UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION III

2443 WARRENVILLE ROAD, SUITE 210

LISLE, IL 60532-4352

May 7, 2012

Mr. Barry Allen

Site Vice President

FirstEnergy Nuclear Operating Company

Davis-Besse Nuclear Power Station

5501 North State Route 2, Mail Stop A-DB-3080

Oak Harbor, OH 43449-9760

SUBJECT: DAVIS-BESSE NUCLEAR POWER STATION REACTOR VESSEL HEAD 

REPLACEMENT AND SHIELD BUILDING CRACKING INSPECTION REPORT 

05000346/2012007(DRS)

Dear Mr. Allen:

On April 17, 2012, the U.S. Nuclear Regulatory Commission (NRC) completed a portion of the 

Replacement Reactor Vessel Closure Head (RVCH) inspection and additional inspections of 

your activities to demonstrate operability of the containment system following your identification 

of subsurface concrete laminar cracks in the shield building cylindrical wall. The reactor vessel 

head replacement inspection discussed in this report was limited to an evaluation of your 

engineering change for the temporary modification and restoration of the shield building. The 

inspection activities conducted subsequent to your identification of laminar cracks in the shield 

building cylindrical wall included activities related to the shield building design and licensing 

bases and shield building operability (capability to perform its specified safety functions). The 

enclosed inspection report documents the inspection results, which were discussed on 

April 17, 2012, with you and other members of your staff. The remaining portions of the Reactor 

Vessel Head Replacement Inspection are covered in other inspection reports.

The inspection examined activities conducted under your license as they relate to safety and 

compliance with the Commission’s rules and regulations and with the conditions of your license. 

The inspectors reviewed selected procedures and records, observed activities, and interviewed 

personnel.

Based on the results of this inspection, no findings of safety significance were identified by the 

NRC. 

Prior to restart from the mid-cycle outage to replace the RVCH, the NRC issued a Confirmatory 

Action Letter (CAL), No. 3-11-001, dated December 2, 2011 (ML11336A355), that confirmed 

commitments by FirstEnergy Nuclear Operating Company (FENOC) regarding the identification 

of cracks in the reinforced concrete shield building at the Davis-Besse Nuclear Power Station. 

Based on an evaluation of FENOC’s extent of condition and technical analysis of the shield 

building cracking, the NRC staff concluded that FENOC provided reasonable assurance that the 

shield building was capable of performing its safety functions. Related NRC conclusions and 

their basis were discussed during a public meeting held on January 5, 2012 (ML12030A141).
B. Allen -2-

In order to provide continued confidence, and as noted in the CAL, FENOC agreed to provide a

Root Cause Report, corrective actions, and a long term monitoring program to the NRC for our 

review by February 28, 2012. The NRC continues to evaluate your root cause activities and 

corrective actions including review of the Root Cause Report. That effort includes evaluating 

whether there were any performance deficiencies associated with the root cause that warrant 

enforcement action.

In accordance with Title 10, Code of Federal Regulations (CFR), Part 50, Section 2.390 of the 

NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any), will be 

available electronically for public inspection in the NRC Public Document Room or from the 

Publicly Available Records System (PARS) component of NRC's Agencywide Documents 

Access and Management System (ADAMS), accessible from the NRC Web site at 

http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA by Kenneth O’Brien Acting for/

Steven A. Reynolds, Director

Division of Reactor Safety

Docket No. 50-346

License No. NPF-3

Enclosure: Inspection Report 05000346/2012007

 w/Attachment: Supplemental Information

cc w/encl: Distribution via ListServ TM
U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Docket No: 50-346

License No: NPF-3

Report No: 05000346/2012007(DRS)

Licensee: FirstEnergy Nuclear Operating Company

(FENOC)

Facility: Davis-Besse Nuclear Power Station

Location: Oak Harbor, OH

Dates: July 18, 2011, through April 17, 2012

Inspectors: J. Neurauter, Senior Reactor Inspector (Lead)

D. Kimble, Senior Resident Inspector

V. Meghani, Reactor Inspector

A. Wilson, Resident Inspector

P. Cardona-Morales, Reactor Inspector

Primary NRR

Technical Specialists: K. Manoly, Senior Level Technical Advisor

A. Sheikh, Senior Structural Engineer 

D. Hoang, Structural Engineer

W. Jessup, Mechanical Engineer

A. Rezai, Materials Engineer

Approved by: David E. Hills, Chief

Engineering Branch 1

Division of Reactor Safety
1 Enclosure

SUMMARY OF FINDINGS

IR 05000346/2012007(DRS); 07/18/2011 - 04/17/2012; Davis-Besse Nuclear Power Station; 

Reactor Vessel Head Replacement and Shield Building Cracking.

