augmented inspection 

team consisting of NRC resident, regional, and headquarters officials.17 The 

inspection team concluded that the cavity was caused by boric acid 

corrosion from leaks through the control rod drive mechanism nozzles in 

the reactor vessel head. Primary water stress corrosion cracking of the 

nozzles caused through-wall cracks, which led to the leakage and eventual 

corrosion of the vessel head. NRC’s inspection team also concluded, among 

other things, that this corrosion had gone undetected for an extended 

period of time—at least 4 years—and significantly compromised the plant’s 

17NRC forms such inspection teams to ensure that the agency investigates significant 

operational events in a timely, objective, systematic, and technically sound manner, and 

identifies and documents the causes of such events.

Source: FirstEnergy.
Page 19 GAO-04-415 Davis-Besse Nuclear Power Plant

safety margins. As of May 2004, NRC had not yet completed other analyses, 

including how long Davis-Besse could have continued to operate with the 

corrosion it had experienced before a vessel head loss-of-coolant accident 

would have occurred.18 However, on May 4, 2004, NRC released preliminary 

results of its analysis of the vessel head and cracked cladding. Based on its 

analysis of conditions that existed on February 16, 2002, NRC estimated 

that Davis-Besse could have operated for another 2 to 13 months without 

the vessel head failing. However, the agency cautioned that this estimate 

was based on several uncertainties associated with the complex network of 

cracks on the cladding and the lack of knowledge about corrosion and 

cracking rates. NRC plans to use this data in preparing its preliminary 

analysis of how, and the likelihood that, the events at Davis-Besse could 

have led to core damage. NRC plans to complete this preliminary analysis 

in the summer of 2004.

NRC also established a special oversight panel to (1) coordinate NRC’s 

efforts to assess FirstEnergy’s performance problems that resulted in the 

corrosion damage, (2) monitor Davis-Besse’s corrective actions, and (3) 

evaluate the plant’s readiness to resume operations. The panel, which is 

referred to as the Davis-Besse Oversight Panel, comprises officials from 

NRC’s Region III office in Lisle, Illinois; NRC headquarters; and the resident 

inspector office at Davis-Besse. In addition to overseeing FirstEnergy’s 

performance during the shutdown and through restart of Davis-Besse, the 

panel holds public meetings in Oak Harbor, Ohio, where the plant is 

located, and nearby Port Clinton, Ohio, to inform the public about its 

oversight of Davis-Besse’s restart efforts and its views on the adequacy of 

these efforts. The panel developed a checklist of issues that FirstEnergy 

had to resolve prior to restarting: (1) replacing the vessel head and 

ensuring the adequacy of other equipment important for safety, (2) 

correcting FirstEnergy programs that led to the corrosion, and (3) ensuring 

FirstEnergy’s readiness to restart. To restart the plant, FirstEnergy, among 

other things, removed the damaged reactor vessel head, purchased and 

installed a new head, replaced management at the plant, and took steps to 

improve key programs that should have prevented or detected the 

corrosion. As of March 2004, when NRC gave its approval for Davis-Besse 

to resume

18NRC has an Accident Sequence Precursor Analysis Program to analyze significant events 

that occur at nuclear power plants to determine how, and the likelihood that, the events 

could have led to core damage.
Page 20 GAO-04-415 Davis-Besse Nuclear Power Plant

operations, the shutdown and preparations for restart had cost FirstEnergy 

approximately $640 million.19

In addition, NRC established a task force to evaluate its regulatory 

processes for assuring reactor pressure vessel head integrity and to identify 

and recommend areas for improvement that may be applicable to either 

NRC or the nuclear power industry. The task force’s report, which was 

issued in September 2002, contains 51 recommendations aimed primarily at 

improving NRC’s process for inspecting and overseeing licensees, 

communicating with industry, and identifying potential emerging technical 

issues that could impact plant safety. NRC developed an action plan to 

implement the report’s recommendations. 

NRC’s Actions to 

Oversee Davis-Besse 

Did Not Provide an 

Accurate Assessment 

of Safety at the Plant 

NRC’s inspections and assessments of FirstEnergy’s operations should 

have but did not provide the agency with an accurate understanding of 

safety conditions at Davis-Besse, and thus NRC failed to identify or prevent 

the vessel head corrosion. Some NRC inspectors were aware of the 

indications of corrosion and leakage that could have alerted NRC to 

corrosion problems at the plant, but they did not have the knowledge to 

recognize the significance of this information. These problems were 

compounded by NRC’s assessments of FirstEnergy that led the agency to 

believe FirstEnergy was a good performer and could or would successfully 

resolve problems before they became significant safety issues. More 

broadly, NRC had a range of information that could have identified and 

prevented the incident at Davis-Besse but did not effectively integrate it 

into its oversight. 

19FirstEnergy spent about $293 million on operations, maintenance, and capital projects 

(including $47 million for the new reactor vessel head) and $348 million to purchase power 

to replace the power that Davis-Besse would have generated over the 2-year shutdown 

period. In contrast, during a more routine refueling outage, Davis-Besse would spend about 

$60 million—about $37 million on operations, maintenance, and capital projects and $23 

million on replacing the power that would have been generated over a 42-day shutdown 

period. These latter estimates are based on the Davis-Besse refueling outage in midcalendar 

year 2000.
Page 21 GAO-04-415 Davis-Besse Nuclear Power Plant

Several Factors Contributed 

to the Inadequacy of NRC’s 

Inspections for Determining 

Plant Conditions

Three separate, but related, NRC inspection program factors contributed to 

the development of the corrosion problems at Davis-Besse. First, resident 

inspectors did not know that the boric acid, rust, and unidentified leaking 

indicated that the reactor vessel head might be degrading. Second, these 

inspectors thought they understood the cause for the indications, based on 

licensee actions to address them. Therefore, resident inspectors, as well as 

regional and headquarters officials, did not fully communicate information 

on the indications or decide how to address them, and therefore took no 

action. Third, because the significance of the symptoms was not fully 

recognized, NRC did not direct sufficient inspector resources to 

aggressively investigate the indicators. NRC might have taken a different 

approach to the Davis-Besse situation if its program to identify emerging 

issues important to safety had pursued earlier concerns about boric acid 

corrosion and cracking and recognized how they could affect safety.

Inspectors Did Not Know Safety 

Significance of Observed 

Problems

NRC limits the amount of unidentified leakage from the reactor coolant 

system to no more than 1 gallon per minute. When this limit is exceeded, 

NRC requires that licensees identify and correct any sources of 

unidentified leakage. NRC also prohibits any leakage from the reactor 

coolant pressure boundary, of which the reactor vessel is a key component. 

Such leakage is prohibited because the pressure boundary is key to 

maintaining adequate coolant around the reactor fuel and thus protects 

public health and safety. Because of this, NRC’s technical specification 

states that licensees are to monitor reactor coolant leakage and shut down 

within 36 hours if leakage is found in the pressure boundary.

In the years leading up to FirstEnergy’s March 2002 discovery that DavisBesse’s vessel head had corroded extensively, NRC had several indications 

of potential leakage problems. First, NRC knew that the rates of leakage in 

the reactor coolant system had increased. Between 1995 and mid-1998, the 

unidentified leakage rate was about 0.06 gallon per minute or less, 

according to FirstEnergy’s monitoring. In mid-1998, the unidentified 

reactor coolant system leakage rate increased significantly—to as much as 

0.8 gallon per minute. The elevated leakage rate was dominated by a known 

problem with a leaking relief valve on the reactor coolant system 

pressurizer tank, which masked the ongoing leak on the reactor pressure 

vessel head. However, the elevated leak rate should have raised concerns. 

To investigate this leakage, as well as to repair other equipment, 

FirstEnergy shut down the plant in mid-1999. It then identified a faulty 

relief valve that accounted for much of the leakage and repaired the valve. 
Page 22 GAO-04-415 Davis-Besse Nuclear Power Plant

However, after restarting Davis-Besse, the unidentified leakage rate 

remained significantly higher than the historical average. Specifically, the 

unidentified leakage rate varied between 0.15 and 0.25 gallon per minute as 

opposed to the historical low of about 0.06 gallon or less. While NRC was 

aware that the rate was higher than before, NRC did not aggressively 

pursue the difference because the rate was well below NRC’s limit of no 

more than 1 gallon per minute, and thus the leak was not viewed as being a 

significant safety concern. Following the repair in 1999, NRC’s inspection 

report concluded that FirstEnergy’s efforts to reduce the leak rate during 

the outage were effective. 

Second, NRC was aware of increased levels of boric acid in the 

containment building—an indication that components containing reactor 

coolant were leaking. So much boric acid was being deposited that 

FirstEnergy officials had to repeatedly clean the containment air cooling 

system and radiation monitor filters. For example, before 1998, the 

containment air coolers seldom needed cleaning, but FirstEnergy had to 

clean them 28 times between late 1998 and May 2001. Between May 2001 

and the mid-February 2002 shutdown, the containment air coolers were not 

cleaned, but at shutdown, FirstEnergy removed 15 5-gallon buckets of 

boric acid from the coolers—which is almost as much as was found on the 

reactor pressure vessel head. Rather than seeing these repeated cleanings 

as an indication of a problem that needed to be addressed, FirstEnergy 

made cleaning the coolers a routine maintenance activity, which NRC did 

not consider significant enough to require additional inspections. 

Furthermore, the radiation monitors, used to sample air from the 

containment building to detect radiation, typically required new filters 

every month. However, from 1998 to 2002, these monitors became clogged 

and inoperable hundreds of times because of boric acid, despite 

FirstEnergy’s efforts to fix the problem. 

Third, NRC was aware that FirstEnergy found rust in the containment 

building. The radiation monitor filters had accumulated dark colored iron 

oxide particles—a product of carbon steel corrosion—that were likely to 

have resulted from a very small steam leak. NRC inspection reports during 

the summer and fall of 1999 noted these indications and, while recognizing 

FirstEnergy’s aggressive attempts to identify the reasons for the 

phenomenon, concluded that they were a “distraction to plant personnel.” 

Several NRC inspection reports noted indications of leakage, boric acid, 

and rust before the agency adopted its new Reactor Oversight Process in 

2000, but because the leakage was within NRC’s technical specifications 

and NRC officials thought that the licensee understood and would fix the 
Page 23 GAO-04-415 Davis-Besse Nuclear Power Plant

problem, NRC did not aggressively pursue the indications. NRC’s new 

oversight process, implemented in the spring of 2000, limited the issues 

that could be discussed in NRC inspection reports to those that the agency 

considers to have more than minor significance. Because the leakage rates 

were below NRC’s limits, NRC’s inspection reports following the 

implementation of NRC’s new oversight process did not identify any 

discussion of these problems at the plant.

Fourth, NRC was aware that FirstEnergy found rust on the Davis-Besse 

reactor vessel head, but it did not recognize its significance. For instance, 

during the 2000 refueling outage, a FirstEnergy official said he showed one 

of the two NRC resident inspectors a report that included photographs of 

rust-colored boric acid on the vessel head. (See fig. 5.) 

Figure 5: Rust and Boric Acid on Davis-Besse’s Vessel Head as Shown to Resident 

Inspector during the 2000 Refueling Outage

Source: FirstEnergy. 
Page 24 GAO-04-415 Davis-Besse Nuclear Power Plant

According to this resident inspector, he did not recall seeing the report or 

photographs but had no reason to doubt the FirstEnergy official’s 

statement. Regardless, he stated that had he seen the photographs, he 

would not have considered the condition to be significant at the time. He 

said that he did not know what the rust and boric acid might have 

indicated, and he assumed that FirstEnergy would take care of the vessel 

head before restarting. The second resident inspector said he reviewed all 

such reports at Davis-Besse but did not recall seeing the photographs or 

this particular report. He stated that it was quite possible that he had read 

the report, but because the licensee had a plan to clean the vessel head, he 

would have concluded that the licensee would correct the matter before 

plant restart. However, FirstEnergy did not accomplish this, even though 

work orders and subsequent licensee reports indicated that this was done. 

According to the NRC resident inspector and NRC regional officials, 

because of the large number of licensee activities that occur during a 

refueling outage, NRC inspectors do not have the time to investigate or 

follow up on every issue, particularly when the issue is not viewed as being 

important to safety. While the resident inspector informed regional officials 

about conditions at Davis-Besse, the regional office did not direct more 

inspection resources to the plant, or instruct the resident inspector to 

conduct more focused oversight. Some NRC regional officials were aware 

of indications of boric acid corrosion at the plant; others were not. 

According to the Office of the Inspector General’s investigation and 2003 

report on Davis-Besse,20 the NRC regional branch chief—who supervised 

the staff responsible for overseeing FirstEnergy’s vessel head inspection 

activities during the 2000 refueling outage—said that he was unaware of 

the boric acid leakage issues at Davis-Besse, including its effects on the 

containment air coolers and the radiation monitor filters. Had his staff been 

requested to look at these specific issues, he might have directed 

inspection resources to that area. (App. I provides a time line showing 

significant events of interest.)

NRC Did Not Fully Communicate 

Indications 

NRC was not fully aware of the indications of a potential problem at DavisBesse because NRC’s process for transmitting information from resident 

inspectors to regional offices and headquarters did not ensure that 

information was fully communicated, evaluated, or used. NRC staff 

communicated information about plant operations through inspection 

reports, licensee assessments, and daily conference calls that included 

20NRC, Office of the Inspector General, NRC’s Oversight of Davis-Besse during the April 

2000 Refueling Outage (Washington, D.C.: Oct. 17, 2003).
Page 25 GAO-04-415 Davis-Besse Nuclear Power Plant

resident, regional, and headquarters officials. According to regional 

officials, information that is not considered important is not routinely 

communicated to NRC management and technical specialists. For 

example, while the resident inspectors at Davis-Besse knew all of the 

indications of leakage, and there was some level of knowledge about these 

indications at the regional office level, the knowledge was not sufficiently 

widespread within NRC to alert a technical specialist who might have 

recognized their safety significance. According to NRC Region III officials, 

the region uses an informal means—memorandums sent to other regions 

and headquarters—of communicating information identified at plants that 

it considers to be important to safety. However, because the indications at 

Davis-Besse were not considered important, officials did not transmit this 

information to headquarters. Further, because the process is informal, 

these officials said they did not know whether—and if so, how—other NRC 

regions or headquarters used this information. 

Similarly, NRC officials said that NRC headquarters had no systematic 

process for communicating information, such as on boric acid corrosion, 

cracking, and small amounts of unidentified leakage, that had not yet risen 

to a relatively high level of concern within the agency, in a timely manner to 

its regions or on-site inspectors. For example, the regional inspector that 

oversaw FirstEnergy’s activities during the 2000 refueling outage, including 

the reactor vessel head inspection, stated that he was not aware of NRC’s 

generic bulletins and letters pertaining to boric acid and corrosion, even 

though NRC issues only a few of these bulletins and generic letters each 

year.21 In addition, according to NRC regional officials and the resident 

inspector at Davis-Besse, there is little time to review technical reports 

about emerging safety issues that NRC compiles because they are too 

lengthy and detailed. Ineffective communication, both within the region 

and between NRC headquarters and the region, was a primary factor cited 

by NRC’s Office of the Inspector General in its investigation of NRC’s 

oversight of Davis-Besse boric acid leakage and corrosion. 22 For example, 

it found that ineffective communication resulted in senior regional 

management being largely unaware of repeated reports of boric acid 

leakage at Davis-Besse. It also found that headquarters, in communications 

with the regions, did not emphasize the issues discussed in its generic 

21Over the last 10 years, NRC has issued an average of about two generic bulletins and about 

four generic letters a year.

22NRC, Office of the Inspector General, NRC’s Oversight of Davis-Besse during the April 

2000 Refueling Outage (Washington, D.C.; Oct. 17, 2003).
Page 26 GAO-04-415 Davis-Besse Nuclear Power Plant

letters or bulletins on boric acid corrosion or cracking. NRC programs for 

informing its inspectors about issues that can reduce safety at nuclear 

power plants were not effective. As a result, NRC inspectors did not 

recognize the significance of the indications at Davis-Besse, fully 

communicate information about the indications, or spend additional effort 

to follow up on the indications. 

Resource Constraints Affected 

NRC Oversight

NRC also did not focus on the indications that the vessel head was 

corroding because of several staff constraints. Region III was directing 

resources to other plants that had experienced problems throughout the 

region, and these plants thus were the subject of increased regulatory 

oversight. For example, during the refueling outages in 1998 and 2000, 

while NRC oversaw FirstEnergy’s inspection of the reactor vessel head, the 

region lacked senior project engineers to devote to Davis-Besse. A vacancy 

existed for a senior project engineer responsible for Davis-Besse from June 

1997 until June 1998, except for a one month period, and from September 

1999 until May 2000, which resulted in fewer inspection hours at the facility 

than would have been normal. Other regional staff were also occupied with 

other plants in the region that were having difficulties, and NRC had 

unfilled vacancies for resident and regional inspector positions that 

strained resources for overseeing Davis-Besse. 

Even if the inspector positions had been filled, it is not certain that the 

inspectors would have aggressively followed up on any of the indications. 

According to our discussions with resident and regional inspectors and our 

on-site review of plant activities, because nuclear power plants are so large, 

with many physical structures, systems, and components, an inspector 

could miss problems that were potentially significant for safety. Licensees 

typically prepare several hundred reports per month for identifying and 

resolving problems, and NRC inspectors have only a limited amount of time 

to follow up on these licensee reports. Consequently, NRC selects and 

oversees the most safety significant structures, systems, and components.

NRC’s Assessment Process 

Did Not Indicate 

Deteriorating Performance 

Under NRC’s Reactor Oversight Process, NRC assesses licensees’ 

performance using two distinct types of information: (1) NRC’s inspection 

results and (2) performance indicators reported by the licensees. These 

indicators, which reflect various aspects of a plant’s operations, include 

data on, for example, the failure or unavailability of certain important 

operating systems, the number of unplanned power changes, and the 

amount of reactor coolant system leakage. NRC evaluates both the 

inspection results and the performance indicators to arrive at licensee 
Page 27 GAO-04-415 Davis-Besse Nuclear Power Plant

assessments, which it then color codes to reflect their safety significance. 