This report covers an announced inspection of IP 71007, “Reactor Vessel Head Replacement,” 

limited to the licensee’s engineering change for the temporary modification and restoration of 

the shield building for the reactor vessel head replacement. In addition, this report covers 

inspection activities conducted subsequent to the licensee’s identification of laminar cracks in 

the shield building cylindrical wall. The inspection was conducted by Region III-based 

engineering inspectors with the assistance of NRR technical specialists. The NRC’s program 

for overseeing the safe operation of commercial nuclear power reactors is described in 

NUREG-1649, “Reactor Oversight Process,” Revision 4, dated December 2006.

A. NRC-Identified and Self-Revealed Findings

No findings of significance were identified.

B. Licensee-Identified Violations

No violations of significance were identified.
2 Enclosure

REPORT DETAILS

4. OTHER ACTIVITIES

Cornerstones: Initiating Events and Barrier Integrity

4OA5 Other Activities

.1 Reactor Vessel Head Replacement (Inspection Procedure 71007) – Engineering 

Changes for Temporary Modification and Restoration of the Shield Building and Shield 

Building Cracking

a. Inspection Scope

On July 18, 2011, the NRC initiated an inspection related to the engineering changes 

supporting the replacement of the reactor vessel closure head (RVCH). The objectives 

of the inspection included: (1) verification that engineering evaluations and design 

changes were completed in conformance with requirements in the facility license, the 

applicable codes and standards, licensing commitments, and the regulations, and (2) 

verification that RVCH removal and replacement activities maintained nuclear 

radiological safety.

Included in the engineering changes that supported installation of a replacement reactor 

vessel closure head (RRVCH) were temporary shield building (SB) and steel 

containment vessel (CV) construction openings to facilitate removal of the existing 

RVCH and installation of the RRVCH. The engineering changes also provided for 

restoration of the construction openings following removal of the existing RVCH and 

transport of the RRVCH to inside the steel CV.

The containment system consists of three basic structures: a steel CV, a reinforced 

concrete SB, and the internal structures (Figure 1). The CV is a cylindrical steel 

pressure vessel which houses components and systems including the reactor vessel 

and reactor coolant piping. The SB is a reinforced concrete structure that surrounds the 

CV and has a cylindrical wall nominally 30 inches thick with vertical and horizontal 

reinforcement on both the inside and outside faces and a shallow dome roof. An annular 

space is provided between the wall of the CV and the SB, and clearance is also provided 

between the CV and the dome of the SB. With the exception of the concrete under the 

CV, there are no structural ties between the CV and the SB above the foundation slab.
3 Enclosure

Figure 1: Simplified Davis-Besse Shield Building and Steel Containment Vessel 

The containment system was designed to provide protection for the public from 

radiological consequences of hypothetical accidents including a break of the largest 

reactor coolant piping. The CV provides primary means to contain the post-accident 

environment and was designed to withstand and hold against accident pressure. The 

identified cracking did not involve the CV. The design basis of the SB provided: 

(1) environmental protection of the containment vessel; (2) for a controlled release of the 

annulus atmosphere during accidents; and (3) shielding from radiation sources within the 

SB. Specifically, the SB’s function was to provide biological shielding and, in case 

radioactive leakage escapes from the CV during accident conditions, to allow the 

Emergency Ventilation System to draw a suction from the annulus region and filter that 

leakage. In addition, the SB protects the CV from external environmental hazards such 

as tornado winds and tornado driven missiles. The SB must also function to withstand 

earthquakes.

The scope of this inspection was limited to the SB including: review of engineering 

evaluations that verified the SB structural integrity, with the temporary construction 

opening; review of engineering evaluations for closure of the temporary construction 

opening; and inspection of the licensee’s activities and actions subsequent to its 

discovery of SB subsurface cracks.

Prior to the SB demolition activities to create a temporary construction opening, the 

inspector reviewed the associated engineering change, the licensee calculation that 

demonstrated SB structural integrity for the temporary condition, and the plans for 
4 Enclosure

restoration of the SB in conformance with requirements in the facility license and the 

applicable codes and standards. In addition, the inspector reviewed licensee controls for 

the SB demolition for consistency with the technical specifications for the containment 

system.

On October 10, 2011, a subsurface concrete crack-like indication was identified by the 

licensee’s contractor performing concrete hydro-demolition for the SB temporary 

construction opening. The indication was found in the flute shoulder area on the vertical 

side (left side looking from the outside) generally along the main reinforcing steel of the 

cylindrical wall, extending to across the top (approximately 6 feet), and across the 

bottom (approximately 4 feet) of the opening. The issue was entered into the 

contractor’s corrective action program (CAP) as condition report (CR) 25539-000-GCAGAMG-00182, “Fractured Concrete Found at Shield Building Construction Opening,” 

dated October 10, 2011. The licensee entered the issue into its CAP as CR-2011-03346, 

“Fractured Concrete Found at 17M Shield Building Construction Opening,” dated 

October 10, 2011, and informed the NRC via the onsite Resident Inspectors’ Office. As 

documented in CR-2011-03346, the SB pressure boundary was declared inoperable 

pending further evaluation based on the unknown impact of the cracking on the SB 

pressure boundary. The licensee also initiated a Mode Hold Restraint to ensure 

completion of evaluations necessary to demonstrate that the SB pressure boundary was 

operable prior to entering Mode 4, a Mode for which the containment system was 

required be operable per the plant’s Technical Specifications.

At the time of the discovery of SB crack-like indications, a senior NRC Region III 

inspector was on-site for inspections related to the RRVCH. Upon notification by the 

licensee, the inspector immediately focused on this emergent safety concern.

Throughout the review of the SB cracking concern, the senior Region III inspector was 

augmented by the resident inspectors, additional Region III inspectors, and NRR 

engineering specialists. The NRC’s independent review and observation of licensee 

activities related to the SB cracking included:

• maintenance of nuclear radiological safety;

• SB examinations to determine the type and extent of SB cracking, including SB 

examination methods and data evaluation;

• review of SB structural operability evaluations for identified cracking; and

• review of the SB original structural design and licensing bases.

The formation of an NRC technical review team, consisting of Region III inspectors and 

NRR technical specialists, ensured a thorough and rigorous review from multiple 

perspectives.

The records reviewed by the inspectors and technical specialists are identified in the 

Attachment to this report.

(1) Maintenance of Nuclear Radiological Safety

Because the containment system was not required to be operable during cold shutdown, 

demolition activities necessary to create the temporary construction opening continued 

to proceed after the SB cracks were identified. The inspectors confirmed that continuing 

this work would not prevent continued investigation of the SB cracking. The inspectors 
5 Enclosure

also verified that the licensee had adequate Mode Hold Restraints in-place such that SB 

and CV demolition and restoration activities maintained nuclear radiological safety.

• The SB concrete demolition was completed prior to the initiation of the CV 

demolition.

• The CV demolition did not commence prior to defueling the reactor vessel.

• Refueling the reactor vessel did not commence prior to repairing the CV 

construction opening.

• Plant restart operations (Mode 4) did not commence until the SB construction 

opening was repaired, and the containment system was determined to be 

operable as required by the Technical Specifications (i.e., the SB was capable to 

perform its design functions with identified laminar cracking).

(2) Extent of Condition Activities

The on-site inspectors observed the licensee’s activities to determine the type and 

extent of SB cracking. Following its initial identification of cracking, the licensee 

expanded SB testing based on an evaluation of the test results and the identification of 

cracked SB areas away from the RRVCH construction opening.

The inspectors observed the crack-like indications identified at the SB opening during 

the hydro-demolition process. The licensee postulated that the apparent cracks could 

be removed using a manual chipping process. Using this method, the crack indications 

along the left and bottom edges essentially disappeared, but the crack at the top of the 

opening did not disappear. Based on the observed crack characteristics, the licensee 

considered the crack to have been a circumferential laminar tear and not a radial 

‘through-thickness’ direction crack.

Based upon inspection of the crack at the top of the SB opening, the licensee 

determined that the extent of the cracking warranted further examination and 

investigation. A contractor was contacted to perform impulse response (IR) testing. The 

IR testing was a nondestructive examination technique, not required or qualified by 

codes and standards for concrete examinations. The licensee utilized IR testing to 

qualitatively assess the presence and extent of potential anomalies (e.g., concrete 

laminar cracks) within the concrete structure. 

The inspectors discussed with the IR testing contractors the capabilities and limitations 

of the IR testing method to identify subsurface laminar crack boundaries. The IR testing 

method measured the structure’s frequency at a specific location and plotted that 

frequency with adjacent readings in order to identify any changes in building frequency. 

Changes in frequency within a short span were possible indications of subsurface 

concrete cracking. The licensee supplemented its IR testing efforts with the taking of 

core borings to validate the IR testing findings as well as to determine the depth of 

suspected laminar cracks.

The IR testing readings were performed on a representative sample of all readily 

accessible areas of the SB, with the progression of IR testing based upon the indications 

of possible cracking that were obtained. From this information, the licensee concluded 

that the laminar cracking initially identified adjacent to the RRVCH opening was not 

restricted to that area. The licensee entered this extent-of-condition issue for the SB 
6 Enclosure

cracking into its CAP as CR-2011-03996 on October 19, 2011, and informed the NRC 

via the onsite Resident Inspectors’ Office.

On October 26, 2011, during investigation actions associated with CR-2011-03346, the 

licensee identified additional areas of concern via IR testing in semicircular zones above 

the main steam line penetrations through the SB. The licensee entered this extent-ofcondition issue for the SB cracking into its CAP as CR-2011-04402 and informed the 

NRC via the onsite Resident Inspectors’ Office. 