Green assessments indicate that performance is acceptable, and thus 

connote a very low risk significance and impact on safety. White, yellow, 

and red assessments each represent a greater degree of safety significance. 

After NRC adopted its Reactor Oversight Process in April 2000, FirstEnergy 

never received anything but green designations for its operations at DavisBesse and was viewed by NRC as a good performer until the 2002 discovery 

of the vessel head corrosion.23 Similarly, prior to adopting the Reactor 

Oversight Process, NRC consistently assessed FirstEnergy as generally 

being a good performer. NRC officials stated, however, that significant 

issues were identified and addressed as warranted throughout this period, 

such as when the agency took enforcement action in response to 

FirstEnergy’s failure to properly repair important components in 1999—a 

failure caused by weaknesses in FirstEnergy’s boric acid corrosion control 

program. 

Key Davis-Besse programs for ensuring the quality and safe operation of 

the plant’s engineered structures, systems, and components include, for 

example, 

• a corrective action program to ensure that problems at the plant that are 

relevant to safety are identified and resolved in a timely manner,

• an operating experience program to ensure that experiences or 

problems that occur are appropriately identified and analyzed to 

determine their significance and relevance to operations, and 

• a plant modification program to ensure that modifications important to 

safety are implemented in a timely manner.

As at other commercial nuclear power plants, NRC conducted routine, 

baseline inspections of Davis-Besse to determine the effectiveness of these 

programs. Reports documenting these inspections noted incidences of 

boric acid leakage, corrosion, and deposits. However, between February 

1997 and March 2000, the regional office’s assessment of the licensee’s 

performance addressed leakage in the reactor coolant system only once 

and never noted the other indications. Furthermore, Davis-Besse was not 

23Before adopting the Reactor Oversight Process, NRC also assessed licensee performance 

based on inspection results and other information; however, NRC did not assign color codes 

to assessment results.
Page 28 GAO-04-415 Davis-Besse Nuclear Power Plant

the subject of intense scrutiny in regional plant assessment meetings 

because plants perceived as good performers—such as Davis-Besse—

received substantially less attention. Between April 2000—when NRC’s 

revised assessment process took effect—until the corrosion was 

discovered in March 2002, none of NRC’s assessments of Davis-Besse’s 

performance noted leakage or other indications of corrosion at the plant. 

As a result, NRC may have missed opportunities to identify weaknesses in 

the Davis-Besse programs intended to detect or prevent the corrosion.

After the corrosion was discovered, NRC analyzed the problems that led to 

the corrosion on the reactor vessel head and concluded that FirstEnergy’s 

programs for overseeing safety at Davis-Besse were weak, as seen in the 

following examples: 

• Davis-Besse’s corrective action program did not result in timely or 

effective actions to prevent indications of leakage from reoccurring in 

the reactor coolant system. 

• FirstEnergy officials did not always enter equipment problems into the 

corrective action program because individuals who had identified the 

problem were often responsible for resolving it.

• For over a decade, FirstEnergy had delayed plant modifications to its 

service structure platform, primarily because of cost. These 

modifications would have improved its ability to inspect the reactor 

vessel head nozzles. As a result, FirstEnergy could conduct only limited 

visual inspections and cleaning of the reactor pressure vessel head 

through the small “mouse-holes” that perforated the service structure. 

NRC was also unaware of the extent to which various aspects of 

FirstEnergy’s safety culture had degraded—that is, FirstEnergy’s 

organization and performance related to ensuring safety at Davis-Besse. 

This degradation had allowed the incident to occur with no forewarning 

because NRC’s inspections and performance indicators do not directly 

assess safety culture. Safety culture is a group of characteristics and 

attitudes within an organization that establish, as an overriding priority, 

that issues affecting nuclear plant safety receive the attention their 

significance warrants. Following FirstEnergy’s March 2002 discovery, NRC 

found numerous indications that FirstEnergy emphasized production over 

plant safety. First, Davis-Besse routinely restarted the plant following an 

outage, even though reactor pressure vessel valves and control rod drive 

mechanisms leaked. Second, staff was unable to remove all of the boric 
Page 29 GAO-04-415 Davis-Besse Nuclear Power Plant

acid deposits from the reactor pressure vessel head because FirstEnergy’s 

schedule to restart the plant dictated the amount of work that could be 

performed. Third, FirstEnergy management was willing to accept degraded 

equipment, which indicated a lack of commitment to resolve issues that 

could potentially compromise safety. Fourth, Davis-Besse’s program that 

was intended to ensure that employees feel free to raise safety concerns 

without fear of retaliation had several weaknesses. For example, in one 

instance, a worker assigned to repair the containment air conditioner was 

not provided a respirator in spite of his concerns that he would inhale boric 

acid residue. According to NRC’s lessons-learned task force report, NRC 

was not aware of weaknesses in this program because its inspections did 

not adequately assess it. 

Given that FirstEnergy concluded that one of the causes for the DavisBesse incident was human performance and management failures, the 

panel overseeing FirstEnergy’s efforts to restart Davis-Besse requested that 

FirstEnergy assess its safety culture before allowing the plant to restart. To 

oversee FirstEnergy’s efforts to improve its safety culture, NRC (1) 

reviewed whether FirstEnergy had adequately identified all of the root 

causes for management and human performance failures at Davis-Besse, 

(2) assessed whether FirstEnergy had identified and implemented 

appropriate corrective actions to resolve these failures, and (3) assessed 

whether FirstEnergy’s corrective actions were effective. As late as 

February 2004, NRC had concerns about whether FirstEnergy’s actions 

would be adequate in the long term. As a result, the Davis-Besse safety 

culture was one of the issues contributing to the delay in restarting the 

plant. In March 2004, NRC’s panel concluded that FirstEnergy’s efforts to 

improve its safety culture were sufficient to allow the plant to restart. In 

doing so, however, NRC officials stated that one of the conditions the panel 

imposed was for FirstEnergy to conduct an independent assessment of the 

safety culture at Davis-Besse annually over the course of the next 5 years. 

NRC Did Not Effectively 

Incorporate Long-Standing 

Knowledge about 

Corrosion, Nozzle Cracking, 

and Leak Detection into Its 

Oversight

NRC has been aware of boric acid corrosion and its potential to affect 

safety since at least 1979. It issued several notices to the nuclear power 

industry about boric acid corrosion and, specifically, the potential for it to 

degrade the reactor coolant pressure boundary. In 1987, two licensees 

found significant corrosion on their reactor pressure vessel heads, which 

heightened NRC’s concern. A subsequent industry study concluded that 

concentrated solutions of boric acid could result in unacceptably high 

corrosion rates—up to 4 inches per year—when primary coolant leaks onto 

surfaces and concentrates at temperatures found on the surface of the 
Page 30 GAO-04-415 Davis-Besse Nuclear Power Plant

reactor vessel.24 After considering this information and several more 

instances of boric acid corrosion at plants, NRC issued a generic letter in 

1988 requesting licensees to implement boric acid corrosion control 

programs. 

In 1990, NRC visited Davis-Besse to assess the adequacy of the plant’s boric 

acid corrosion control program. At that time, NRC concluded that the 

program was acceptable. However, in 1999, NRC became aware that 

FirstEnergy’s boric acid corrosion control program was inadequate 

because boric acid had corroded several bolts on a valve, and NRC issued a 

violation. As a result of the violation, FirstEnergy agreed to review its boric 

acid corrosion procedures and enhance its program. NRC inspectors 

evaluated FirstEnergy’s completed and planned actions to improve the 

boric acid corrosion control program and found them to be adequate. 

According to NRC officials, they never inspected the remaining actions—

assuming that the planned actions had been implemented effectively. In 

2000, NRC adopted its new Reactor Oversight Process and discontinued its 

inspection procedure for plants’ corrosion control programs because these 

inspections had rarely been conducted due to higher priorities. Thus, NRC 

had no reliable or routine way to ensure that the nuclear power industry 

fully implemented boric acid corrosion control programs. 

NRC also did not routinely review operating experiences at reactors, both 

in the United States and abroad, to keep abreast of boric acid 

developments and determine the need to emphasize this problem. Indeed, 

NRC did not fully understand the circumstances in which boric acid would 

result in corrosion, rather than flash to steam. Similarly, NRC did not know 

the rate at which carbon steel would corrode under different conditions. 

This lack of knowledge may be linked to shortcomings in its program to 

review operating experiences at reactors, which could have been 

exacerbated by the 1999 elimination of the office specifically responsible 

for reviewing operating experiences.25 This office was responsible for, 

among other things, (1) coordinating operational data collection, (2) 

24Westinghouse Electric Company, Corrosion Effects of Boric Acid Leakage on Steel under 

Plant Operating Conditions—A Review of Available Data (Pittsburgh: October 1987).

25NRC’s Office for Analysis and Evaluation of Operating Data was established in response to 

a recommendation that we made to the agency in 1978 that it have a systematic process for 

analyzing operating experience and feeding this information back to licensees and the 

industry. NRC eliminated this office, and its responsibilities were transferred to other NRC 

offices in an effort to gain efficiencies.
Page 31 GAO-04-415 Davis-Besse Nuclear Power Plant

systematically analyzing and evaluating operational experience, (3) 

providing feedback on operational experience to improve safety, (4) 

assessing the effectiveness of the agencywide program, and (5) acting as a 

focal point for interaction with outside organizations on issues pertaining 

to operational safety data analysis and evaluation. According to NRC 

officials who had overseen Davis-Besse at the time of the incident, they 

would not have suspected the reactor vessel head or cracked head 

penetration nozzles as the source of the filter clogging and unidentified 

leakage because they had not been informed that these could be potential 

problems. According to these officials, the vessel head was “not on the 

radar screen.” 

With regard to nozzle cracking, NRC, for more than two decades, was 

aware of the potential for nozzles and other components made of alloy 600 

to crack. While cracks were found at nuclear power plants, NRC 

considered their safety significance to be low because the cracks were not 

developing rapidly. In contrast, other countries considered the safety 

significance of such cracks to be much higher. For example, concern over 

alloy 600 cracking led France, as a preventive measure, to institute 

requirements for an extensive nondestructive examination inspection 

program for vessel head penetration nozzles, including the removal of 

insulation, during every fuel outage. When any indications of cracking were 

observed, even more frequent inspections were required, which, because of 

economic considerations, resulted in the replacement of vessel heads when 

indications were found. The effort to replace the vessel heads is still under 

way. Japan replaced those vessel heads whose nozzles it considered most 

susceptible to cracking, even though no cracks had yet been found. Both 

France and Sweden also installed enhanced leakage monitoring systems to 

detect leaks early. However, according to NRC, such systems cannot detect 

the small amounts of leakage that may be typical from cracked nozzles.

NRC recognized that an integrated, long-term program, including periodic 

inspections and monitoring of vessel heads to check for nozzle cracking, 

was necessary. In 1997, it issued a generic letter that summarized NRC’s 

efforts to address cracking of control rod drive mechanism nozzles and 

requested information on licensees’ plans to inspect nozzles at their 

reactors. More specifically, this letter asked licensees to provide NRC with 

descriptions of their inspections of these nozzles and any plans for 

enhanced inspections to detect cracks. At that time, NRC was planning to 

review this information to determine if enhanced licensee inspections were 

warranted. Based on its review of this information, NRC concluded that the 

current inspection program was sufficient. As a result, between 1998 and 
Page 32 GAO-04-415 Davis-Besse Nuclear Power Plant

2001, NRC did not issue or solicit additional information on nozzle cracking 

or assess its requirements for inspecting reactor vessels to determine 

whether they were sufficient to detect cracks. At Davis-Besse, NRC also did 

not determine if FirstEnergy had plans or was implementing any plans for 

enhanced nozzle inspections, as noted in the 1997 generic letter. NRC took 

no further action until the cracks were found in 2001 at the Oconee plant, in 

South Carolina. NRC attributed its lack of focus on nozzle cracking, in part, 

to the agency’s inability to effectively review, assess, and follow up on 

industry operating experience events. Furthermore, as with boric acid 

corrosion, NRC did not obtain or analyze any new data about cracking that 

would have supported making changes in either its regulations or 

inspections to better identify or prevent corrosion on the vessel head at 

Davis-Besse. 

NRC’s technical specifications regarding allowable leakage rates also 

contributed to the corrosion at Davis-Besse because the amount of leakage 

that can cause extensive corrosion can be significantly less than the level 

that NRC’s specifications allow. According to NRC officials, NRC’s 

requirements, established in 1973, were based on the best available 

technology at that time. The task of measuring identified and unidentified 

leakage from the reactor coolant system is not precise. It requires licensees 

to estimate the amount of coolant that the reactor is supposed to contain 

and identify any difference in coolant levels. They then have to account for 

the estimated difference in the actual amount of coolant to arrive at a 

leakage rate; to do this, they identify all sources and amounts of leakage by, 

among other things, measuring the amount of water contained in various 

sump collection systems. If these sources do not account for the difference, 

licensees know they have an unidentified source of leakage. This estimate 

can vary significantly from day to day between negative and positive 

numbers.

According to analyses that FirstEnergy conducted after it identified the 

corrosion in March 2002, the leakage rates from the nozzle cracks were 

significantly below NRC’s reactor coolant system unidentified leakage rate 

of 1 gallon per minute. Specifically, the leakage from the nozzle around 

which the vessel head corrosion occurred was predicted to be 0.025 gallon 

per minute. If such small leakage can result in such extensive corrosion, 

identifying if and where such leakage occurs is important. NRC staff 

recognized as early as 1993 it would be prudent for the nuclear power 

industry to consider implementing an enhanced method for detecting small 

leaks during plant operation, but NRC did not require this action, and the 

industry has not taken steps to do so. Furthermore, NRC has not 
Page 33 GAO-04-415 Davis-Besse Nuclear Power Plant

consistently enforced its requirement for reactor coolant pressure 

boundary leakage. As a result, the NRC Davis-Besse task force concluded 

that inconsistent enforcement may have reinforced a belief that alloy 600 

nozzle leakage was not actually or potentially a safety significant issue. 

NRC’s Process for 

Deciding Whether to 

Allow a Delayed DavisBesse Shutdown 

Lacked Credibility

Although FirstEnergy operated Davis-Besse without incident until shutting 

it down in February 2002, certain aspects of NRC’s deliberations allowing 

the delayed shutdown raise questions about the credibility of the agency’s 

decision making, if not about the Davis-Besse decision itself. NRC does not 

have specific guidance for deciding on plant shutdowns. Instead, agency 

officials turned to guidance developed for a different purpose—reviewing 

requests to amend license operating conditions—and even then did not 

always adhere to this guidance. In addition, NRC did not document its 

decision-making process, as called for by its guidance, and its letter to 

FirstEnergy to lay out the basis for the decision—sent a year after the 

decision—did not fully explain the decision. NRC’s lack of guidance, 

coupled with the lack of documentation, precludes us from independently 

judging whether NRC’s decision was reasonable. Finally, some NRC 

officials stated that the shutdown decision was based, in part, on the 

agency’s probabilistic risk assessment (PRA) calculations of the risk that 

Davis-Besse would pose if it delayed its shutdown and inspection. 

However, as noted by our consultants, the calculations were flawed, and 

NRC’s decision makers did not always follow the agency’s guidance for 

developing and using such calculations.

NRC Did Not Have Specific 

Guidance for Deciding on 

Plant Shutdowns

NRC believed that Davis-Besse could have posed a potential safety risk 

because it was, in all likelihood, failing to comply with NRC’s technical 

specification that no leakage occur in the reactor coolant pressure 

boundary. Its belief was based on the following indicators of probable 

leakage:

• All six of the other reactors manufactured by the same company as 

Davis-Besse’s reactor had cracked nozzles and identified leakage.26

• Three of these six reactors had identified circumferential cracking.

26Davis-Besse’s manufacturer was the Babcock and Wilcox Company, which is an operating 

unit of McDermott International.
Page 34 GAO-04-415 Davis-Besse Nuclear Power Plant

• FirstEnergy had not performed a recent visual examination of all of its 

nozzles.

Furthermore, a FirstEnergy manager agreed that cracks and leakage were 

likely.

NRC has the authority to shut down a plant when it is clear that the plant is 

in violation of important safety requirements, and it is clear that the plant 

poses a risk to public health and safety.27 Thus, if a licensee is not 

complying with technical specifications, such as those for no allowable 

reactor vessel pressure boundary leakage, NRC can order a plant to shut 

down. However, NRC decided that it could not require Davis-Besse to shut 

down on the basis of other plants’ cracked nozzles and identified leakage or 

the manager’s acknowledgement of a probable leak. Instead, it believed it 

needed more direct, or absolute, proof of a leak to order a shutdown. This 

standard of proof has been questioned. According to the Union of 

Concerned Scientists,28 for example, if NRC needed irrefutable proof in 

every case of suspected problems, the agency would probably never issue a 

shutdown order. In effect, in this case NRC created a Catch-22: It needed 

irrefutable proof to order a shutdown but could not get this proof without 

shutting down the plant and requiring that the reactor be inspected. 

Despite NRC’s responsibility for ensuring that the public is adequately 

protected from accidents at commercial nuclear power plants, NRC does 

not have specific guidance for shutting down a plant when the plant may 

pose a risk to public health and safety, even though it may be complying 

with NRC requirements. It also has no specific guidance or standards for 

quality of evidence needed to determine that a plant may pose an undue 

risk. Lacking direct or absolute proof of leakage at Davis-Besse, NRC 

instead drafted a shutdown order on the basis that a potentially hazardous 

condition may have existed at the plant. NRC had no guidance for 

developing such a shutdown order, and therefore, it used its guidance for 

reviewing license amendment requests. NRC officials recognized that this 

guidance was not specifically designed to determine whether NRC should 

shut down a power plant such as Davis-Besse. However, NRC officials 

27Ordinarily, NRC would not suspend a license for a failure to meet a requirement unless the 

failure was willful and adequate corrective action had not been taken. 

28The Union of Concerned Scientists is a nonprofit partnership of scientists and citizens that 

augments scientific analyses and policy development for identifying environmental 

solutions to issues such as energy production.
Page 35 GAO-04-415 Davis-Besse Nuclear Power Plant

stated that this guidance was the best available for deciding on a shutdown 

because, although the review was not to amend a license, the factors that 

NRC needed to consider in making the decision and that were contained in 

the guidance were applicable to the Davis-Besse situation. 

To use its guidance for reviewing license amendment requests, NRC first 

determined that the situation at Davis-Besse posed a special circumstance 

because new information revealed a substantially greater potential for a 

known hazard to occur, even if Davis-Besse was in compliance with the 

technical specification for leakage from the reactor coolant pressure 

boundary. The special circumstance stemmed from NRC’s determination 

that requirements for conducting vessel head inspections were not 

sufficient to detect nozzle cracking and, thus, small leaks.29 According to 

NRC officials, this determination allowed NRC to use its guidance for 

reviewing license amendment requests when deciding whether to order a 

shutdown.

The Extent of NRC’s 

Reliance on License 

Amendment Guidance Is 

Not Clear

Under NRC’s license amendment guidance, NRC considers how the license 

change affects risk, but not how it has previously assessed licensee 

performance, such as whether the licensee was viewed as a good 

performer. With regard to the Davis-Besse decision, the guidance directed 

NRC to determine whether the plant would comply with five NRC safety 

principles if it operated beyond December 2001 without inspecting the 

reactor vessel head. As applied to Davis-Besse, these principles were 

whether the plant would (1) continue to meet requirements for vessel head 

inspections, (2) maintain sufficient defense-in-depth, (3) maintain 

sufficient safety margins, (4) have little increase in the likelihood of a core 

damage accident, and (5) monitor the vessel head and nozzles. The 

guidance, however, does not specify how to apply these safety principles, 

how NRC can demonstrate it has followed the principles and ensured they 

are met, or whether any one principle takes precedence over the others. 

The guidance also does not indicate what actions NRC or licensees should 

take if some or all of the principles are not met. 

29Specifically, reactor vessel head inspection requirements do not require that insulation be 

removed. Because of this, reactor vessel head inspections performed without removing the 

insulation above the vessel head could not result in 100 percent of the nozzles being visually 

inspected.
Page 36 GAO-04-415 Davis-Besse Nuclear Power Plant

In mid-September 2001, NRC staff concluded that Davis-Besse complied 

with the first safety principle but did not meet the remaining four. 

According to the staff, Davis-Besse did not meet three safety principles 

because the requirements for vessel head inspections were not adequate. 

Specifically, the requirements do not require the inspector to remove the 

insulation above the vessel head, and thus allow all of the nozzles to be 

visually inspected. NRC therefore could not ensure that FirstEnergy was 

maintaining defense-in-depth and adequate safety margins or sufficiently 

monitoring the vessel head and nozzles. The staff believed that Davis-Besse 

did not meet the fourth safety principle because the risk estimate of core 

damage approached an unacceptable level and the estimate itself was 

highly uncertain. 

Between early October and the end of November 2001, NRC requested and 

received additional information from FirstEnergy regarding its risk 

estimate of core damage—its PRA estimate—and met with the company to 

determine the basis for the estimate. NRC was also developing its own risk 

estimate, although its numbers kept changing. At some point during this 

time, NRC staff also concluded that the first safety principle was probably 

not being met, although the basis for this conclusion is not known.

At the end of November 2001, NRC contacted FirstEnergy and informed it 

that a shutdown order had been forwarded to the NRC commissioners and 

asked if FirstEnergy could take any actions that would persuade NRC to 

not issue the shutdown order. The following day, FirstEnergy proposed 

measures to mitigate the potential for and consequences of an accident. 

These measures included, among other things, lowering the operating 

temperature from 605 degrees Fahrenheit to 598 degrees Fahrenheit to 

reduce the driving force for stress corrosion cracking on the nozzles, 

identifying a specific operator to initiate emergency cooling in response to 

an accident, and moving the scheduled refueling outage up from March 31, 

2002, to no later than February 16, 2002. NRC staff discussed these 

measures, and NRC management asked the staff if they were concerned 

about extending Davis-Besse’s operations until mid-February 2002. While 

some of the staff were concerned about continued operations, none 

indicated to NRC management that cracking in control rod drive 

mechanism nozzles was likely extensive enough to cause a nozzle to eject 

from the vessel head, thus making it unsafe to operate. NRC formally 

accepted FirstEnergy’s compromise proposal within several days, thus 

abandoning its shutdown order. 
Page 37 GAO-04-415 Davis-Besse Nuclear Power Plant

NRC Did Not Fully Explain 

or Document the Basis for 

Its Decision

We could not fully assess NRC’s basis for accepting FirstEnergy’s proposal. 

NRC did not document its deliberations, even though its guidance requires 

that it do so. This documentation is to include the data, methods, and 

assessment criteria used; the basis for the decisions made; and essential 

correspondence sufficient to document the persons, places, and matters 

dealt with by NRC. Specifically, the guidance requires that the 

documentation contain sufficient detail to make possible a “proper 

scrutiny” of NRC decisions by authorized outside agencies and provide 

evidence of how basic decisions were formed, including oral decisions. 

NRC’s guidance also states that NRC should document all important staff 

meetings. 

In reviewing NRC’s documentation on the Davis-Besse decision, we found 

no evidence of an in-depth or formal analysis of how Davis-Besse’s 

proposed measures would affect the plant’s ability to satisfy the five safety 

principles. Thus, it is unclear whether the safety principles contained in the 

guidance were met by the measures that FirstEnergy proposed. However, 

several NRC officials stated that FirstEnergy’s proposed measures had no 

impact on plant operations or safety. For example, according to one NRC 

official, FirstEnergy’s proposal to reduce the operating temperature would 

have had little impact on safety because the small drop in operating 

temperature over a 7-week period would have had little effect on the 

growth rate of any cracks in a nozzle. As such, this official considered the 

measures as “window dressing.” A proposed measure that NRC staff did 

consider as having a significant impact on the risk was for FirstEnergy to 

dedicate an operator for manually turning on safety equipment in the event 

that a nozzle was ejected. Subsequent to approving the delayed shutdown, 

NRC learned that FirstEnergy had not, in fact, planned to dedicate an 

operator for this task—rather, FirstEnergy planned to have an operator do 

this task in addition to other regularly assigned duties. 

According to an NRC official, once NRC decided not to issue a shutdown 

order for December 2001, NRC staff needed to discuss how NRC’s 

assessment of whether the five safety principles had been met had changed 

in the course of the staff’s deliberations. However, there was no evidence in 

the agency’s records to support that this discussion was held, and other key 

meetings, such as the one in which the agency made its decision to allow 

Davis-Besse to operate past December 31, 2001, were not documented. 

Without documentation, it is not clear what factors influenced NRC’s 

decision. For example, according to the NRC Office of the Inspector 

General’s December 2002 report that examined the Davis-Besse incident, 

NRC’s decision was driven in large part by a desire to lessen the financial 
Page 38 GAO-04-415 Davis-Besse Nuclear Power Plant

impact on FirstEnergy that would result from an early shutdown.30 While 

NRC disputed this finding, we found no evidence in the agency’s records to 

support or refute its position.

In December 2001, when NRC informed FirstEnergy that it accepted the 

company’s proposed measures and the February 16, 2002, shutdown date, it 

also said that the company would receive NRC’s assessment in the near 

future. However, NRC did not provide the assessment until a full year 

later—in December 2002. In addition, the December 2002 assessment, 

which includes a four-page evaluation, does not fully explain how the 

safety principles were used or met—other than by stating that if the 

likelihood of nozzle failure were judged to be small, then adequate 

protection would be ensured. Even though NRC’s regulations regarding the 

reactor coolant pressure boundary dictate that the reactor have an 

extremely low probability of failing, NRC stated it did not believe that 

Davis-Besse needed to demonstrate strict conformance with this 

regulation. As evidence of the small likelihood of failure, NRC cited the 

small size of cracks found at other power plants, as well as its preliminary 

assessment of nozzle cracking, which projected crack growth rates. NRC 

concluded that 7 weeks of additional operation would not result in an 

appreciable increase in the size of the cracks.31 While NRC included its 

calculated estimates of the risk that Davis-Besse would pose, it did not 

detail how it calculated its estimates.

NRC’s PRA Estimate Was 

Flawed and Its Use in 

Deciding to Delay the 

Shutdown Is Unclear 

In moving forward with its more risk-informed regulatory approach, NRC 

has established a policy to increase the use of PRA methods as a means to 

promote regulatory stability and efficiency. Using PRA methods, NRC and 

the nuclear power industry can estimate the likelihood that different 

accident scenarios at nuclear power plants will result in reactor core 

damage and a release of radioactive materials. For example, one of these 

accident scenarios begins with a “medium break” loss-of-coolant accident 

in which the reactor coolant system is breached and a midsize—about 2- to 

4-inch—hole is formed that allows coolant to escape from the reactor 

30NRC, Office of the Inspector General, NRC’s Regulation of Davis-Besse Regarding 

Damage to the Reactor Vessel Head (Washington, D.C.; Dec. 30, 2002). 

31NRC, Preliminary Staff Technical Assessment for Pressurized Water Reactor Vessel 

Head Penetration Nozzles Associated with NRC Bulletin 2001-01, “Circumferential 

Cracking of Reactor Pressure Vessel Head Penetration Nozzles” (Washington, D.C.; Nov. 6, 

2001). 
Page 39 GAO-04-415 Davis-Besse Nuclear Power Plant

pressure boundary. The probability of such an accident scenario occurring 

and the consequences of that accident take into account key engineering 

safety system failure rates and human error probabilities that influence 

how well the engineered systems would be able to mitigate the 

consequences of an accident and ensure no radioactive release from the 

plant. 

For Davis-Besse, NRC needed two estimates: one for the frequency of a 

nozzle ejecting and causing a loss-of-coolant accident and one for the 

probability that a loss-of-coolant accident would result in core damage. 

NRC first established an estimate, based partially on information provided 

by FirstEnergy, for the frequency of a plant developing a cracked nozzle 

that would initiate a medium break loss-of-coolant accident. NRC 

estimated that the frequency of this occurring would be about 2x10-2, or 1 

chance in 50,32 per year. NRC then used an estimate, which FirstEnergy 

provided, for the probability of core damage given a medium break loss-ofcoolant accident. This probability estimate was 2.7x10-3, or about 1 chance 

in 370.33 Multiplying these two numbers, NRC estimated that the potential 

for a nozzle to crack and cause a loss-of-coolant accident would increase 

the frequency of core damage at Davis-Besse by about 5.4x10-5 per year, or 

about 1 in 18,500 per year.34 Converting this frequency to a probability 

associated with continued operation for 7 weeks, NRC calculated that the 

increase in the probability of core damage was approximately 5x10-6, or 1 

chance in 200,000.35 While NRC officials currently disagree that this was the 

number it used, this is the number that it included in its December 2002 

assessment provided to FirstEnergy. Further, we found no evidence in the 

agency’s records to support NRC’s current assertion. 

According to our consultants, the way NRC calculated and used the PRA 

estimate was inadequate in several respects. (See app. II for the 

consultants’ detailed report.) First, NRC’s calculations did not take into 

32Here is how to calculate the frequency estimate: 2x10-2 equates to 0.02, or 2/100, which 

equals 1/50. 

33Here is how to calculate the probability estimate: 2.7x10-3 equates to 0.0027, or 27/10,000, 

which equals 1/370.37.

34Here is how to calculate the frequency estimate: 5.4x10-5 equates to 0.000054, or 

54/1,000,000, which equals 1/18,518.52.

35Here is how to calculate the probability estimate: 5x10-6 equates to 0.000005, or 5/1,000,000, 

which equals 1/200,000.
Page 40 GAO-04-415 Davis-Besse Nuclear Power Plant

account several factors, such as the possibility of corrosion and axial 

cracking that could lead to leakage. For example, the consultants 

concluded that NRC’s estimate of risk was incorrectly too small, primarily 

because the calculation did not consider corrosion of the vessel head. In 

reviewing how NRC developed and used its PRA estimates for Davis-Besse, 

our consultants noted that the calculated risk was smaller than it should 

have been because the calculations did not consider corrosion of the 

reactor vessel from the boric acid coolant leaking through cracks in the 

nozzles. According to the consultants, apparently all NRC staff involved in 

the Davis-Besse decision were aware that coolant under high pressure was 

leaking from valves, flanges, and possibly from cracks but evidently 

thought that the coolant would immediately flash into steam and 

noncorrosive compounds of boric acid. Our consultants, however, stated 

that because boric acid could potentially cause corrosion, except at 

temperatures much higher than 600 degrees Fahrenheit, NRC should have 

anticipated that corrosion could occur. Our consultants further stated that 

as evaporation occurs, boric acid becomes more concentrated in the 

remaining liquid—making it far more corrosive—and as vapor pressure 

decreases, evaporation is further slowed. They said it should be expected 

that some of the boric acid in the escaping coolant could reach the metal 

surfaces as wet or moist, highly corrosive material underlying the surface 

layers of dry noncorrosive boric acid, which is evidently what happened at 

Davis-Besse. 

Our consultants concluded that NRC staff should have been aware of the 

experience at French nuclear power plants, where boric acid corrosion 

from leaking reactor coolant had been identified during the previous 

decade, the safety significance had been recognized, and safety procedures 

to mitigate the problem had been implemented. Furthermore, tests had 

been conducted by the nuclear power industry and in government 

laboratories on boric acid corrosion that were widely available to NRC. 

They stated that keeping abreast of safety issues at similar plants, whether 

domestic or foreign, and conveying relevant safety information to licensees 

are important functions of NRC’s safety program. According to NRC, the 

agency was aware of the experience at French nuclear power plants. For 

example, NRC concluded, in a December 15, 1994, internal NRC memo, 

that primary coolant leakage from a through-wall crack could cause boric 

acid corrosion of the vessel head. However, because it concluded that some 

analyses indicated that it would take at least 6 to 9 years before any 

corrosion would challenge the structural integrity of the head, NRC 

concluded that cracking was not a short-term safety issue. 
Page 41 GAO-04-415 Davis-Besse Nuclear Power Plant

Our consultants also stated that NRC’s risk analysis was inadequate 

because the analysis concerned only the formation and propagation of 

circumferential cracks that could result in nozzle failure, loss of coolant, 

and even control rod ejection. Although there is less chance of axial cracks 

causing complete nozzle failure, these cracks open additional pathways for 

coolant leakage. In addition, their long crevices provide considerably 

greater opportunity for the coolant to concentrate near the surface of the 

vessel head. However, according to our consultants, NRC was convinced 

that the boric acid they saw resulted from leaking flanges above the reactor 

vessel head, as opposed to axial cracks in the nozzles. 

Second, NRC’s analysis was inadequate because it did not include the 

uncertainty of its risk estimate and use the uncertainty analysis in the 

Davis-Besse decision-making process, although NRC staff should have 

recognized large uncertainties associated with its risk estimate. Our 

consultants also concluded that NRC failed to take into account the large 

uncertainties associated with estimates of the frequency of core damage 

resulting from the failure of nozzles. PRA estimates for nuclear power 

plants are subject to significant uncertainties associated with human errors 

and other common causes of system component failures, and it is 

important that proper uncertainty analyses be performed for any PRA 

study. NRC guidance and other NRC reports on advancing PRA technology 

for risk-informed decisions emphasize the need to understand and 

characterize uncertainties in PRA estimates. Our consultants stated that 

had the NRC staff estimated the margin of error or uncertainty associated 

with its PRA estimate for Davis-Besse, the uncertainty would likely have 

been so high as to render the estimate of questionable value. 

Third, NRC’s analysis was inadequate because the risk estimates were 

higher than generally considered acceptable under NRC guidance. Despite 

PRA’s important role in the decision, our consultants found that NRC did 

not follow its own guidance for ensuring that the estimated risk was within 

levels acceptable to the agency. NRC required the nuclear power industry 

to develop a baseline estimate for how frequently a core damage accident 

could occur at every nuclear power plant in the United States. This baseline 

estimate is used as a basis for deciding whether changes at a plant that 

affect the core damage frequency are acceptable. The baseline core 

damage frequency estimate for the Davis-Besse plant was between 4x10-5
Page 42 GAO-04-415 Davis-Besse Nuclear Power Plant

and 6.6x10-5 per year (which is between 1 chance in 25,00036 per year and 

about 1 chance in 15,15037 per year). NRC guidance for reviewing and 

approving license amendment requests indicates that any plant-specific 

change resulting in an increase in the frequency of core damage of 1x10-5

per year (which is 1 chance in 100,000 per year) or more would fall within 

the highest risk zone: In this case, NRC would generally not approve the 

change because the risk criterion would not be met. If a license change 

would result in a core damage frequency change of 1x10-5 per year to 1x10-6 

per year (which is 1 chance in 100,000 per year to 1 chance in 1 million per 

year), the risk criterion would be considered marginally met and NRC 

would consider approving the change but would require additional 

analysis. Finally, if a license change would result in a core damage 

frequency change of 1x10-6 per year (which is 1 chance in 1 million per 

year) or less, the risk would fall within the lowest risk zone and NRC would 

consider the risk criterion to be met and would generally consider 

approving the change without requiring additional analysis. (See fig. 6.) 

36Here is how to calculate the frequency estimate: 4x10-5 equates to 0.00004, or 4/100,000, 

which equals 1/25,000.

37Here is how to calculate the frequency estimate: 6.6x10-5 equates to 0.000066, or 

66/1,000,000, which equals 1/15,151.51.
Page 43 GAO-04-415 Davis-Besse Nuclear Power Plant

Figure 6: NRC’s Acceptance Guidelines for Core Damage Frequency

a

Risk criterion is met and license changes would generally be considered.

b

Risk criterion is considered marginally met and while license changes are generally considered, they 

require additional analysis.

c

Risk criterion is not met and license changes are generally not allowed.