On October 31, 2011, the licensee identified additional indications of concrete cracking 

during IR testing towards the top of the SB wall, approximately between the 780 ft and 

800 ft elevations. The licensee entered this extent-of-condition issue for the SB cracking 

into its CAP as CR-2011-04648, informed the NRC via the onsite Resident Inspectors’ 

Office, and continued to investigate further to determine if any additional adverse 

conditions existed.

The cylindrical SB wall contains an inner and an outer rebar mat, each located 

approximately 3 inches into the concrete from their respective inner and outer wall 

surfaces. A rebar mat is a grid of reinforcing steel bars that adds strength to the 

concrete. In addition, the structure includes additional concrete elements (called 

shoulders) that extend out from the 2.5 foot thick wall to provide the appearance of 8 

evenly spaced, vertical cutouts (called flutes) in the outer wall surface (Figure 2). These 

shoulders (16 in total) are a maximum thickness on each side of the flute (approximately 

18 inches thick) and gradually blend into the 2.5 foot thick wall away from the flute. 

These shoulders are reinforced by additional rebar located below their concrete surfaces

(Figure 3).

Figure 2: Davis-Besse Shield Building
7 Enclosure

Figure 3: Flute and Shoulder Cross-Section

The NRC technical review team concluded that the licensee’s method to determine the 

extent of SB cracking was a logical expansion of the testing, based on areas of identified 

cracking and the extent of identified cracking in these areas; i.e., initial IR testing 

mapping and core boring were performed at the SB temporary construction opening 

affected by hydro-demolition; IR testing mapping and core boring was expanded to SB 

areas near the main steam line penetrations; IR testing mapping and core boring was 

further expanded to upper SB regions. 

In addition, the licensee performed petrographic examination of two core borings in 

accordance with the American Society for Testing and Materials Specification C856, 

“Standard Practice for Petrographic Examination of Hardened Concrete.” The test 

results were documented in CTL Group Report: “Laboratory Evaluation of Shield 

Building Concrete Cores A and D,” dated October 27, 2011 and were provided to the 

NRC for review. 

Through its extent of condition activities, the licensee determined that the cracking was

laminar in nature and generally extended circumferentially along or near the outer edge 

of the outer rebar mat for the credited concrete (2.5 foot thick cylinder). Laminar 

cracking is cracking of material, including concrete, in which there is a separation of 

layers, or planes. The laminar cracking was primarily associated with the shoulder 

regions, although some cracking was identified outside the shoulder regions near the 

main steam lines and near the top of the SB cylinder. The cracks were very tight 

(meaning the gap between the crack surfaces was extremely small). No cracking was 

identified inboard of the outer rebar mat (deeper toward the inner wall surface) around 

either the equipment access opening or in the core borings. The cracking was observed 
8 Enclosure

to be more severe on the south facing regions of the building. The cracking was interior 

to the wall surface and was not visually discernable until the licensee cut into the wall to 

create the access opening.

The inspectors verified that the licensee utilized the IR testing methodology as a nondestructive screening to identify potentially cracked regions for further evaluation and 

that core borings were taken to confirm crack boundaries based on IR test results. The 

inspectors noted good correlation between the IR testing and the confirmatory core 

borings. The inspectors reviewed test reports for two core borings that were subjected 

to further laboratory examination. Additional core borings were analyzed through the 

licensee’s root cause efforts and NRC review of those activities is a part of the ongoing 

NRC root cause and corrective action inspection and will be documented in the 

associated NRC inspection report.

The NRC technical review team engaged the licensee’s engineers and consultants in 

technical discussions while evaluating the licensee’s actions related to SB cracking 

extent of condition. As a result of NRC questions, the licensee performed additional SB 

examinations to justify licensee conclusions regarding the extent of cracking at the SB 

outer face wall, laminar cracks did not penetrate inside the outer face reinforcement, and 

the SB inner face wall was not subject to laminar cracking.

(3) SB Operability Evaluations

The licensee provided the inspectors with technical reports and calculations that 

evaluated the SB structural integrity based on IR testing mapping and core boring

results. The NRC technical review team provided extensive review of the licensee’s 

analysis and related calculations, and held multiple interactions with the licensee’s 

engineers and consultants over several weeks to discuss NRC concerns and questions. 

As a result, the licensee’s analysis and associated calculations were subject to several 

iterative changes during that timeframe including: 

• Between November 1 and November 9, 2011, the licensee provided

CR-2011-03346, “Mode Hold Resolution” dated November 2, 2011 and 

CR-2011-044402, “Mode Hold Resolution” dated November 3, 2011 and,

“Davis-Besse Shield Building Cracking Investigation and Assessment Report,” 

Revision 0, dated November 3, 2011, and several supporting calculations. 