However, NRC’s PRA estimate for Davis-Besse—an increase in the 

frequency of core damage of 5.4x10-5, or 1 chance in about 18,500 per 

year—was higher than the acceptable level. While an NRC official who 

helped develop the risk estimate said that additional NRC and industry 

guidance was used to evaluate whether its PRA estimate was acceptable, 

this guidance also suggests that NRC’s estimate was too high. NRC’s 

estimate of the increase in the frequency of core damage of 5.4x10-5 per 

year equates to an increase in the probability of core damage of 5x10-6 , or 1 

chance in 200,000, for the 7-week period December 31, 2001, to February 

16, 2002.35 NRC’s guidance for evaluating requests to relax NRC technical 

specifications suggests that a probability increase higher than 5x10-7, or 1 

chance in 2 million38, is considered unacceptable for relaxing the 

specifications. Thus, NRC’s estimate would not be considered acceptable 

38Here is how to calculate the probability estimate: 5x10-7 equates to 0.0000005, or 

5/10,000,000, which equals 1/2,000,000.
Page 44 GAO-04-415 Davis-Besse Nuclear Power Plant

under this guidance. NRC’s estimate would also not be considered 

acceptable under Electric Power Research Institute or Nuclear Energy 

Institute guidance unless further action were taken to evaluate or manage 

risk. According to NRC officials, NRC viewed its PRA estimate as being 

within acceptable bounds because it was a temporary situation—7 

weeks—and NRC had, at other times, allowed much higher levels of risk at 

other plants. However, at the time that NRC made its decision, it did not 

document the basis for accepting this risk estimate, even though NRC’s 

guidance explicitly states that the decision on whether PRA results are 

acceptable must be based on a full understanding of the contributors to the 

PRA results and the reasoning must be well documented. In defense of its 

decision, NRC officials said that the process they used to arrive at the 

decision is used to make about 1,500 licensing decisions such as this each 

year. 

Lastly, NRC’s analysis was inadequate because the agency does not have 

clear guidance for how PRA estimates are to be used in the decisionmaking process. Our consultants concluded that NRC’s process for riskinformed decision making is ill-defined, lacks guidelines for how it is 

supposed to work, and is not uniformly transparent within NRC. According 

to NRC officials involved in the Davis-Besse decision, NRC’s guidance is 

not clear on the use of PRA in the decision-making process. For example, 

while NRC has extensive guidance, this guidance does not outline to what 

extent or how the resultant PRA risk number and uncertainty should be 

weighed with respect to the ultimate decision. One factor complicating this 

issue is the lack of a predetermined methodology to weigh risks expressed 

in PRA numbers against traditional deterministic results and other 

factors.39 Absent this guidance, the value assigned to the PRA analysis is 

largely at the discretion of the decision maker. The process, which NRC 

stated is robust, can result in a decision in which PRA played no role, a 

partial role, or one in which it was the sole deciding factor. According to 

our consultants, this situation is made worse by the lack of guidelines for 

how, or by whom, decisions in general are made at NRC.

It is not clear how NRC staff used the PRA risk estimate in the Davis-Besse 

decision-making process. For example, according to one NRC official who 

39The deterministic approach considers a set of safety challenges and how those challenges 

should be mitigated through engineering safety margins and quality assurance standards. 

The probabilistic approach extends this by allowing for the consideration of a broader set of 

safety challenges, prioritizing safety challenges based on risk significance, and allowing for 

a broader set of mitigation mechanisms.
Page 45 GAO-04-415 Davis-Besse Nuclear Power Plant

was familiar with some of the data on nozzle cracking, these data were not 

sufficient for making a good probabilistic decision. He stated that he 

favored issuing an order requiring that Davis-Besse be shut down by the 

end of December 2001 because he believed the available data were not 

sufficient to assure a low enough probability for a nozzle to be ejected. 

Other officials indicated that they accepted FirstEnergy’s proposed 

February 16, 2002, shutdown date based largely on NRC’s PRA estimate for 

a nozzle to crack and be ejected. According to one of these officials, 

allowing the additional 7 weeks of operating time was not sufficiently risk 

significant under NRC’s guidance. He stated that safety margins at the plant 

were preserved and the PRA number was within an acceptable range. Still 

another official said he discounted the PRA estimate and did not use it at all 

when recommending that NRC accept FirstEnergy’s compromise proposal. 

This official also stated that it was likely that many of the staff did base 

their conclusions on the PRA estimate. According to our consultants, 

although the extent to which the PRA risk analysis influenced the decision 

making will probably never be known, it is apparent that it did play an 

important role in the decision to allow the shutdown delay. 

NRC Has Made 

Progress in 

Implementing 

Recommended 

Changes, but Is Not 

Addressing Important 

Systemic Issues

NRC has made significant progress in implementing the actions 

recommended by the Davis-Besse lessons-learned task force. While NRC 

has implemented slightly less than half—21 of the 51—recommendations 

as of March 2004, it is scheduled to have more than 70 percent of them 

implemented by the end of 2004. For example, NRC has already taken 

actions to improve staff training and inspections that would appear to help 

address the concern that NRC inspectors viewed FirstEnergy as a good 

performer and thus did not subject Davis-Besse to the level of scrutiny or 

questioning that they should have. It is not certain when actions to 

implement the remaining recommendations will occur, in part because of 

resource constraints. NRC also faces challenges in fully implementing the 

recommendations, also in part because of resource constraints, both in the 

staff needed to develop specific corrective actions and in the additional 

staff responsibilities and duties to carry them out. Further, while NRC is 

making progress, the agency is not addressing three systemic issues 

highlighted by the Davis-Besse experience: (1) an inability to detect 

weakness or deterioration in FirstEnergy’s safety culture, (2) deficiencies 

in NRC’s process for deciding on a shutdown, and (3) lack of management 

controls to track, on a longer-term basis, the effectiveness of actions 

implemented in response to incidents such as Davis-Besse, so that they do 

not occur at another power plant.
Page 46 GAO-04-415 Davis-Besse Nuclear Power Plant

NRC Does Not Expect to 

Complete Its Actions until 

2006, in Part Because of 

Resource Constraints

NRC’s lessons-learned task force for Davis-Besse developed 51 

recommendations to address the weaknesses that contributed to the DavisBesse incident. Of these 51 recommendations, NRC rejected 2 because it 

concluded that agency processes or procedures already provided for the 

recommendations’ intent to be effectively carried out.40 To address the 

remaining 49 recommendations, NRC developed a plan in March 2003 that 

included, for each recommendation, the actions to be taken, the 

responsible NRC office, and the schedule for completing the actions. When 

developing its schedule, NRC placed the highest priority on implementing 

recommendations that were most directly related to the underlying causes 

of the Davis-Besse incident as well as those recommendations responding 

to vessel head corrosion. NRC assigned a lower priority to the remaining 

recommendations, which were to be integrated into the planning activities 

of those NRC offices assigned responsibility for taking action on the 

recommendations. In assigning these differing priorities, NRC officials 

stated they recognized that the agency has many other pressing matters to 

address that are not related to the Davis-Besse incident, such as renewing 

operating licenses, and they did not want to divert resources away from 

these activities. (App. III contains a complete list of the task force’s 

recommendations, NRC actions, and the status of the recommendations as 

of March 2004.)

To better track the status of the agency’s actions to implement the 

recommendations, we split two of the 49 recommendations that NRC 

accepted into 4; therefore, our analysis reflects NRC’s response to 51 

recommendations. As shown in table 1, as of March 2004, NRC had made 

progress in implementing the recommendations, although some 

completion dates have slipped. 

40These two recommendations were for NRC to (1) review how industry considers 

economic factors in making decisions to repair equipment and consider these factors in 

developing guidance for nonvisual inspections of vessel head penetration nozzles, and (2) 

revise the criteria for reviewing industry topical reports that have not been formally 

submitted to NRC for review but that have generic safety implications.
Page 47 GAO-04-415 Davis-Besse Nuclear Power Plant

Table 1: Status of Davis-Besse Lessons-Learned Task Force Recommendations, as 

of March 2004

Source: GAO analysis of NRC data.

Note: This table does not include the two recommendations NRC rejected.

As the table shows, as of March 2004, NRC had implemented 21 

recommendations and scheduled another 17 for completion by December 

2004. However, some slippage has already occurred in this schedule—

primarily because of resource constraints—and NRC has rescheduled 

completion of some recommendations. NRC’s time frames for completing 

the recommendations depend on several factors—the recommendations’ 

priority, the amount of work required to develop and implement actions, 

and the need to first complete actions on other related recommendations. 

Of the 21 implemented recommendations, 10 called upon NRC to revise or 

enhance its inspection guidance or training. For example, NRC revised the 

guidance it uses to assess the implementation of licensees’ programs to 

identify and resolve problems before they affect operations. It took this 

action because the task force had concluded that FirstEnergy’s weak 

corrective action program implementation was a major contributor to the 

Davis-Besse incident. NRC has also developed Web-based training modules 

to improve NRC inspectors’ knowledge of boric acid corrosion and nozzle 

cracking. The other 11 completed recommendations concerned actions 

such as 

• collecting and analyzing foreign and domestic information on alloy 600 

nozzle cracking, 

• fully implementing and revising guidance to better assure that licensees 

carry out their commitments to make operational changes, and 

• establishing measurements for resident inspector staffing levels and 

requirements.

Status 

Number of 

recommendations

Completed as of March 2004 21

Scheduled for completion April through December 2004 17

Scheduled for completion in 2005 6

Completion date yet to be determined 7

Total 51
Page 48 GAO-04-415 Davis-Besse Nuclear Power Plant

By the end of 2004, NRC expects to complete another 17 recommendations, 

12 of which generally address broad oversight or programmatic issues, and 

5 of which provide for additional inspection guidance and training. On the 

broader issues, for example, NRC is scheduled to complete a review of the 

effectiveness of its response to past NRC lessons-learned task force reports 

by April 2004. By December 2004, NRC expects to have a framework 

established for moving forward with implementing recommended 

improvements to its agencywide operating experience program. 

In 2005, 4 of the 6 recommendations scheduled for completion concern 

leakage from the reactor coolant system. For example, NRC is to (1) 

develop guidance and criteria for assessing licensees’ responses to 

increasing leakage levels and (2) determine whether licensees should 

install enhanced systems to detect leakage from the reactor coolant 

system. The fifth recommendation calls for NRC to inspect the adequacy of 

licensees’ programs for controlling boric acid corrosion, and the final 

recommendation calls on NRC to assess the basis for canceling a series of 

inspection procedures in 2001. 

NRC did not assign completion dates to 7 recommendations because, 

among other things, their completion depends on completing other 

recommendations or because of limited resources. Even though it has not 

assigned completion dates for these recommendations, NRC has begun to 

work on 5 of the 7: 

• Two recommendations will be addressed when requirements for vessel 

head inspections are revised. To date, NRC has taken some related, but 

temporary, actions. For example, since February 2003, it has required 

licensees to more extensively examine their reactor vessel heads. NRC 

has also issued a series of temporary instructions for NRC inspectors to 

oversee the enhanced examinations. NRC expects to replace these 

temporary steps with revised requirements for vessel head inspections.

• Two recommendations call upon NRC to revise requirements for 

detecting leaks in the reactor coolant pressure boundary. In response, 

NRC has, for example, begun to review its barrier integrity requirements 

and has contracted for research on enhanced detection capabilities.

• One recommendation is directed at improving follow-up of licensee 

actions taken in response to NRC generic communications. NRC is 

currently developing a temporary inspection procedure to assess the 

effectiveness of licensee actions taken in response to generic 
Page 49 GAO-04-415 Davis-Besse Nuclear Power Plant

communications. Additionally, as a long-term change in the operating 

experience program, the agency plans to improve the verification of 

how effective its generic communications are.

The remaining two recommendations address NRC’s need to (1) evaluate 

the adequacy of methods for analyzing the risks posed by passive 

components, such as reactor vessels, and integrate these methods and risks 

into NRC’s decision-making process and (2) review a sample of plant 

assessments conducted between 1998 and 2000 to determine if any 

identified plant safety issues have not been adequately assessed. NRC has 

not yet taken action on these recommendations. 

Some recommendations will require substantial resources to develop and 

implement. As a result, some implementation dates have slipped and some 

plans in response to the recommendations have changed in scope. For 

example, owing to resource constraints, NRC has postponed indefinitely 

the evaluation of methods to analyze the risk associated with passive 

reactor components such as the vessel head. Also, in part due to resource 

constraints, NRC has reconceptualized its plan to review licensee actions 

in response to previous generic communications, such as bulletins and 

letters. 

Staff resources will be strained because implementing the 

recommendations adds additional responsibilities or duties—that is, more 

inspections, training, and reviews of licensee reports. For example, NRC’s 

revised inspection guidance for more thorough examinations of reactor 

vessel heads and nozzles, as well as new requirements for NRC oversight of 

licensees’ corrective action programs, will require at least an additional 200 

hours of inspection per reactor per year. As of February 2004, NRC was 

also revising other inspection requirements that are likely to place 

additional demands on inspectors’ time. Thus, to respond to these 

increased demands, NRC will either need to add inspectors or reduce 

oversight of other licensee activities. 

To its credit, in its 2004 budget plan, NRC increased the level of resources 

for some inspection activities. However, it is not certain that these 

increases will be maintained. The number of inspection hours has fallen by 

more than one-third between 1995 and 2001. In addition, NRC is aware that 

resident inspector vacancies are filled with staff having varying levels of 

experience—from the basic level that would be expected from a newly 

qualified inspector to the advanced level that is achieved after several 

years’ experience. According to the latest available data, as of May 2003, 
Page 50 GAO-04-415 Davis-Besse Nuclear Power Plant

about 12 percent of sites had only one resident inspector; the remaining 88 

percent had two inspectors of varying levels of experience. Because of this 

situation, NRC augments these inspection resources with regional 

inspectors and contractors to ensure that, at a minimum, its baseline 

inspection program can be implemented throughout the year. Because of 

surges in the demand for inspections, NRC in 2003 increased its use of 

contractors and temporarily pulled qualified inspectors from other jobs to 

help complete the baseline inspection program for every plant. According 

to NRC, it did not expect to require such measures in 2004.

Similarly, NRC may require additional staff to identify and evaluate plants’ 

operating experiences and communicate the results to licensees, as the 

task force recommended. NRC has currently budgeted an increase of three 

full-time staff in fiscal year 2006 to implement a centralized system, or 

clearinghouse, for managing the operating experience program. However, 

according to an NRC official, questions remain about the level of resources 

needed to fully implement the task force recommendations. NRC’s 

operating experience office, before it was disbanded in 1999, had about 33 

staff whose primary responsibility was to collect, evaluate, and 

communicate activities associated with safety performance trends, as 

reflected in licensees’ operating experiences, and participate in developing 

rulemakings. However, it is too early to know the effectiveness of this 

clearinghouse approach and the adequacy of resources in the other offices 

available for collecting and analyzing operating experience information. 

Neither the operating experience office before it was disbanded nor the 

other offices flagged boric acid corrosion, cracking, or leakage as problems 

warranting significantly greater oversight by NRC, licensees, or the nuclear 

power industry. 

NRC Has Not Proposed Any 

Specific Actions to Correct 

Systemic Weaknesses in 

Oversight and DecisionMaking Processes

NRC’s Davis-Besse task force did not make any recommendations to 

address two systemic problems: evaluating licensees’ commitment to 

safety and improving the agency’s process for deciding on a shutdown.

NRC’s Task Force 

Recommendations Did Not 

Address Licensee Safety Culture

NRC’s task force identified numerous problems at Davis-Besse that 

indicated human performance and management failures and concluded 

that FirstEnergy did not foster an environment that was fully conducive to 

ensuring that plant safety issues received appropriate attention. Although 
Page 51 GAO-04-415 Davis-Besse Nuclear Power Plant

the task force report did not use the term safety culture, as evidence of 

FirstEnergy’s safety culture problems, the task force pointed to 

• an imbalance between production and safety, as evidenced by 

FirstEnergy’s efforts to address symptoms (such as regular cleanup of 

boric acid deposits) rather than causes (finding the source of the leaks 

during refueling outages);

• a lack of management involvement in or oversight of work at DavisBesse that was important for maintaining safety;

• a lack of a questioning attitude by senior FirstEnergy managers with 

regard to vessel head inspections and cleaning activities;

• ineffective and untimely corrective action;

• a long-standing acceptance of degraded equipment; and

• inadequate engineering rigor. 

The task force concluded that NRC’s implementation of guidance for 

inspecting and assessing a safety-conscious work environment and 

employee concerns programs failed to identify significant safety problems. 

Although the task force did not make any specific recommendations that 

NRC develop a means to assess licensees’ safety culture, it did recommend 

changes to focus more effort on assessing programs to promote a safetyconscious work environment. 

NRC has taken little direct action in response to this task force 

recommendation. However, to help enhance NRC’s capability to assess 

licensee safety culture by indirect means, NRC modified the wording in, 

and revised its inspection procedure for, assessing licensees’ ability to 

identify and resolve problems, such as malfunctioning plant equipment. 

These revisions included requiring inspectors to 

• review all licensee reports on plant conditions,

• analyze trends in plant conditions to determine the existence of 

potentially significant safety issues, and

• expand the scope of their reviews to the prior 5 years in order to identify 

recurring issues. 
Page 52 GAO-04-415 Davis-Besse Nuclear Power Plant

This problem identification and resolution inspection procedure is 

intended to assess the end results of management’s safety commitment 

rather than the commitment itself. However, by measuring only the end 

results, early signs of a deteriorating safety culture and declining 

management performance may not be readily visible and may be hard to 

interpret until clear violations of NRC’s regulations occur. Furthermore, 

because NRC directs its inspections at problems that it recognizes as being 

more important to safety, NRC may overlook other problems until they 

develop into significant and immediate safety problems. Conditions at a 

plant can quickly degrade to the extent that they can compromise public 

health and safety. 