These provided technical analysis and details to support conclusions described 

earlier in onsite presentations provided to NRC managers and staff entitled,

“Davis Besse Shield Building Cracking – NRC Informal Presentation” and 

“Bechtel Structural Evaluation of Davis Besse Shield Building.”

• Between November 17 and November 23, the licensee provided, ”Davis Besse 

Shield Building Cracking Investigation and Assessment Report,” Revision 1,

dated November 23, 2011, “Davis Besse Shield Building and Technical Summary 

Report,” Revision 0, dated November 17, 2011, and Revision 1, dated 

November 21, 2011, and several supporting calculations. On November 26 and 

November 27, 2011, a portion of the NRC technical review team conducted 

extensive onsite discussions with licensee technical staff and the licensee’s 

contractors that performed the supporting calculations. The remainder of the 

NRC technical review team continued to provide offsite support. 
9 Enclosure

• On December 1, 2011, the licensee provided revised calculations to address 

remaining concerns of the NRC technical review team. On December 1 and 

December 2, 2011, a portion of the NRC technical review team conducted 

additional onsite discussions with the licensee technical staff and the licensee’s 

contractors to ensure appropriate understanding of the latest calculation 

revisions. The remainder of the NRC technical review team continued to provide 

offsite support. 

On December 2, 2011, the NRC technical review team concluded that the licensee 

provided reasonable assurance that the SB had sufficient structural capacity to perform 

its design functions if subjected to a postulated design basis earthquake, tornado wind, 

or tornado generated missiles. The basis for the consensus conclusion included:

• IR testing mapping and confirmatory core borings provided sufficient 

characterization of the extent of cracking.

• Licensee calculations incorporated the SB extent of cracking into the structural 

evaluations in a conservative manner.

• The assumptions and engineering judgment used in the structural evaluations 

were reasonable and appropriate.

• The finite element model adequately reflected the load path through the structure 

for the degraded SB condition.

• The licensee calculations evaluated design basis earthquake, tornado wind and 

differential pressure, and tornado generated missile loads. The licensee used 

the alternative differential pressure design load from Regulatory Guide 1.76, 

“Design-Basis Tornado and Tornado Missiles for Nuclear Power Plants,”

Revision 1, applicable for the Davis-Besse site.

• The licensee calculations confirmed the calculated building stresses remained 

within acceptance limits specified in the original design and licensing bases. 

• The licensee identified additional conservatism in the calculations that was not 

credited and could provide additional safety margin.

(4) SB Design and Licensing Bases

The licensee provided Calculation C-CSS-099.20-054, “Evaluation of Shield Building for 

the Permanent Condition With Outside Vertical Reinforcement Removed at Cracking 

Areas,” Revision 002, dated November 20, 2011, and C-CSS-099.20-056, “Evaluation of 

Shield Building Hoop Reinforcement With Observed Cracking,” Revision 000, dated 

November 21, 2011, as supporting its conclusion that the intent of the current SB design 

and licensing bases was still satisfied despite the cracking.

The inspectors reviewed the SB current design and licensing bases to determine if the 

above calculations were in conformance with the current design and licensing bases. As 

part of the review of SB laminar cracking, the inspectors reviewed the original SB design 

calculations, the associated industry codes and standards identified in the SB design 

and licensing bases, and industry guidance for evaluating existing concrete structures.
10 Enclosure

The SB was designed, in-part, using rules and requirements from American Concrete 

Institute (ACI) 307-69, “Specification for the Design and Construction of Reinforced 

Concrete Chimneys.” This design standard specifies both inner face and outer face 

reinforcement for a cylindrical wall greater than 18 inches in thickness. The inspector 

did not identify alternative design rules in ACI 307-69 that addresses laminar cracking in 

proximity to the outer face reinforcement mat. In addition, the SB design was checked 

by the Ultimate Strength Design Method in accordance with ACI 318-63, “Building Code 

Requirements for Reinforced Concrete.” The inspectors did not identify industry codes 

or standards that addressed concrete reinforcement effectiveness in proximity to laminar 

cracking. 

Therefore, the inspectors, after consultation with the NRC technical review team,

questioned if laminar cracking in proximity to the outer face reinforcement was a 

condition not in conformance with the current design basis. The inspectors noted that 

revised calculations provided on December 1, 2011, did not include such a claim, and 

instead the licensee continued to review the previous conclusion. 

After additional evaluation and interactions with the inspectors, the licensee concluded in 

its Shield Building Root Cause Report dated February 27, 2012, that the SB, with the 

laminar cracking in its walls, was operable but non-conforming to the current design and 

licensing bases with regard to the design stress analysis methodology, and the tornado 

allowable stress values. 

• Davis-Besse’s Updated Safety Analysis Report (USAR) Section 3.8.2.2.5 and 

Design Criteria Manual (DCM) Section II.H.2.5.1.5 specified the analysis 

methodologies used for the SB design. These documents stated that the SB wall 

was designed using, “Analysis of Spherical Shells” from Section III of the 

1968 Edition of the American Society of Mechanical Engineers Boiler and 

Pressure Vessel Code. In the initial condition assessment, Calculations 

C-CSS-099.20-054 and C-CSS-099.20-056 used the “ANSYS” computer 

software to study the effect of the laminar cracks on the function of the SB. 

• The USAR Section 3.8.2.2.6 and Design Criteria Manual (DCM) Section 

II.H.2.5.1.5 defined the load combinations and allowed stresses for the SB 

design. Study Calculation C-CSS-099.20-056 documented that the calculated 

stress for the tornado wind and differential pressure load exceeded the allowable 

stress value in the design and licensing basis, but was within the allowable limit 

using the alternative differential pressure design load of Regulatory Guide 1.76, 

“Design-Basis Tornado and Tornado Missiles for Nuclear Power Plants,” 

Revision 1.

The inspectors noted that any future calculations used to evaluate the SB with the 

concrete laminar cracking will have to conform to the design and licensing bases. The 

licensee documented in its Root Cause Report its intention to generate an engineering 

plan by December 1, 2012, to re-establish the SB licensing basis.

(5) SB Cracking Root Cause

Prior to restart from the mid-cycle outage to replace the RVCH, the NRC issued 

Confirmatory Action Letter (CAL), No. 3-11-001, dated December 2, 2011 

(ML11336A355), that confirmed commitments by FirstEnergy Nuclear Operating 
11 Enclosure

Company (FENOC) regarding the identification of cracks in the reinforced SB at the 

Davis-Besse Nuclear Power Station. Based on evaluation of FENOC’s extent of 

condition and technical analysis of the SB cracking, the NRC staff concluded that 

FENOC provided reasonable assurance that the SB was capable of performing its safety 

functions. In order to provide continued confidence, FENOC agreed to provide a root 

cause report, corrective actions, and a long term monitoring program to the NRC for 

review by February 28, 2012.

The NRC technical review team’s rationale for not needing a root cause evaluation prior 

to startup included:

• The NRC technical review team concluded the SB could perform its safety 

function in the current degraded state

• The licensee took actions to monitor whether the condition was worsening over 

the short term and committed to develop a longer term monitoring program.

Because of the additional licensee actions with respect to extent of condition and 

evaluation of operability and licensing basis, resulting from NRC review and interactions 

with NRC staff, the licensee captured associated performance concerns in its CAP as 

CR-2012-1472 and CR-2012-5708.

NRC inspectors continued to evaluate the licensee’s root cause activities including 

review of the Root Cause Report in a separate ongoing inspection. That effort included

evaluating whether there are any performance deficiencies associated with the root 

cause that warranted enforcement action.

b. Findings

No findings of significance were identified. Specifically, the licensee revised the 

operability evaluation and supporting calculations including superseding its conclusion 

on licensing basing prior to restarting the plant. In particular, the licensee demonstrated 

the containment system was operable prior to the entering Mode 4.

4OA6 Management Meetings

.1 Exit Meeting Summary

On April 17, 2012, the inspectors presented the inspection results via telephone to the 

Site Vice President, Mr. Barry Allen, and other members of the licensee staff. The 

licensee acknowledged the issues presented. The inspectors confirmed that none of the 

potential report input discussed was considered proprietary. 

ATTACHMENT: SUPPLEMENTAL INFORMATION
1 Attachment

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

B. Allen, Site Vice President

R. Baird, Staff Engineer, Design Engineering

B. Boles, Director, Site Operations

K. Byrd, Director, Site Engineering

T. Chowdhary, NRC Liaison

C. Daft, Component Engineering

J. Dominy, Director, Site Maintenance

B. Hennessy, Supervisor, Fleet Oversight

T. Henry, Advanced Nuclear Engineer, Design Engineering

J. Hook, Manager, Design Engineering

D. Imlay, Director, Site Performance Improvement

D. Pace, FENOC Senior Vice President, Engineering

G. Wolf, Supervisor, Regulatory Compliance
2 Attachment

LIST OF ITEMS OPENED, CLOSED AND DISCUSSED

Opened

None

Closed

None

Discussed

None
3 Attachment

LIST OF DOCUMENTS REVIEWED

The following is a partial list of documents reviewed during the inspection. Inclusion on this list 

does not imply that the NRC inspector reviewed the documents in their entirety, but rather that 

selected sections or portions of the documents were evaluated as part of the overall inspection 

effort. Inclusion of a document on this list does not imply NRC acceptance of the document or 

any part of it, unless this is stated in the body of the inspection report.