The International Atomic Energy Agency and its member nations have 

developed guidance and procedures for assessing safety culture at nuclear 

power plants, and today several countries, such as Brazil, Canada, Finland, 

Sweden, and the United Kingdom, assess plant safety culture or licensees’ 

own assessments of their safety culture.41 In assessing safety culture, an 

advisory group to the agency suggests that regulatory agencies examine 

whether, for example, (1) employee workloads are not excessive, (2) staff 

training is sufficient, (3) responsibility for safety has been clearly assigned 

within the organization, (4) the corporation has clearly communicated its 

safety policy, and (5) managers sufficiently emphasize safety during plant 

meetings. One reason for assessing safety culture, according to the 

Canadian Nuclear Safety Commission, is because management and human 

performance aspects are among the leading causes of unplanned events at 

licensed nuclear facilities, particularly in light of pressures such as 

deregulation of the electricity market. Finland specifically requires that 

nuclear power plants maintain an advanced safety culture and its 

inspections target the importance that has been embedded in factors 

affecting safety, including management. NRC had begun considering 

methods for assessing organizational factors, including safety culture, but 

in 1998, NRC’s commissioners decided that the agency should have a 

performance-based inspection program of overall plant performance and 

should infer licensee management performance and competency from the 

results of that program. They chose this approach instead of one of four 

other options: 

41The International Atomic Energy Agency is an international organization affiliated with the 

United Nations that provides advice and assistance to its members on nuclear safety 

matters.
Page 53 GAO-04-415 Davis-Besse Nuclear Power Plant

• conduct performance-based inspections in all areas of facility operation 

and design, but not infer or articulate conclusions regarding the 

performance of licensee management;

• assess the performance of licensee management through targeted 

operations-based inspections using specific inspection procedures, 

trained staff, and contractors to assess licensee management—a task 

that would require the development of inspection procedures and 

significant training—and to document inspection results;

• assess the performance of licensee management as part of the routine 

inspection program by specifically evaluating and documenting 

management performance attributes—a larger effort that would require 

the development of assessment tools to evaluate safety culture as well 

as additional resources; or 

• assess the competency of licensee management by evaluating 

management competency attributes—an even larger effort that would 

require that implementation options and their impacts be assessed.

When adopting the proposal to infer licensee management performance 

from the results of its performance-based inspection program, NRC 

eliminated any resource expenditures specifically directed at developing a 

systematic method of inferring management performance and competency. 

NRC stated that it currently has a number of means to assess safety culture 

that provide indirect insights into licensee safety culture. These means 

include, for example, (1) insights from augmented inspection teams, (2) 

lessons-learned reviews, and (3) information obtained in the course of 

conducting inspections under the Reactor Oversight Process. However, 

insights from augmented inspection teams and lessons-learned reviews are 

reactionary and do not prevent problems such as those that occurred at 

Davis-Besse. Further, before the Davis-Besse incident, NRC assumed its 

oversight process would adequately identify problems with licensees’ 

safety culture. However, NRC has no formalized process for collectively 

assessing information obtained in the course of its problem identification 

and resolution inspection to ensure that individual inspection results would 

identify poor management performance. NRC stated that its licensee 

assessments consider inputs such as inspection results and insights, 

correspondence to licensees related to inspection observations, input from 

resident inspectors, and the results of any special investigations. However, 

this information may not be sufficient to inform NRC of problems at a plant 

in advance of these problems becoming safety significant. 
Page 54 GAO-04-415 Davis-Besse Nuclear Power Plant

In part because of Davis-Besse, NRC’s Advisory Committee on Reactor 

Safeguards42 recommended that NRC again pursue the development of a 

methodology for assessing safety culture. It also asked NRC to consider 

expanding research to identify leading indicators of degradation in human 

performance and work to develop a consistent comprehensive 

methodology for quantifying human performance. During an October 2003 

public meeting of the advisory committee’s Human Performance 

Subcommittee, the subcommittee’s members again reiterated the need for 

NRC to assess safety culture. Specifically, the members recognized that 

certain aspects of safety culture, such as beliefs, perceptions, and 

management philosophies, are ultimately the nuclear power industry’s 

responsibility but stated that NRC should deal with patterns of behavior 

and human performance, as well as organizational structures and 

processes. At this meeting, NRC officials discussed potential safety culture 

indicators that NRC could use, including, among other things, how many 

times a problem recurs at a plant, timeliness in correcting problems, 

number of temporary modifications, and individual program and process 

error rates. Committee members recommended that NRC test various 

safety culture indicators to determine whether (1) such indicators should 

ultimately be incorporated into the Reactor Oversight Process and (2) a 

significance determination process could be developed for safety culture. 

As of March 2004, NRC had yet to respond to the advisory committee’s 

recommendation. 

Despite the lack of action to address safety culture issues, NRC’s concern 

over FirstEnergy’s safety culture at Davis-Besse was one of the last issues 

resolved before the agency approved Davis-Besse’s restart. NRC undertook 

a series of inspections to examine Davis-Besse’s safety culture and 

determine whether FirstEnergy had (1) correctly identified the underlying 

causes associated with its declining safety culture, (2) implemented 

appropriate actions to correct safety culture problems, and (3) developed a 

process for monitoring to ensure that actions taken were effective for 

resolving safety culture problems. In December 2003, NRC noted 

significant improvements in the safety culture at Davis-Besse, but 

expressed concern with the sustainability of Davis-Besse’s performance in 

this area. For example, a survey of FirstEnergy and contract employees 

conducted by FirstEnergy in November 2003 indicated that about 17 

42The Advisory Committee on Reactor Safeguards is an independent committee comprising 

nuclear experts that advises NRC on matters of licensing and safety-related issues, and 

provides technical advice to aid the NRC commissioners’ decision-making process. 
Page 55 GAO-04-415 Davis-Besse Nuclear Power Plant

percent of employees believed that management cared more about cost 

and schedule than resolving safety and quality issues—again, production 

over safety.

NRC’s Task Force 

Recommendations Did Not 

Address NRC’s Decision-Making 

Process

NRC’s task force also did not analyze NRC’s process for deciding not to 

order a shutdown of the Davis-Besse plant. It noted that NRC’s written 

rationale for accepting FirstEnergy’s justification for continued plant 

operation had not yet been prepared and recommended that NRC change 

guidance requiring NRC to adequately document such decisions. It also 

made a recommendation to strengthen guidance for verifying information 

provided by licensees. According to an NRC official on the task force, the 

task force did not assess the decision-making process in detail because the 

task force was charged with determining why the degradation at DavisBesse was not prevented and because NRC had coordinated with NRC’s 

Office of the Inspector General, which was reviewing NRC’s decision 

making. 

NRC’s Failure to Track the 

Resolution of Identified 

Problems May Allow the 

Problems to Recur 

The NRC task force conducted a preliminary review of prior lessonslearned task force reports to determine whether they suggested any 

recurring or similar problems. As a result of this preliminary review, the 

task force recommended that a more detailed review be conducted to 

determine if actions that NRC took as a result of those reviews were 

effective. These previous task force reports included: Indian Point 2 in 

Buchanan, New York, in February 2000; Millstone in Waterford, 

Connecticut, in October 1993; and South Texas Project in Wadsworth, 

Texas, from 1988 to 1994.43 NRC’s more detailed review, as of May 2004, was 

still under way. We also reviewed these reports to determine whether they 

suggested any recurring problems and found that they highlighted broad 

areas of continuing programmatic weaknesses, as seen in the following 

examples: 

• Inspector training and information sharing. All three of the other task 

forces also identified inspector training issues and problems with 

information collection and sharing. The Indian Point task force called 

43NRC formed the Indian Point lessons-learned task force in response to a steam-generatortube rupture that forced a reactor shutdown. NRC formed the Millstone lessons-learned task 

force because the plant operated outside its design standards while refueling. NRC formed 

the South Texas task force in response to concerns about the effectiveness of NRC’s 

inspection program and the adequacy of the licensee’s employee concerns program.
Page 56 GAO-04-415 Davis-Besse Nuclear Power Plant

upon NRC to develop a process for promptly disseminating technical 

information to NRC inspectors so that they can review and apply the 

information in their inspection program.

• Oversight of licensee corrective action programs. Two of the three task 

forces also identified inadequate oversight of licensee corrective action 

programs. The South Texas task force recommended improving 

assessments of licensees’ corrective action programs to ensure that 

NRC identifies broader licensee problems.

• Better identification of problems. Two of the three task force reports 

also noted the need for NRC to develop a better process for identifying 

problem plants, and one report noted the need for NRC inspectors to 

more aggressively question licensees’ activities. 

Over the past two decades, we have also reported on underlying causes 

similar to those that contributed, in part, to the incident at Davis-Besse. 

(See Related GAO Products.) For example, with respect to the safety 

culture at nuclear power plants, in 1986, 1995, and 1997, we reported on 

issues relevant to NRC assessing plant management so that significant 

problems could be detected and corrected before they led to incidents such 

as the one that later occurred at Davis-Besse. Regardless of our 1997 

recommendation that NRC require that the assessment of management’s 

competency and performance be a mandatory component of NRC’s 

inspection process, NRC subsequently withdrew funding to accomplish 

this. In terms of inspections, in 1995 we reported that NRC, itself, had 

concluded that the agency was not effectively integrating information on 

previously identified and long-standing issues to determine if the issues 

indicated systemic weaknesses in plant operations. This report further 

noted that NRC was not using such information to focus future inspection 

activities. In 1997 and 2001, we reported on weaknesses in NRC’s 

inspections of licensees’ corrective action programs. Finally, with respect 

to learning from plants’ operating experiences, in 1984 we noted that NRC 

needed to improve its methods for consolidating information so that it 

could evaluate safety trends and ensure that generic issues are resolved at 

individual plants. These recurring issues indicate that NRC’s actions, in 

response to individual plant incidents and recommendations to improve 

oversight, are not always institutionalized. 

NRC guidance requires that resolutions to action plans be described and 

documented, and while NRC is monitoring the status of actions taken in 

response to Davis-Besse task force recommendations and preparing 
Page 57 GAO-04-415 Davis-Besse Nuclear Power Plant

quarterly and semiannual reports on the status of actions taken, the DavisBesse action plan does not specify how long NRC will monitor them. It also 

does not describe how long NRC will prepare quarterly and semiannual 

status reports, even though, according to NRC officials, these semiannual 

status reports will continue until all items are completed and the agency is 

required to issue a final summary report. The plan also does not specify 

what criteria the agency will use to determine when the actions in response 

to specific task force recommendations are completed. Furthermore, 

NRC’s action plan does not require NRC to assess the long-term 

effectiveness of recommended actions, even though, according to NRC 

officials, some activities already have an effectiveness review included. As 

in the past and in response to prior lessons-learned task force reports and 

recommendations, NRC has no management control in place for assessing 

the long-term effectiveness of efforts resulting from the recommendations. 

NRC officials acknowledged the need for a management control, such as an 

agencywide tracking system, to ensure that actions taken in response to 

task force recommendations effectively resolve the underlying issue over 

the long term, but the officials have no plans to establish such a system. 

Conclusions It is unlikely, given the actions that NRC has taken to date, that extensive 

reactor vessel corrosion will occur any time soon at another domestic 

nuclear power plant. However, we do not yet have adequate assurances 

from NRC that many of the factors that contributed to the incident at DavisBesse will be fully addressed. These factors include NRC’s failure to keep 

abreast of safety significant issues by collecting information on operating 

experiences at plants, assessing their relative safety significance, and 

effectively communicating information within the agency to ensure that 

oversight is fully informed. The underlying causes of the Davis-Besse 

incident underscore the potential for another incident unrelated to boric 

acid corrosion or cracked control rod drive mechanism nozzles to occur. 

This potential is reinforced by the fact that both prior NRC lessons-learned 

task forces and we have found similar weaknesses in many of the same 

NRC programs that led to the Davis-Besse incident. NRC has not followed 

up on prior task force recommendations to assess whether the lessons 

learned were institutionalized. NRC’s actions to implement the Davis-Besse 

lessons-learned task force recommendations, to be fully effective, will 

require an extensive effort on NRC’s part to ensure that these are 

effectively incorporated into the agency’s processes. However, NRC has not 

estimated the amount of resources necessary to carry out these 

recommendations, and we are concerned that resource limitations could 

constrain their effectiveness. For this reason, it is important for NRC to not 
Page 58 GAO-04-415 Davis-Besse Nuclear Power Plant

only monitor the implementation of Davis-Besse task force 

recommendations, but also determine their effectiveness, in the long term, 

and the impact that resource constraints may have on them. These actions 

are even more important because the nation’s fleet of nuclear power plants 

is aging.

Because the Davis-Besse task force did not address NRC’s unwillingness to 

directly assess licensee safety culture, we are concerned that NRC’s 

oversight will continue to be reactive rather than proactive. NRC’s 

oversight can result in NRC making a determination that a licensee’s 

performance is good one day, yet the next day NRC discovers the 

performance to be unacceptably risky to public health and safety. Such a 

situation does not occur overnight: Long-standing action or inaction on the 

part of the licensee causes unacceptably risky and degraded conditions. 

NRC needs better information to preclude such conditions. Given the 

complexity of nuclear power plants, the number of physical structures, 

systems, and components, and the manner in which NRC inspectors must 

sample to assess whether licensees are complying with NRC requirements 

and license specifications, it is possible that NRC will not identify licensees 

that value production over safety. While we recognize the difficulty in 

assessing licensee safety culture, we believe it is sufficiently important to 

develop a means to do so. 

Given the limited information NRC had at the time and that an accident did 

not occur during the delay in Davis-Besse’s shutdown, we do not 

necessarily question the decision the agency made. However, we are 

concerned about NRC’s process for making that decision. It used guidance 

intended to make decisions for another purpose, did not rigorously apply 

the guidance, established an unrealistically high standard of evidence to 

issue a shutdown order, relied on incomplete and faulty PRA analyses and 

licensee evidence, and did not document key decisions and data. It is 

extremely unusual for NRC to order a nuclear power plant to shut down. 

Given this fact, it is more imperative that NRC have guidance to use when 

technical specifications or requirements may be met, yet questions arise 

over whether sufficient safety is being maintained. This guidance does not 

need to be a risk-based approach, but rather a more structured riskinformed approach that is sufficiently flexible to ensure that the guidance 

is applicable under different circumstances. This is important because NRC 

annually makes about 1,500 licensing decisions relating to operating 

commercial nuclear power plants. While we recognize the challenges NRC 

will face in developing such guidance, the large number and wide variety of 
Page 59 GAO-04-415 Davis-Besse Nuclear Power Plant

decisions strongly highlight the need for NRC to ensure that its decisionmaking process and decisions are sound and defensible. 

Recommendations for 

Executive Action

To ensure that NRC aggressively and comprehensively addresses the 

weaknesses that contributed to the Davis-Besse incident and could 

contribute to problems at nuclear power plants in the future, we are 

recommending that the NRC commissioners take the following five 

actions:

• Determine the resource implications of the task force’s 

recommendations and reallocate the agency’s resources, as appropriate, 

to better ensure that NRC effectively implements the recommendations. 

• Develop a management control approach to track, on a long-term basis, 

implementation of the recommendations made by the Davis-Besse 

lessons-learned task force and future task forces. This approach, at a 

minimum, should assign accountability for implementing each 

recommendation and include information on the status of major actions, 

how each recommendation will be judged as completed, and how its 

effectiveness will be assessed. The approach should also provide for 

regular—quarterly or semiannual—reports to the NRC commissioners 

on the status of and obstacles to full implementation of the 

recommendations. 

• Develop a methodology to assess licensees’ safety culture that includes 

indicators of and inspection information on patterns of licensee 

performance, as well as on licensees’ organization and processes. NRC 

should collect and analyze this data either during the course of the 

agency’s routine inspection program or during separate targeted 

assessments, or during both routine and targeted inspections and 

assessments, to provide an early warning of deteriorating or declining 

performance and future safety problems. 

• Develop specific guidance and a well-defined process for deciding on 

when to shut down a nuclear power plant. The guidance should clearly 

set out the process to be used, the safety-related factors to be 

considered, the weight that should be assigned to each factor, and the 

standards for judging the quality of the evidence considered. 

• Improve NRC’s use of probabilistic risk assessment estimates in 

decision making by (1) ensuring that the risk estimates, uncertainties, 
Page 60 GAO-04-415 Davis-Besse Nuclear Power Plant

and assumptions made in developing the estimates are fully defined, 

documented, and communicated to NRC decision makers; and (2) 

providing guidance to decision makers on how to consider the relative 

importance, validity, and reliability of quantitative risk estimates in 

conjunction with other qualitative safety-related factors. 

Agency Comments and 

Our Evaluation

We provided a draft of this report to NRC for review and comment. We 

received written comments from the agency’s Executive Director for 

Operations. In its written comments, NRC generally addressed only those 

findings and recommendations with which it disagreed. Although 

commenting that it agreed with many of the report’s findings, NRC 

expressed an overall concern that the report does not appropriately 

characterize or provide a balanced perspective on NRC’s actions 

surrounding the discovery of the Davis-Besse reactor vessel head condition 

or NRC’s actions to incorporate the lessons learned from that experience 

into its processes. Specifically, NRC stated that the report does not 

acknowledge that NRC must rely heavily on its licensees to provide it with 

complete and accurate information, as required by its regulations. NRC 

also expressed concern about the report’s characterization of its use of risk 

estimates—specifically the report’s statement that NRC’s estimate of risk 

exceeded the risk levels generally accepted by the agency. In addition, NRC 

disagreed with two of our recommendations: (1) to develop specific 

guidance and a well-defined process for deciding on when to shut down a 

plant and (2) to develop a methodology to assess licensees’ safety culture.

With respect to NRC’s overall concern, we believe that the report 

accurately captures NRC’s performance. Our draft report, in discussing 

NRC’s regulatory and oversight role and responsibilities, stated that 

according to NRC, the completeness and accuracy of the information 

provided by licensees is an important aspect of the agency’s oversight. To 

respond further to NRC’s concern, we added a statement to the effect that 

licensees are required under NRC’s regulations to provide the agency with 

complete and accurate information. While we do not want to diminish the 

importance of this responsibility on the part of the licensees, we believe 

that NRC also has a responsibility, in designing its oversight program, to 

implement management controls, including inspection and enforcement, to 

ensure that it has accurate information on and is sufficiently aware of plant 

conditions. In this respect, it was NRC’s decision to rely on the premise that 

the information provided by FirstEnergy was complete and accurate. As we 

point out in the report, the degradation of the vessel head at Davis-Besse 

occurred over several years. NRC knew about several indications that 
Page 61 GAO-04-415 Davis-Besse Nuclear Power Plant

problems were occurring at the plant, and the agency could have requested 

and obtained additional information about the vessel head condition. 