Corrective Action Program Documents Reviewed:

- Bechtel Condition Report 25539-000-GCA-GAMG-00182; Fractured Concrete Found at Shield 

Building Construction Opening; dated October 10, 2011

- CR-2011-3232; Shield Building Rebar Cover Less Than Drawing Requirement; dated 

October 8, 2011

- CR-2011-3346; Fractured Concrete Found at 17M Shield Building Construction Opening; 

dated October 10, 2011

- CR-2011-3996; Extent of Condition for Shield Building Fracture Indications; dated 

October 19, 2011

- CR-2011-4190; Surface Cracks identified on Fluted Areas of Shield Building; dated 

October 23, 2011

- CR-2011-4214; Core Bore Found Additional Crack in Shield Building Architectural Flute Area; 

dated October 24, 2011

- CR-2011-4402; Fractured Concrete Found at 17M Shield Building at Main Steam Line 

Penetrations; dated October 26, 2011

- CR-2011-4507; Isolated Crack Indication Identified by Impulse Response Testing of Shield 

Building; dated October 28, 2011

- CR-2011-4648; Shield Building Impulse Response Indications Above Elevation 780; dated 

October 31, 2011

- CR-2011-4973; As-Found Concrete Cover and Spacing of Reinforcement Steel (Rebar) Do 

Not Meet Specified Design Requirements of 17M Shield Building Construction Opening; dated 

November 6, 2011

- CR-2011-5475; Concrete Cracking at Top of Shield Building Wall; dated November 16, 2011

- CR-2011-5648; Concrete Cracking in Shoulder 4/Flute 2 Region of Shield Building (Azimuth 

67.5); dated November 18, 2011

- CR-2011-5904; Errors Identified in Shield Building Crack Calculation C-CSS-059.20-056; 

dated November 25, 2011

- CR-2011-6185; Error in Calculation C-CSS-059.20-056 Revision 01; dated December 1, 2011

- CR-2012-0071; As-Found Concrete Cover and Spacing of reinforcement Steel (Rebar) Do Not 

Meet Specified Design Requirements in Locations Adjacent to the 17M Shield Building 

Opening; dated January 3, 2012

Corrective Action Program Documents Issued during Inspection: 

- CR-2012-1472; NRC Potential Violation of 50.59 on the Shield Building Cracking; dated 

February 29, 2012

- CR-2012-5708; NRC Observation Regarding Design Control of Shield Building Calculation; 

dated April 13, 2012
4 Attachment

Calculations: 

- C-CSS-099.20-045; Evaluation of Shield Building for the Construction Opening – SGR –

RVCH Replacement; Revision 0

- C-CSS-099.20-046; Evaluation of Shield Building for the Permanent Condition; Revision 0

- C-CSS-099.20-047; Restoration of Shield Building Construction Opening; Revision 0

- C-CSS-099.20-048; Seismic II/I Evaluation of Shield Building Framework in Annulus; 

Revision 0

- C-CSS-099.20-050; Tornado Depressurization for the RV Head Replacement; Revision 0

- C-CSS-099.20-053 Evaluation of Shield Building for the Interim Condition with Outside Vertical 

Reinforcement Removed at Each Flute Shoulder; Revision 0 dated November 7, 2011

- C-CSS-099.20-054; Evaluation of Shield Building for the Permanent Condition with Outside 

Vertical Reinforcement Removed at Each Flute Shoulder; Revision 0 dated October 31, 2011

- C-CSS-099.20-054; Evaluation of Shield Building for the Permanent Condition with Outside 

Vertical Reinforcement Removed at Each Flute Shoulder; Revision 1 dated November 8, 2011

- C-CSS-099.20-054; Evaluation of Shield Building for the Permanent Condition with Outside 

Vertical Reinforcement Removed at Each Flute Shoulder; Revision 2 dated 

November 20, 2011

- C-CSS-099.20-054; Evaluation of Shield Building for the Permanent Condition with Outside 

Vertical Reinforcement Removed at Each Flute Shoulder; Revision 3 dated December 1, 2011

- C-CSS-099.20-055; II/I Evaluation for Architectural Flute Shoulder; Revision 0 dated 

October 31, 2011

- C-CSS-099.20-056; Evaluation of Shield Building Hoop Reinforcement with Observed 

Cracking; Revision 0 November 21, 2011

- C-CSS-099.20-056; Evaluation of Shield Building Hoop Reinforcement with Observed 

Cracking; Revision 1 dated December 1, 2011

- C-CSS-099.20-056; Evaluation of Shield Building Hoop Reinforcement with Observed 

Cracking; Revision 2 dated December 5, 2011

- C-EE-099.01-001; Containment Conduits Affected by Tornado Missiles - Containment Vessel 

construction Opening – 17M and18 RFO; Revision 0

- VC01-B01-001; Shield Building – Thermal Stresses – Shield Wall (Independent Checking); 