We also believe that the report’s characterization of NRC’s use of risk 

estimates is accurate. The NRC risk estimate that we and our consultants 

found for the period leading up to the December 2001 decision on DavisBesse’s shutdown, including the risk estimate used by the staff during key 

briefings of NRC management, indicated that the estimate for core damage 

frequency was 5.4x10-5, as used in the report. The 5x10-6 referenced in 

NRC’s December 2002 safety evaluation is for core damage probability, 

which equates to a core damage frequency of approximately 5x10-5—a level 

that is in excess of the level generally accepted by the agency. The 

impression of our consultants is that some confusion about the differences 

in these terms may exist among NRC staff.

Concerning NRC’s disagreement with our recommendation to develop 

specific guidance for making plant shutdown decisions, NRC stated that its 

regulations, guidance, and processes are robust and do provide sufficient 

guidance in the vast majority of situations. The agency added that from 

time to time a unique situation may present itself wherein sufficient 

information may not exist or the information available may not be 

sufficiently clear to apply existing rules and regulations definitively. 

According to NRC, in these unique instances, the agency’s most senior 

managers, after consultation with staff experts and given all of the 

information available at the time, decide whether to require a plant 

shutdown. While we agree that NRC has an array of guidance for making 

decisions, we continue to believe that NRC needs specific guidance and a 

well-defined process for deciding when to shut down a plant. As discussed 

in our report, the agency used its guidance for approving license change 

requests to make the decision on when to shut down Davis-Besse. Although 

NRC’s array of guidance provides flexibility, we do not believe that it 

provides the structure, direction, and accountability needed for important 

decisions such as the one on Davis-Besse’s shutdown.

In disagreeing with our recommendation concerning the need for a 

methodology to assess licensees’ safety culture, NRC said that the 

Commission, to date, has specifically decided not to conduct direct 

evaluations or inspections of safety culture as a routine part of assessing 

licensee performance due to the subjective nature of such evaluations. 

According to NRC, as regulators, agency officials are not charged with 

managing licensees’ facilities, and direct involvement with organizational 

structure and processes crosses over to a management function. We 
Page 62 GAO-04-415 Davis-Besse Nuclear Power Plant

understand NRC’s position that it is not charged with managing licensees’ 

facilities, and we are not suggesting that NRC should prescribe or regulate 

the licensees’ organizational structure or processes. Our recommendation 

is aimed at NRC monitoring trends in licensees’ safety culture as an early 

warning of declining performance and safety problems. Such early 

warnings can help preclude NRC from assessing a licensee as being a good 

performer one day, and the next day being faced with a situation that it 

considers a potentially significant safety risk. As discussed in the report, 

considerable guidance is available on safety culture assessment, and other 

countries have established safety culture programs. 

NRC’s written response also contained technical comments, which we have 

incorporated into the report, as appropriate. (NRC’s comments and our 

responses are presented in app. IV.)

As arranged with your staff, unless you publicly announce its contents 

earlier, we plan no further distribution of this report until 30 days from its 

issue date. At that time, we plan to provide copies of this report to the 

appropriate congressional committees; the Chairman, NRC; the Director, 

Office of Management and Budget; and other interested parties. We will 

also make copies available to others upon request. In addition, this report 

will be available at no charge on the GAO Web site at http://www.gao.gov. If 

you or your staff have any questions, please call me at (202) 512-3841. Key 

contributors to this report are listed in appendix V.

Jim Wells

Director, Natural Resources

 and Environment
Page 63 GAO-04-415 Davis-Besse Nuclear Power Plant

List of Congressional Requesters

The Honorable George V. Voinovich

United States Senate

The Honorable Dennis J. Kucinich

House of Representatives

The Honorable Steven C. LaTourette

House of Representatives
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Appendix I

Time Line Relating Significant Events of Apendxies

Interest ApendxiI
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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse ApendxiI
Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix II

Analysis of the Nuclear Regulatory 

Commission’s Probabilistic Risk Assessment 

for Davis-Besse

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Appendix III

Davis-Besse Task Force Recommendations to 

NRC and Their Status, as of March 2004 ApendxiI

Recommendation NRC actions and status as of March 2004

Completed recommendations

Either fully implement or revise guidance to manage licensee 

commitments. Determine whether the periodic report on 

commitment changes submitted by licensees should continue. 

Revised instructions for these submittals and reviews to ensure 

that these tasks are accomplished. Completed in May 2003.

Determine if stress corrosion cracking models are appropriate for 

predicting susceptibility of vessel head penetration nozzles to 

pressurized water stress corrosion cracking. Determine if additional 

analysis and testing is needed to reduce modeling uncertainties for 

their continued applicability in regulatory decision making. 

Evaluated existing stress corrosion cracking models for their 

continuing use in determining susceptibility. Completed in July 

2003.

Revise the problem identification and resolution approach so that 

safety problems noted in daily licensee reports are reviewed and 

assessed. Enhance guidance to prescribe the format of information 

that is screened when deciding which problems to review. 

Revised inspection procedure for determining licensee ability to 

promptly identify and resolve conditions adverse to quality or 

safety. Completed in September 2003.

Provide enhanced inspection guidance to pursue issues and 

problems identified during reviews of plant operations. 

Revised inspection procedure for determining licensee capability 

to promptly identify and resolve conditions adverse to quality or 

safety. Completed in September 2003.

Revise inspection guidance to provide for longer-term follow-up of 

previously identified issues that have not progressed to an 

inspection finding. 

Revised inspection procedure for determining licensee capability 

to promptly identify and resolve conditions adverse to quality or 

safety. Completed in September 2003.

Revise inspection guidance to assess (1) the safety implications of 

long-standing unresolved licensee equipment problems, (2) the 

impact of phased in corrective actions, and (3) the implications of 

deferred plant modifications. 

Revised inspection procedure for determining licensee capability 

to identify and resolve conditions adverse to quality or safety. 

Completed in September 2003.

Revise inspection guidance to allow for establishing reactor 

oversight panels even when a significant performance problem, as 

defined under NRC's Reactor Oversight Process, does not exist. 

Revised inspection guidance for establishing reactor oversight 

panels. Completed in October 2003.

Assess the scope and adequacy of requirements for licensees to 

review operating experience. 

Included in NRC’s recommendation to develop a program for 

collecting, analyzing, and disseminating information on 

experiences at operating reactors. Completed in November 2003.

Ensure inspector training includes (1) boric acid corrosion effects 

and control, and (2) pressurized water stress corrosion cracking of 

nickel-based alloy nozzles. 

Developed and implemented Web-based training and a means for 

ensuring training is completed. Completed in December 2003.

Provide training and reinforce expectations to managers and staff to 

(1) maintain a questioning attitude during inspection activities, (2) 

develop inspection insights from Davis-Besse on symptoms of 

reactor coolant leakage, (3) communicate expectations to follow up 

recurring and unresolved problems, and (4) maintain an awareness 

of surroundings while conducting inspections. Establish 

mechanisms to perpetuate this training. 

Developed Web-based inspector training and a means for ensuring 

that training has been completed. NRC headquarters provided an 

overview of the training to NRC regional offices. (Training modules 

will be added and updated as needed.) Completed in December 

2003. 

Reinforce expectations that regional management should make 

every effort to visit each reactor at least once every 2 years. 

Discussed at regional counterparts meeting. Completed in 

December 2003.

Develop guidance to address impacts of regional oversight panels 

on regional resource allocations and organizational alignment.

Evaluated past and present oversight panels. Developed 

enhanced inspection approaches for oversight panels and issued 

revised procedures. Completed in December 2003.
Appendix III

Davis-Besse Task Force Recommendations to 

NRC and Their Status, as of March 2004

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Evaluate (1) the capacity to retain operating experience information 

and perform long-term operating experience reviews; (2) thresholds, 

criteria, and guidance for initiating generic communications; (3) 

opportunities for more gains in effectiveness and efficiency by 

realigning the organization (i.e., feasibility of a centralized operating 

experience "clearinghouse"); (4) effectiveness of the generic Issues 

program; and (5) effectiveness of internal dissemination of operating 

experience information to end users. 

Developed program objectives and attributes and obtained 

management endorsement of a plan to implement the 

recommendation. Developed specific recommendations to improve 

program. Evaluation completed in November 2003. 

(Implementation of recommendations resulting from this evaluation 

expected to be completed in December 2004.)

Ensure that generic requirements or guidance are not 

inappropriately affected when making unrelated changes to other 

programs, processes, guidance, etc. 

Revised inspection guidance. Completed in February 2004.

Develop inspection guidance to assess scheduler influences on 

amount of work performed during refueling outages. 

Revised the appropriate inspection procedure. Completed in 

February 2004.

Establish guidance to ensure that NRC decisions allowing licensees 

to deviate from guidelines and recommendations issued in generic 

communications are adequately documented. 

Update guidance to address documentation. Develop training and 

distribute to NRC offices and regions to emphasize compliance 

with the updated guidance. Follow up to assess the effectiveness 

of the training. Completed follow-up in February 2004.

Develop or revise inspection guidance to ensure that NRC reviews 

vessel head penetration nozzles and the reactor vessel head during 

licensee inspection activities. 

Develop or revise inspection guidance to ensure that nozzles and 

the vessel head are reviewed during licensee inspection. Issued 

interim guidance in August 2003 and a temporary inspection 

procedure in September 2003. Additional guidance expected in 

March 2004.

Develop inspection guidance to assess (1) repetitive or multiple 

technical specification actions in NRC inspection or licensee reports, 

and (2) radiation dose implications for conducting repetitive tasks. 

Revise the appropriate inspection procedure to reflect this need. 

Completion expected in March 2004.

Develop guidance to periodically inspect licensees’ boric acid 

corrosion control programs. 

Issued temporary guidance in November 2003. Completion of 

further inspection guidance changes expected in March 2004.

Reinforce expectations for managers responsible for overseeing 

operations at nuclear power plants regarding site visits, coordination 

with resident inspectors, and assignment duration. Reinforce 

expectations to question information about operating conditions and 

strengthen guidance for reviewing license amendments to 

emphasize consideration of current system conditions, reliability, 

and performance data in safety evaluation reports. Strengthen 

guidance for verifying licensee-provided information. 

Update project manager handbook that provides guidance on 

activities to be conducted during site visits and interactions with 

NRC regional staff. Also, revise guidance for considering plant 

conditions during licensing action and amendment reviews. 

Completion expected in March 2004.

Assemble and analyze foreign and domestic information on Alloy 

600 nozzle cracking. If additional regulatory action is warranted, 

propose a course of action and implement a schedule to address 

the results. 

Assemble and analyze alloy 600 cracking data. Completion 

expected in March 2004.

Recommendations due to be completed between April and December 2004

Conduct an effectiveness review of actions taken in response to past 

NRC lessons-learned reviews. 

Review past lessons-learned actions. Completion expected in April 

2004.

Provide inspection and oversight refresher training to managers and 

staff. 

Develop a training module. Completion expected in June 2004.

(Continued From Previous Page)

Recommendation NRC actions and status as of March 2004
Appendix III

Davis-Besse Task Force Recommendations to 

NRC and Their Status, as of March 2004

Page 91 GAO-04-415 Davis-Besse Nuclear Power Plant

Establish guidance for accepting owners group and industry 

recommended resolutions for generic communications and generic 

issues, including guidance for verifying that actions are taken. 

Revise office instructions to provide recommended guidance. 

Completion expected in June 2004.

Review inspection guidance to determine the inspection level that is 

sufficient during refueling outages, including inspecting reactor 

areas inaccessible during normal operations and passive 

components. 

Revised an inspection procedure to reflect these changes. Some 

inspection procedure changes were completed in November 2003, 

and additional changes are expected in August 2004. 

Evaluate, and revise as necessary, guidance for proposing 

candidate generic issues. 

Evaluate and revise guidance. Completion expected in October 

2004

Assemble and analyze foreign and domestic information on boric 

acid corrosion of carbon steel. If additional regulatory action is 

warranted, propose a course of action and implement a schedule to 

address the results. 

Review Argonne National Laboratory study on boric acid 

corrosion. Analyze data to revise inspection requirements. 

Completion expected in October 2004.

Conduct a follow-on verification of licensee actions to implement a 

sample of significant generic communications with emphasis on 

those that are programmatic in nature. 

Screen candidate generic communications to identify those most 

appropriate for follow-up using management-approved criteria. 

Develop and approve verification plan. Completion expected in 

November 2004.

Strengthen inspection guidance for periodically reviewing licensee 

operating experience. 

Incorporated into the recommendation pertaining to NRC’s 

capacity to retain operating experience information. Completion 

expected in December 2004.

Enhance the effectiveness of processes for collecting, reviewing, 

assessing, storing, retrieving, and disseminating foreign operating 

experience. 

Incorporated into the recommendation pertaining to NRC’s 

capacity to retain operating experience information. Completion 

expected in December 2004.

Update operating experience guidance to reflect the changes 

implemented in response to recommendations for operating 

experience. 

Incorporated into the recommendation pertaining to NRC’s 

capacity to retain operating experience information. Completion 

expected in December 2004.

Review a sample of NRC evaluations of licensee actions made in 

response to owners groups’ commitments to identify whether 

intended actions were effectively implemented. 

Conduct the recommended review. Completion expected in 

December 2004.

Develop general inspection guidance to periodically verify that 

licensees implement owners groups’ commitments. 

Develop inspection procedure to provide a mechanism for regions 

to support project managers’ ability to verify that licensees 

implement commitments. Completion expected in December 2004.

Conduct follow-on verification of licensee actions pertaining to a 

sample of resolved generic issues. 

No specific actions have been identified. Completion expected in 

December 2004.

Review the range of baseline inspections and plant assessment 

processes to determine sufficiency to identify and dispose of 

problems like those at Davis-Besse. 

No specific actions have been identified. Completion expected in 

December 2004.

Identify alternative mechanisms to independently assess licensee 

plant performance for self-assessing NRC oversight processes and 

determine the feasibility of such mechanisms.

No specific actions have been identified. Completion expected in 

December 2004.

Establish measurements for resident inspector staffing levels and 

requirements, including standards for satisfying minimum staffing 

levels. 

Develop standardized staffing measures and implement details. 

Metrics were developed in December 2003. Completion expected 

in December 2004.

Structure and focus inspections to assess licensee employee 

concerns and a "safety conscious work environment.” 

No specific actions have been identified. Completion expected in 

December 2004.

(Continued From Previous Page)

Recommendation NRC actions and status as of March 2004
Appendix III

Davis-Besse Task Force Recommendations to 

NRC and Their Status, as of March 2004

Page 92 GAO-04-415 Davis-Besse Nuclear Power Plant

Recommendations due to be completed in calendar year 2005

Develop inspection guidance and criteria for addressing licensee 

response to increasing leakage levels and/or adverse trends in 

unidentified reactor coolant system leakage. 

Develop recommendations for guidance with action levels to 

trigger greater NRC interaction with licensees in response to 

increased leakage. Completion expected in January 2005.

Reassess the basis for the cancellation, in 2001, of certain 

inspection procedures (i.e., boric acid control programs and 

operational experience feedback) to assess if these procedures are 

still applicable. 

Review revised procedures and reactivate as necessary. 

Completion expected in March 2005.

Assess requirements for licensee procedures to respond to plant 

alarms for leakage to determine whether requirements are sufficient 

to identify reactor coolant pressure boundary leakage. 

Review and assess adequacy of requirements and develop 

recommendations to (1) improve procedures to identify leakage 

from boundary, (2) establish consistent technical specifications for 

leakage, and (3) use enhanced leakage detection systems. 

Completion expected in March 2005. 

Determine whether licensees should install enhanced systems to 

detect leakage from the reactor coolant system. 

Re-evaluate the basis for current leakage requirements and 

assess the capabilities of current leakage detection systems. 

Develop recommendations to (1) improve procedures for 

identifying leakage, (2) establish consistent technical 

specifications, and (3) use enhanced leakage detection systems. 

Completion expected in March 2005

Inspect the adequacy of licensee’s programs to control boric acid 

corrosion, including effectiveness of implementation. 

Develop guidance to assess adequacy of corrosion control 

programs, including implementation and effectiveness, and 

evaluate the status of this effort after the first year of inspections. 

Guidance expected to be developed by March 2004. Follow-up 

scheduled for completion in March 2005.

Continue ongoing efforts to review and improve the usefulness of 

barrier integrity performance indicators and evaluate the use of 

primary system leakage that licensees have identified but not yet 

corrected as a potential indicator. 

Develop and implement improved performance indicators based 

on current requirements and measurements. Explore the use of 

additional performance indicators to track the number, duration, 

and rate of system leakage. Determine the feasibility of 

establishing a risk-informed performance indicator for barrier 

integrity. Completion expected in December 2005.

Recommendations whose completion dates have yet to be determined

Encourage the American Society of Mechanical Engineers to revise 

inspection requirements for nickel-based alloy nozzles. Encourage 

changes to requirements for nonvisual, nondestructive inspections 

of vessel head penetration nozzles. Alternatively, revise NRC 

regulations to address the nature and scope of these inspections. 

Monitor and provide input to industry efforts to develop revised 

inspection requirements. Participate in American Society of 

Mechanical Engineers’ meetings and communicate with 

appropriate stakeholders. Decide whether to endorse the revised 

American Society of Mechanical Engineers’ code requirements. 

These actions parallel a larger NRC rulemaking effort. Completion 

date yet to be determined.

Revise processes to require short- and long-term verification of 

licensee actions to respond to significant NRC generic 

communications before closing out issues. 

Target date to be set upon completion of review of NRC’s generic 

communications program. Completion date yet to be determined. 

Determine whether licensee reactor vessel head inspection 

summary reports should be submitted to NRC and, if so, revise 

submission requirements and report disposition guidance, as 

appropriate. 

Will be included as part of revised American Society of Mechanical 

Engineers’ requirements for inspection of reactor vessel heads and 

vessel head penetration nozzles. Completion date yet to be 

determined.

(Continued From Previous Page)

Recommendation NRC actions and status as of March 2004
Appendix III

Davis-Besse Task Force Recommendations to 

NRC and Their Status, as of March 2004

Page 93 GAO-04-415 Davis-Besse Nuclear Power Plant

Source: GAO analysis of NRC data.

Evaluate the adequacy of methods for analyzing the risk of passive 

component degradation and integrate these methods and risks into 

NRC’s decision-making processes. 

No specific actions have been identified. Completion date yet to be 

determined.

Review pressurized water reactor technical specifications to identify 

plants that have nonstandard reactor coolant pressure boundary 

leakage requirements and change specifications to make them 

consistent among all plants. 

Assessed plants for nonstandard technical specifications. 

Completed in July 2003. Change leakage detection specifications 

in coordination with other changes in leakage detection 

requirements. Completion date yet to be determined.

Improve requirements for unidentified leakage in reactor coolant 

system to ensure they are sufficient to (1) discriminate between 

unidentified leaks from the coolant system and leaks from the 

reactor coolant pressure boundary and (2) ensure that plants do not 

operate with pressure boundary leakage. 

Issue regulations implementing the improved requirements when 

these requirements are determined. Completion date yet to be 

determined.

NRC should review a sample of plant assessments conducted 

between 1998 and 2000 to determine if any identified plant safety 

issues have not been adequately assessed. 

No specific actions have been identified. Completion expected in 

March 2004. 

Recommendations rejected by NRC management

Review industry approaches licensees use to consider economic 

factors for inspection and repair and consider this information in 

formulating future positions on the performance of non-visual 

inspections of vessel head penetration nozzles. 

Recommendation rejected by NRC management. No completion 

date. 

Revise the criteria for review of industry topical reports to allow for 

NRC staff review of safety-significant reports that have generic 

implications but have not been formally submitted for NRC review in 

accordance with the existing criteria. 

Recommendation rejected by NRC management. No completion 

date. 

(Continued From Previous Page)

Recommendation NRC actions and status as of March 2004
Page 94 GAO-04-415 Davis-Besse Nuclear Power Plant

Appendix IV

Comments from the Nuclear Regulatory 

Commission ApendxiIV

Note: GAO comments 

supplementing those in 

the report text appear 

at the end of this 

appendix.

See comment 1.

See comment 2.
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See comment 3.

See comment 4.

See comment 5.
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See comment 1.

See comment 2.
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Commission

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See comment 3.
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Commission

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See comment 6.

See comment 7.

See comment 8.

See comment 9.
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See comment 10.

See comment 11.

See comment 12.
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See comment 4.

See comment 13.
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Comments from the Nuclear Regulatory 

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See comment 14.

See comment 15.
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Appendix IV

Comments from the Nuclear Regulatory 

Commission

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See comment 16.

See comment 17.

See comment 18.
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Comments from the Nuclear Regulatory 

Commission

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See comment 19.

See comment 20.
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Comments from the Nuclear Regulatory 

Commission

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See comment 21.

See comment 22.

See comment 23.
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Comments from the Nuclear Regulatory 

Commission

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See comment 24.

See comment 25.
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Comments from the Nuclear Regulatory 

Commission

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See comment 26.

See comment 5.

See comment 5.
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See comment 5.
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Commission

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See comment 27.

See comment 28.
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Commission

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Commission

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The following are GAO’s comments on the Nuclear Regulatory 

Commission’s letter dated May 5, 2004.

GAO Comments 1. We agree with NRC that 10 C.F.R. § 50.9 requires that information 

provided to NRC by a licensee be complete and accurate in all material 

respects, and we have added this information to the report. NRC also 

states that in carrying out its oversight responsibilities, NRC must “rely 

heavily” on licensees providing accurate information. However, we 

believe that NRC’s oversight program should not place undue reliance 

on applicants providing complete and accurate information. NRC also 

recognizes that it cannot rely solely on information from licensees, as 

evidenced by its inspection program and process for determining the 

significance of licensee violations. Under this process, NRC considers 

whether there are any willful aspects associated with the violation—

including the deliberate intent to violate a license requirement or 

regulation or falsify information. We believe that management controls, 

including inspection and enforcement, should be implemented by NRC 

so as to verify whether licensee-submitted information considered to 

be important for ensuring safety is complete and accurate as required 

by the regulation. In this regard, as stated in NRC’s enforcement policy 

guidance, NRC is authorized to conduct inspections and investigations 

(Atomic Energy Act § 161); revoke licenses for, among other things, a 

licensee’s making material false statements or failing to build or 

operate a facility in accordance with the terms of the license (Atomic 

Energy Act § 186); and impose civil penalties for a licensee’s knowing 

failure to provide certain safety information to NRC (Energy 

Reorganization Act § 206). 

With regard to the draft report conveying the expectation that NRC 

should have known about the thick layer of boron on the reactor vessel 

head, we note in the draft report that since at least 1998, NRC was 

aware that (1) FirstEnergy’s boric acid corrosion control program was 

inadequate, (2) radiation monitors within the containment area were 

continuously being clogged by boric acid deposits, (3) the containment 

air cooling system had to be cleaned repeatedly because of boric acid 

buildup, (4) corrosion was occurring within containment as evidenced 

by rust particles being found, and (5) the unidentified leakage rate had 

increased above the level that historically had been found at the plant. 

NRC was also aware of the repeated but ineffective attempts by 

FirstEnergy to correct many of these recurring problems—evidence 

that the licensee’s programs to identify and correct problems were not 
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Comments from the Nuclear Regulatory 

Commission

Page 115 GAO-04-415 Davis-Besse Nuclear Power Plant

effective. Given these indications at Davis-Besse, NRC could have 

taken more aggressive follow-up action to determine the underlying 

causes. For example, NRC could have taken action during the fuel 

outage in 1998, the shutdown to repair valves in mid-1999, or the fuel 

outage in 2000 to ensure that staff with sufficient knowledge 

appropriately investigated the types of conditions that could cause 

these indications, or followed up to ensure that FirstEnergy had fully 

investigated and successfully resolved the cause of the indications. 

2. With respect to the responsibility of the licensee to provide complete 

and accurate information, see comment 1. As to the Davis-Besse 

lessons-learned task force finding, we agree that some information 

provided by FirstEnergy in response to Bulletin 2001-01 may have been 

inconsistent with some information subsequently identified by NRC’s 

lessons-learned task force, and that had some of this information been 

known in the fall of 2001, the vessel head leakage and degradation may 

have been identified sooner than March 2002. This information 

included (1) the boric acid accumulations found on the vessel head by 

FirstEnergy in 1998 and 2000, (2) FirstEnergy’s limited ability to 

visually inspect the vessel head, (3) FirstEnergy’s boric acid corrosion 

control procedures relative to the vessel head, (4) FirstEnergy’s 

program to address the corrosive effects of small amounts of reactor 

coolant leakage, (5) previous nozzle inspection results, (6) the bases for 

FirstEnergy’s conclusion that another source of leakage—control rod 

drive mechanism flanges—was the source of boric acid deposits on the 

vessel head that obscured multiple nozzles, and (7) photographs of 

vessel head penetration nozzles. However, various NRC officials knew 

some of this information, other information should have been known 

by NRC, and the remaining information could have been obtained had 

NRC requested it from FirstEnergy. For example, according to the 

senior resident inspector, he reviewed every Davis-Besse condition 

report on a daily basis to determine whether the licensee properly 

categorized the safety significance of the conditions. Vessel head 

conditions found by FirstEnergy in 1998 and 2000 were noted in such 

condition reports or in potential-condition-adverse-to-quality reports. 

According to a FirstEnergy official, photographs of the pressure vessel 

head nozzles were specifically provided to NRC’s resident inspector, 

who, although he did not specifically recall seeing the photographs, 

stated that he had no reason to doubt the FirstEnergy official’s 

statement. NRC had been aware, in 1999, of limitations in FirstEnergy’s 

boric acid corrosion control program and, while it cited FirstEnergy for 

its failure to adequately implement the program, NRC officials did not 
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Comments from the Nuclear Regulatory 

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follow up to determine if the program had improved. Lastly, while NRC 

questioned the information provided by FirstEnergy in its submissions 

to NRC in response to Bulletin 2001-01 (regarding vessel head 

penetration nozzle inspections), NRC staff did not independently 

review and assess information pertaining to the results of past reactor 

pressure vessel head inspections and vessel head penetration nozzle 

inspections. Similarly, NRC did not independently assess the 

information concerning the extent and nature of the boric acid 

accumulations found on the vessel head by the licensee during past 

inspections. 

On page 2 of the report, we note that the Department of Justice has an 

ongoing investigation concerning the completeness and accuracy of 

information that FirstEnergy provided to NRC on the conditions at 

Davis-Besse. The investigation may or may not find that FirstEnergy 

provided inaccurate or incomplete information. While NRC notes that it 

might have detected something months earlier if information had been 

known in the fall of 2001, we would also note that the degradation of 

the reactor vessel head likely took years to occur. 

3. We believe that the statement is correct. NRC produced an estimate of 

5x10-5 per year for the change in core damage frequency, as we state in 

the report. NRC specifically documented this calculation in its 

December 2002 assessment: 

“The NRC staff estimated that, giving credit only to the [FirstEnergy] inspection 

performed in 1996, the probability of a [control rod drive mechanism] nozzle 

ejection during the period of operation from December 31, 2001, to February 16, 

2002, was in the range of 2E-3 and was an increase in the overall [loss of coolant 

accident] probability for the plant. The increase in core damage probability and 

large early release probability were estimated as approximately 5E-6 and 5E-08, 

respectively.”1

The probability of a large early release—5E-6—equates to a frequency 

of 5x10-5 per year.2

 As we note in the report, according to NRC’s 

1

The numbers 2E-3, 5E-6, and 5E-8 can also be written as 2x10-3, 5x10-6, and 5x10-8. 

2

The probability of an event occurring is the product of the frequency of an event and a 

given time period. In this case, the time period—7 weeks—was approximated as one-tenth 

of the year. Thus, 5.4x10-5 per year multiplied by 0.10 equates to a probability of 5.4x10-6. 

According to NRC, it revised 5.4x10-6 to 5.0x10-6 to account for uncertainties.
Appendix IV

Comments from the Nuclear Regulatory 

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regulatory guide 1.174, this frequency would be in the highest risk zone 

and NRC would generally not approve the requested change. 

On several occasions, we met with the NRC staff that developed the 

risk estimate in an attempt to understand how it was calculated. We 

obtained from NRC staff the risk estimate information provided to 

senior management in late November 2001, as well as several 

explanations of how the staff developed its calculations. We were 

provided with no evidence that NRC estimated the frequency of core 

damage as being 5x10-6 per year until February 2004, after our 

consultants and we had challenged NRC’s estimate as being in the 

highest risk zone under NRC’s regulatory guide 1.174. Furthermore, 

several NRC staff involved in deciding whether to issue the order to 

shut down Davis-Besse, or to allow it to continue operating until 

February 16, 2002, stated that the risk estimate they used was relatively 

high. 

4. We agree that existing regulations provide a spectrum of conditions 

under which a plant shutdown could occur and that could be 

interpreted as covering the vast majority of situations. However, we 

continue to believe that NRC lacks sufficient guidance for making plant 

shutdown decisions. We disagree on two grounds: First, the decisionmaking guidance used by NRC to shut down Davis-Besse was guidance 

for approving license change requests. This guidance provides general 

direction on how to make risk-informed decisions when licensees 

request license changes. It does not address important aspects of 

decision-making involved in deciding whether to shut down a plant. It 

also does not provide direction on how NRC should weigh 

deterministic factors in relation to probabilistic factors in making 

shutdown decisions. Secondly, while NRC views the flexibility afforded 

by its existing array of guidance as a strength, we are concerned that, 

even on the basis of the same information or circumstances, staff can 

arrive at very different decisions. Without more specific guidance, NRC 

will continue to lack accountability and the degree of credibility needed 

to convince the industry and the public that its shutdown decisions are 

sufficiently sound and reasoned for protecting public health and safety. 

5. We are aware that the commissioners have specifically decided not to 

conduct direct evaluations or inspections of safety culture. We agree 

that as regulators, NRC is not charged with managing licensees’ 

facilities, but disagree that any direct NRC involvement with safety 

culture crosses over to a management function. Management is an 
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Comments from the Nuclear Regulatory 

Commission

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embodiment of corporate beliefs and perceptions that affect 

management strategies, goals, and philosophies. These, in turn, impact 

licensee programs and processes and employee behaviors that have 

safety outcomes. We believe that NRC should not assess corporate 

beliefs and perceptions or management strategies, goals, or 

philosophies. Rather, we believe that NRC has a responsibility to assess 

licensee programs and processes, as well as employee behaviors. We 

cite several areas of safety culture in the report as being examples of 

various aspects of safety culture that NRC can assess which do not 

constitute “management functions.” The International Atomic Energy 

Agency has extensive guidance on assessing additional aspects of 

licensee performance and indicators of safety culture.3

 Such 

assessments can provide early indications of declining safety culture 

prior to when negative safety outcomes occur, such as at Davis-Besse. 

We also agree that NRC has indirect means by which it attempts to 

assess safety culture. For example, NRC’s problem identification and 

resolution inspection procedure’s stated objective is to provide an early 

warning of potential performance issues and insight into whether 

licensees have established safety conscious work environments. 

However, we do not believe that the implementation of the inspection 

procedure has been demonstrated to be effective in meeting its stated 

objectives. The inspection procedure directs inspectors to screen and 

analyze trends in all reported power plant issues. In doing so, the 

procedure directs that inspectors annually review 3 to 6 issues out of 

potentially thousands of issues that can arise and that are related to 

various structures, systems, and components necessary for the safe 

operation of the plant. This requires that inspectors judgmentally 

sample 3 to 6 issues on which they will focus their inspection 

resources. While we do not necessarily question inspector judgment 

when sampling for these 3 to 6 issues, NRC inspectors stated that due 

to the large number of issues that they can sample from, they try to 

focus on those issues that they believe have the most relevance for 

safety. Thus, if an issue is not yet perceived as being important to 

safety, it is less likely to be selected for follow up. Further, even if an 

issue were selected for follow up and this indicated that the licensee 

did not properly identify and resolve underlying problems that 

contributed to the issue, according to NRC officials, it is highly unlikely 

3

The International Atomic Energy Agency, International Nuclear Safety Advisory Group, 

Safety Culture (Vienna, Austria: February 1991).
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Page 119 GAO-04-415 Davis-Besse Nuclear Power Plant

that this one issue would rise to a high enough level of significance for 

it to be noted under NRC’s Reactor Oversight Process. Additionally, the 

procedure is dependant on the inspector being aware of, and having the 

capability to, identify issues or trends in the area of safety culture. 

According to NRC officials, inspectors are not trained in what to look 

for when assessing licensee safety culture because they are, by and 

large, nuclear engineers. While they may have an intuition that 

something is wrong, they may not know how to assess it in terms of 

safety culture. 

Additional specific examples NRC cites for indirectly assessing a 

selected number of safety culture aspects have the following 

limitations: 

• NRC’s inspection procedure for assessing licensees’ employee 

concerns program is not frequently used. According to NRC Region 

III officials, approval to conduct such an inspection must be given by 

the regional administrator and the justification for the inspection to 

be performed has to be based on a very high level of evidence that a 

problem exists. Because of this, these officials said that the 

inspection procedure has only been implemented twice in Region III.

• NRC’s allegation program provides a way for individuals working at 

NRC-regulated plants and the public to provide safety and regulatory 

concerns directly to NRC. It is a reactive program by nature because 

it is dependent upon licensees’ employees feeling free and able to 

come forward to NRC with information about potential licensee 

misconduct. While NRC follows up on those plants that have a much 

higher number of allegations than other plants to determine what 

actions licensees are taking to address any trends in the nature of the 

allegations, the number of allegations may not always provide an 

indication of a poor safety culture, and in fact, may be the reverse. 

For example, the number of allegations at Davis-Besse prior to the 

discovery of the cavity in the reactor head in March 2002 was 

relatively small. Between 1997 and 2001, NRC received 10 allegations 

from individuals at the plant. In contrast, NRC received an average of 

31 allegations per plant over the same 5-year period from individuals 

at other plants. 

• NRC’s lessons-learned reviews, such as the one conducted for DavisBesse, are generally conducted when an incident having potentially 

serious safety consequences has already occurred. 
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• With respect to NRC’s enforcement of employee protection 

regulations, NRC, under its current enforcement policy, would 

normally only take enforcement action when violations are of very 

significant or significant regulatory concern. This regulatory concern 

pertains to NRC’s primary responsibility for ensuring safety and 

safeguards and protecting the environment. Examples of such 

violations would include the failure of a system designed to prevent a 

serious safety incident not working when it is needed, a licensed 

operator being inebriated while at the control of a nuclear reactor, 

and the failure to obtain prior NRC approval for a license change that 

has implications for safety. If violations of employee protection 

regulations do not pose very significant or significant safety, 

safeguards, or environmental concerns, NRC may consider such 

violations minor. In such cases, NRC would not normally document 

such violations in inspection reports or records, and would not take 

enforcement action.

• NRC’s Reactor Oversight Process, instituted in April 2000, focuses on 

seven specific “cornerstones” that support the safety of plant 

operations to ensure reactor safety, radiation safety, and security. 

These cornerstones are: (1) the occurrence of operations and events 

that could lead to a possible accident if safety systems did not work, 

(2) the ability of safety systems to function as intended, (3) the 

integrity of the three safety barriers, (4) the effectiveness of 

emergency preparedness, (5) the effectiveness of occupational 

radiation safety, (6) the ability to protect the public from radioactive 

releases, and (7) the ability to physically protect the plant. NRC’s 

process also includes three elements that cut across these seven 

cornerstones: (1) human performance, (2) a licensee’s safetyconscious work environment, and (3) problem identification and 

resolution. NRC assumes that problems in any of these three 

crosscutting areas will be evidenced in one or more of the seven 

cornerstones in advance of any serious compromise in the safety of a 

plant. However, as evidenced by the Davis-Besse incident, this 

assumption has not proved to be true.

NRC also cites lessons-learned task force recommendations to improve 

NRC’s ability to detect problems in licensee’s safety culture, as a means 

to achieve our recommendation to directly assess licensee safety 

culture. These lessons-learned task force recommendations include (1) 

developing inspection guidance to assess the effect that a licensee’s 

fuel outage shutdown schedule has on the scope of work conducted 
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during a shutdown; (2) revising inspection guidance to provide for 

assessing the safety implications of long-standing, unresolved 

problems; corrective actions being phased in over the course of several 

years or refueling outages; and deferred plant modifications; (3) 

revising the problem identification and resolution inspection approach 

and guidance; and (4) reviewing the range of NRC’s inspections and 

assessment processes and other NRC programs to determine whether 

they are sufficient to identify and dispose of the types of problems 

experienced at Davis-Besse. While we commend these 

recommendations, we do not believe that revising such guidance will 

necessarily alert NRC inspectors to early declines in licensee safety 

culture before they result in negative safety outcomes. Further, because 

of the nature of NRC’s process for determining the relative safety 

significance of violations under NRC’s new Reactor Oversight Process, 

we do not believe that any indications of such declines will result in a 

cited violation. 