Revision 0

- VC02-B01-005; Shield Building – Thermal Stresses; Revision 0

- VC03-B01-003; Membrane Stress in Shield Building Wall; Revision 0

- VC03-B01-004; Shield Building: Summarized Stresses and Reinforcement Design; Revision 0

- VS01-B01-003; Seismic Analysis of the Containment Structure; Revision 0

Drawings: 

- C-0100;Shield Building Foundation Plan & Details, Sheet 1; Revision 5

- C-0110; Shield Building Roof Plan Wall Sections & Details; Revision 6

- C-011A; Shield Building Developed Elevation; Revision 1 dated November 22, 2011

- C-011A; Shield Building Developed Elevation; Revision 2 dated November 29, 2011

- C-0112; Shield Building Details, Sheet 1; Revision 10

Miscellaneous: 

- CTL Impulse Response Mobility Plot; Revision 1 dated November 23, 2011

- Sketch SKZ904; Shield Building Developed Elevation; Revision 0 dated November 3, 2011
5 Attachment

Modifications:

- ECP 10-0458-000; SGR-17M-Shield Building Construction Opening; Revision 0

- ECP 10-0458-001; SGR-17M-Install Shield Building Construction Opening; Revision 0

- ECP 10-0458-002; SGR-17M-Restore Shield Building Construction Opening; Revision 0

- ECP 10-0458-003; SGR-17M-Ventilation of Annulus during Construction; Revision 0

Reports: 

- CTL Group Report: Laboratory Evaluation of Shield Building Concrete Cores A and D; dated 

October 27, 2011

- Davis-Besse Shield Building Cracking Investigation and Assessment Report; Revision 0 dated 

November 3, 2011

- Davis-Besse Shield Building Cracking Investigation and Assessment Report; Revision 1 dated 

November 23, 2011

- Davis-Besse Shield Building Investigation and Technical Summary; Revision 0 dated 

November 17, 2011

- Davis-Besse Shield Building Investigation and Technical Summary; Revision 1 dated 

November 21, 2011

References:

- ACI 307; Specification for the Design and Construction of Reinforced Concrete Chimneys; 

1969 Edition

- ACI 318; Building Code Requirements for Reinforced Concrete; 1963 Edition

- ACI 349.3R; Evaluation of Existing Nuclear Safety-Related Concrete Structures; 2010 Edition

- Davis Besse Updated Final Safety Analysis Report; Revision 28
6 Attachment

LIST OF ACRONYMS USED

ACI American Concrete Institute

ADAMS Agencywide Document Access Management System

CAL Confirmatory Action Letter

CAP Corrective Action Program

CFR Code of Federal Regulations

CR Condition Report

CV Containment Vessel

DCM Design Criteria Manual

DRS Division of Reactor Safety

FENOC FirstEnergy Nuclear Operating Company

IR Impulse Response

NRC U.S. Nuclear Regulatory Commission

NRR Division of Nuclear Reactor Regulation

PARS Publicly Available Records System

RRVCH Replacement Reactor Vessel Closure Head

RVCH Reactor Vessel Closure Head

SB Shield Building

USAR Updated Safety Analysis Report
B. Allen -2-

In order to provide continued confidence, and as noted in the CAL, FENOC agreed to provide a 

Root Cause Report, corrective actions, and a long term monitoring program to the NRC for 

review by February 28, 2012. NRC inspectors continue to evaluate your root cause activities 

and corrective actions including review of the Root Cause Report in a separate ongoing 

inspection. That effort includes evaluating whether there are any performance deficiencies 

associated with the root cause that warrant enforcement action.

In accordance with Title 10, Code of Federal Regulations (CFR), Part 50, Section 2.390 of the 

NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any), will be 

available electronically for public inspection in the NRC Public Document Room or from the 

Publicly Available Records System (PARS) component of NRC's Agencywide Documents 

Access and Management System (ADAMS), accessible from the NRC Web site at 

http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA by Kenneth O’Brien Acting for/

Steven A. Reynolds, Director

Division of Reactor Safety

Docket No. 50-346

License No. NPF-3

Enclosure: Inspection Report 05000346/2012007

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OFFICE RIII RIII RIII RIII

NAME JNeurauter: DHills KO’Brine for 

SReynolds

DATE 05/03/12 04/23/12 05/07/12

OFFICIAL RECORD COPY
Letter to Barry Allen from Steven Reynolds dated May 7, 2012.

SUBJECT: DAVIS-BESSE NUCLEAR POWER STATION REACTOR VESSEL HEAD 

REPLACEMENT AND SHIELD BUILDING CRACKING INSPECTION REPORT 

05000346/2012007(DRS)

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