6. We have revised the report to reflect that boron in the form of boric 

acid crystals is dissolved in the cooling water. (See p. 13.)

7. On page 41 of the report, we recognize that NRC also relied on 

information provided by FirstEnergy regarding the condition of the 

vessel head. For example, in developing its risk estimate, NRC credited 

FirstEnergy with a vessel head inspection conducted in 1996. However, 

NRC decided that the information provided by FirstEnergy 

documenting vessel head inspections in 1998 and 2000 was of such 

poor quality that it did not credit FirstEnergy with having conducted 

them. As a result, NRC’s risk estimate was higher than had these 

inspections been given credit. 

8. The statement made by the NRC regional branch chief was taken 

directly from NRC’s Office of the Inspector General report on NRC’s 

oversight of Davis-Besse during the April 2000 refueling outage.4

9. We agree that up until the Davis-Besse event, NRC had not concluded 

that boric acid corrosion was a high priority issue. We clarified the text 

of the report to reflect this comment. (See p. 25.) 

4

NRC, Office of the Inspector General, NRC’s Oversight of Davis-Besse during the April 

2000 Refueling Outage (Washington, D.C.: Oct. 17, 2003).
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10. We agree that plant operators in France decided to replace their vessel 

heads in lieu of performing the extensive inspections instituted by the 

French regulatory authority. The report has been revised to add these 

details. (See p. 31.)

11. We agree that caked-on boron, in combination with leakage, could 

accelerate corrosion rates under certain conditions. However, even 

without caked-on boron, corrosion rates could be quite high. 

Westinghouse’s 1987 report on the corrosive effects of boric acid 

leakage concluded that the general corrosion rate of carbon steel can 

be unacceptably high under conditions that can prevail when primary 

coolant leaks onto surfaces and concentrates at the temperatures that 

are found on reactor surfaces. In one series of tests that it performed, 

boric acid solutions corroded carbon steel at a rate of about 0.4 inches 

per month, or about 4.8 inches a year. This was irrespective of any 

caked-on boron. In 1987, as a result of that report and extensive boric 

acid corrosion found at two other nuclear reactors that year—Salem 

unit 2 and San Onofre unit 2—NRC concluded that a review of existing 

inspection programs may be warranted to ensure that adequate 

monitoring procedures are in place to detect boric acid leakage and 

corrosion before it can result in significant degradation of the reactor 

coolant pressure boundary. However, NRC did not take any additional 

action. 

12. We agree that NRC has requirements and processes that provide a 

number of circumstances in which a plant shutdown would or could be 

required. We also recognize that there were no legal objections to the 

draft enforcement order to shut down the plant, and that the basis for 

not issuing the order was NRC’s belief that the plant did not pose an 

unacceptable risk to public health and safety. The statement in our 

report that NRC is referring to is discussing one of these 

circumstances—the licensee’s failure to meet NRC’s technical 

specification—and whether NRC believed that it had enough proof that 

the technical specification was not being met. The statement is not 

discussing the basis for NRC issuing an enforcement order. We revised 

the report to clarify this point. (See p. 34.)

13. The basis for our statement that NRC staff concluded that the first 

safety principle was probably not met was its November 29, 2001, 

briefing to NRC’s Executive Director’s Office and its November 30, 

2001, briefing to the NRC commissioners’ technical assistants. These 

briefings, the basis for which are included in documented briefing 
Appendix IV

Comments from the Nuclear Regulatory 

Commission

Page 123 GAO-04-415 Davis-Besse Nuclear Power Plant

slides, took place shortly before NRC formally notified FirstEnergy on 

December 4, 2001, that it would accept its compromise shutdown date.

14. We are referring to the same document that NRC is referring to—NRC’s 

December 3, 2002, response to FirstEnergy (NRC’s ADAMS accession 

number ML023300539). The response consists of a 2-page transmittal 

letter and an 7.3-page enclosure. The 7.3-page enclosure is 3 pages of 

background and 4.3 pages of the agency’s assessment. The assessment 

includes statements that the safety principles were met but does not 

provide an explanation of how NRC considered or weighed 

deterministic and probabilistic information in concluding that each of 

the safety factors were met. For example, NRC concluded that the 

likelihood of a loss-of-coolant accident was acceptably small because 

of the (1) staff’s preliminary technical assessment for control rod drive 

mechanism cracking, (2) evidence of cracking found at other plants 

similar to Davis-Besse, (3) analytical work performed by NRC’s 

research staff in support of the effort, and (4) information provided by 

FirstEnergy regarding past inspections at Davis-Besse. However, the 

assessment does not explain how these four pieces of information 

successfully demonstrated if and how each of the safety principles was 

met. The assessment also states that NRC examined the five safety 

principles, the fifth of which is the ability to monitor the effects of a 

risk-informed decision. The assessment is silent on whether this 

principle is met. However, in NRC’s November 29, 2001, briefing to 

NRC’s Executive Director’s Office and in its November 30, 2001, 

briefing to the NRC commissioners’ technical assistants, NRC 

concluded that this safety principle was not met. As noted above, NRC 

formally notified FirstEnergy on December 4, 2001, that it would accept 

FirstEnergy’s February 16, 2002, shutdown date.

15. See comment 3. We do not agree that the report statements 

mischaracterize the facts. Rather, we are concerned that NRC is 

misusing basic quantitative mathematics. In addition, with regard to 

NRC’s concept of an annual average change in the frequency of core 

damage, NRC stated that the agency averaged the frequency of core 

damage that would exist for the 7-week period of time (representing 

the period of time between December 31, 2001, and February 16, 2002) 

over the entire 1-year period, using the assumption that the frequency 

of core damage would be zero for the remainder of the year—February 

17, 2002, to December 31, 2002. According to our consultants, this 

calculation artificially reduced NRC’s risk estimate to a level that is 

acceptable under NRC’s guidance. By this logic, our consultants stated, 
Appendix IV

Comments from the Nuclear Regulatory 

Commission

Page 124 GAO-04-415 Davis-Besse Nuclear Power Plant

risks can always be reduced by spreading them over time; by assuming 

another 10 years of plant operation (or even longer) NRC could find 

that its calculated “risks” are completely negligible. They further stated 

that NRC’s approach is akin to arguing that an individual, who drives 

100 miles per hour 10 percent of the time, with his car otherwise 

garaged, should not be cited because his time-average speed is only 10 

miles per hour. 

Further, our consultants concluded that the “annual-average” core 

damage frequency approach was also clearly unnecessary, since one 

need only convert a core damage frequency to a core damage 

probability to handle part-year cases like the Davis-Besse case. Lastly, 

we find no basis for the calculation in any NRC guidance. According to 

our consultants, this new interpretation of NRC’s guidance is at best 

unusual and certainly is inconsistent with NRC’s guidelines regarding 

the use of an incremental core damage frequency. This interpretation 

also reinforces our consultants’ impression that perhaps there was, in 

November 2001 and possibly is still today, some confusion among the 

NRC staff regarding basic quantitative metrics that should be 

considered in evaluating regulatory and safety issues. As noted in 

comment 3, we found no evidence of this calculation prior to February 

2004. 

16. While we agree that vessel head corrosion as extensive as later found at 

Davis-Besse was not anticipated, NRC had known that leakage of the 

primary coolant from a through-wall crack could cause boric acid 

corrosion of the vessel head, as evidenced by the Westinghouse work 

cited above. Regardless of information provided to NRC by individual 

licensees, such as FirstEnergy, NRC’s model should account for known 

risks, including the potential for corrosion. 

17. We agree that NRC was aware of control rod drive mechanism nozzle 

cracking at French nuclear power plants. NRC provided us additional 

information consisting of a December 15, 1994, internal memo, in which 

NRC concluded that primary coolant leakage from a through-wall crack 

could cause boric acid corrosion of the vessel head. However, because 

some analyses indicated that it would take at least 6 to 9 years before 

any corrosion would challenge the structural integrity of the head, NRC 

concluded that cracking was not a short-term safety issue. We revised 

the report to include this additional information. (See p. 40.)

18. See comment 15.
Appendix IV

Comments from the Nuclear Regulatory 

Commission

Page 125 GAO-04-415 Davis-Besse Nuclear Power Plant

19. We agree that while not directly relevant to the Davis-Besse situation, 

NRC uses regulatory guide 1.177 to make decisions on whether certain 

equipment can be inoperable while a nuclear reactor is operating, 

which can pose very high instantaneous risks for very short periods of 

time. However, we include the reference to this particular guidance in 

the report because it was cited by an NRC official involved in the DavisBesse decision-making process as another piece of guidance used in 

judging whether the risk that Davis-Besse posed was acceptable. 

20. While regulatory guide 1.174 comprises 25 pages of guidance on how to 

use risk in making decisions on whether to allow license changes, it 

does not lay out how NRC staff are to use quantitative estimates of risk 

or probabilistic factors, or how robust these estimates must be in order 

to be considered along with more deterministic factors. The regulatory 

guide, which was first issued in mid-1998, had been in effect for only 

about 1.5 years when NRC staff was tasked with making their decision 

on Davis-Besse. According to the Deputy Executive Director of Nuclear 

Reactor Programs at the time the decision was being made, the agency 

was trying to bring the staff through the risk-informed decision-making 

process because Davis-Besse was a learning tool. He further stated that 

it was really the first time the agency had used the risk-informed 

decision-making process on operational decisions as opposed to 

programmatic decisions for licensing. At the time the decision was 

made, and currently, NRC has no guidance or criteria for use in 

assessing the quality of risk estimates or clear guidance or criteria for 

how risk estimates are to be weighed against other risk factors. 

21. The December 3, 2002, safety assessment or evaluation did state that 

the estimated increase in core damage frequency was consistent with 

NRC’s regulatory guidelines. However, as noted in comment 3, we 

disagree with this conclusion. In addition, while we agree that NRC has 

staff with risk assessment disciplines, we found no reference to these 

staff in NRC’s safety evaluation. We also found no reference to NRC’s 

statement that these staff gave more weight to deterministic factors in 

arriving at the agency’s decision. While we endorse NRC’s 

consideration of deterministic as well as probabilistic factors and the 

use of a risk-informed decision-making process, we continue to 

maintain that NRC needs clear guidance and criteria for the quality of 

risk estimates, standards of evidence, and how to apply deterministic 

as well as probabilistic factors in plant shutdown decisions. As the 

agency continues to incorporate a risk-informed process into much of 

its regulatory guidance and programs, such criteria will be increasingly 
Appendix IV

Comments from the Nuclear Regulatory 

Commission

Page 126 GAO-04-415 Davis-Besse Nuclear Power Plant

important when making shutdown as well as other types of decisions 

regarding nuclear power plants.

22. The information that NRC provided us indicates that completion dates 

for 2 of the 22 high priority recommendations have slipped.5

 One, the 

completion date for encouraging the American Society of Mechanical 

Engineers to revise vessel head penetration nozzle inspection 

requirements or, alternatively, for revising NRC’s regulations for vessel 

head inspections has slipped from June 2004 to June 2006. Two, the 

completion date for assessing NRC’s requirements that licensees have 

procedures for responding to plant leakage alarms to determine if the 

requirements are sufficient for identifying reactor coolant pressure 

boundary leakage has slipped from March 2004 to March 2005.

23. We agree with this comment and have revised the report to reflect this 

clarification. (See p. 49.)

24. Our estimate of at least an additional 200 hours of inspection per 

reactor per year is based on:

• NRC’s new requirement that its resident inspectors review all 

licensee corrective action items on a daily basis (approximately 30 

minutes per day). Given that reactors are intended to operate 

continuously throughout the year, this results in about 3.5 hours per 

week for reviewing corrective action items, or about 182 hours per 

year. In addition, resident inspections are now required to determine, 

on a semi-annual basis, whether such corrective action items reflect 

any trends in licensee performance (16 to 24 hours per year). The 

total increase for these new requirements is about 198 to 206 hours 

per reactor per year. 

• A new NRC requirement that its resident inspectors validate that 

licensees comply with additional inspection commitments made in 

response to NRC’s 2002 generic bulletin regarding reactor pressure 

vessel head and vessel head penetration nozzles. This requirement 

results in an additional 15 to 50 hours per reactor per fuel outage.

5

Of NRC’s 21 high priority recommendations, we categorized 1 recommendation as 2 so that 

we could better track actions taken to implement it. Thus, we have 22 recommendations 

categorized as high priority.
Appendix IV

Comments from the Nuclear Regulatory 

Commission

Page 127 GAO-04-415 Davis-Besse Nuclear Power Plant

25. Our draft report included a discussion that NRC management’s failure 

to recognize the scope or breadth of actions and resources necessary to 

fully implement task force recommendations could adversely affect 

how effective the actions may be. We made this statement based on 

NRC’s initial response to the Office of the Inspector General’s October 

2003 report on Davis-Besse.6 That report concluded that ineffective 

communication within NRC’s Region III and between Region III and 

NRC headquarters contributed to the Davis-Besse incident. NRC, in its 

January 2004 response to the report, stated that among other things, it 

had developed training on boric acid corrosion and revised its 

inspection program to require semi-annual trend reviews. In February 

2004, the Office of the Inspector General criticized NRC for limiting the 

agency’s efforts in responding to its findings. Specifically, it stated that 

NRC did not address underlying and generic communication failures 

identified in the Office’s report. In response to the criticism, on April 19, 

2004 (while our draft report was with NRC for review and comment), 

NRC provided the Office of the Inspector General with additional 

information to demonstrate that its actions to improve communication 

within the agency were broader than indicated in the agency’s January 

2004 response. Based on NRC’s April 19, 2004, response and the Office’s 

agreement that NRC’s actions appropriately address its concerns about 

communication within the agency, we deleted this discussion in the 

report. 

26. We recognize that the lessons-learned task force did not make a 

recommendation for improving the agency’s decision-making process 

because the task force coordinated with the Office of the Inspector 

General regarding the scope of their respective review activities and 

because the task force was primarily charged with determining why the 

vessel head degradation was not prevented. (See p. 55.) 

27. We agree that NRC’s December 3, 2002, documentation of its decision 

was prepared in response to a finding by the Davis-Besse lessonslearned task force. We revised our report to incorporate this fact. (See 

p. 55.) 

28. We agree that NRC’s lessons-learned task force conducted a 

preliminary review of reports from previous lessons-learned task forces 

6

NRC, Office of the Inspector General, NRC’s Oversight of Davis-Besse during the 2000 

Refueling Outage (Washington, D.C.: Oct. 17, 2003).
Appendix IV

Comments from the Nuclear Regulatory 

Commission

Page 128 GAO-04-415 Davis-Besse Nuclear Power Plant

and, as a result of that review, made a recommendation that the agency 

perform a more detailed effectiveness review of the actions taken in 

response to those reviews. We revised the report to reflect that NRC’s 

detailed review is currently underway. (See p. 55.)
Page 129 GAO-04-415 Davis-Besse Nuclear Power Plant

Appendix V

GAO Contacts and Staff Acknowledgments ApendxiV

GAO Contacts Jim Wells, (202) 512-3841

Ray Smith, (202) 512-6551

Staff 

Acknowledgments

In addition, Heather L. Barker, David L. Brack, William F. Fenzel, Michael L. 

Krafve, William J. Lanouette, Marcia Brouns McWreath, Judy K. Pagano, 

Keith A. Rhodes, and Carol Hernstadt Shulman made key contributions to 

this report.
Page 130 GAO-04-415 Davis-Besse Nuclear Power Plant

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to Address Safety Issues Common to Nuclear Power Plants. GAO-RCED84-149.Washington, D.C.: September 19, 1984.

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Safety Regulation. GAO-RCED-85-11. Washington, D.C.: June 19, 1985.

The Nuclear Regulatory Commission Should Report on Progress in 

Implementing Lessons Learned from the Three Mile Island Accident.

GAO-RCED-85-72. Washington, D.C.: July 19, 1985.

Nuclear Regulation: Oversight of Quality Assurance at Nuclear Power 

Plants Needs Improvement. GAO-RCED-86-41.Washington, D.C.: January 

23, 1986.

Nuclear Regulation: Efforts to Ensure Nuclear Power Plant Safety Can Be 

Strengthened. GAO-RCED-87-141. Washington, D.C.: August 13, 1987.

Nuclear Regulation: NRC’s Restart Actions Appear Reasonable—but 

Criteria Needed. GAO-RCED-89-95.Washington, D.C.: May 4, 1989.

Nuclear Regulation: NRC’s Efforts to Ensure Effective Plant Maintenance 

Are Incomplete. GAO-RCED-91-36. Washington, D.C.: December 17, 1990.

Nuclear Regulation: NRC’s Relationship with the Institute of Nuclear 

Power Operations. GAO-RCED-91-122. Washington, D.C.: May 16, 1991.

Nuclear Regulation: Weaknesses in NRC’s Inspection Program at a South 

Texas Nuclear Power Plant. GAO-RCED-96-10. Washington, D.C.: October 

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Nuclear Regulation: Preventing Problem Plants Requires More Effective 

NRC Action. GAO-RCED-97-145. Washington, D.C.: May 30, 1997.

Nuclear Regulatory Commission: Preventing Problem Plants Requires 

More Effective Action by NRC. GAO-T-RCED-98-252. Washington, D.C.: 

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Nuclear Regulation: Challenges Confronting NRC in a Changing 

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Nuclear Regulatory Commission: Status of Achieving Key Outcomes and 

Addressing Major Management Challenges. GAO-01-760. Washington, 

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Managing for Results: Efforts to Strengthen the Link between Resources 

and Results at the Nuclear Regulatory Commission. GAO-03-258. 

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Nuclear Regulatory Commission: Oversight of Security at Commercial 

Nuclear Power Plants Needs to Be Strengthened. GAO-03-752. Washington, 

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(360292)
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