IR 05000346/2013009

IR 05000346-13-009; on 09/16/2013 - 03/28/2014, Davis-Besse Nuclear Power Station; Design and Licensing Basis of the Shield Building with Laminar Cracking Inspection

ML14132A259
Person / Time
Site:	Davis Besse
Issue date:	05/12/2014
From:	Dave Hills NRC/RGN-III/DRS/EB1
To:	Lieb R FirstEnergy Nuclear Operating Co
James Neurauter
References
IR-13-009
Download: ML14132A259 (17)

Text

UNITED STATES May 12, 2014

SUBJECT:

DAVIS-BESSE NUCLEAR POWER STATION - DESIGN AND LICENSING BASIS OF THE SHIELD BUILDING WITH LAMINAR CRACKING INSPECTION REPORT 05000346/2013009

Dear Mr. Lieb:

On March 28, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection to evaluate your actions to address the design and licensing basis of the Davis-Besse shield building with respect to the laminar cracking as specified in your Root Cause Report (ADAMS Accession Nos. ML120600056 and ML12142A053) as Direct Cause Corrective Action No. 2. In particular, the inspectors focused on testing and technical analysis and the regulatory evaluation performed by your staff and contractors to support your conclusion that your acceptance of the laminar cracking in the shield building wall with respect to the design and licensing basis did not require NRC approval. This inspection was conducted as an inspection sample pursuant to NRC Inspection Procedure 71152, Problem Identification and Resolution, under the baseline inspection program.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected records and interviewed personnel. The enclosed report documents the results of this inspection, which were discussed on March 28, 2014, with you and other members of your staff.

The NRC inspectors did not identify any findings or violations of more than minor significance.

However, the NRC inspectors identified an unresolved item regarding whether your associated 10 CFR 50.59 evaluation provided appropriate rationale to support your licensing basis conclusion. Additional NRC evaluation is planned to resolve this inspection item. In accordance with Title 10, Code of Federal Regulations (CFR), 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/

David E. Hills, Chief Engineering Branch 1 Division of Reactor Safety Docket No. 50-346 License No. NPF-3

Enclosure:

Inspection Report 05000346/2013009 w/Attachment: Supplemental Information

REGION III==

Docket No: 50-346 License No: NPF-3 Report No: 05000346/2013009 Licensee: FirstEnergy Nuclear Operating Company (FENOC)

Facility: Davis-Besse Nuclear Power Station Location: Oak Harbor, OH Dates: September 16, 2013 through March 28, 2014 Inspector: J. Neurauter, Senior Engineering Inspector Approved by: David E. Hills, Chief Engineering Branch 1 Division of Reactor Safety Enclosure

SUMMARY OF FINDINGS

IR 05000346/2013009; 09/16/2013 - 03/28/2014, Davis-Besse Nuclear Power Station; Design and Licensing Basis of the Shield Building with Laminar Cracking Inspection.

This report covers a 6-month period of inspection by one regional inspector. No findings were identified. The significance of most findings is indicated by their color (Green, White, Yellow,

Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process" (SDP). Findings for which the SDP does not apply may be "Green" or be assigned a severity level after NRC management review. 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. Inspector-Identified and Self-Revealed Findings No findings of significance were identified.

Licensee-Identified Violations

No violations of significance were identified.

REPORT DETAILS

SHIELD BUILDING DESCRIPTION The containment system for the Davis-Besse site consists of three basic structures: a steel containment vessel (CV), a reinforced concrete shield building (SB), and the internal structures (Sketch 1). The CV is a cylindrical steel pressure vessel with hemispherical dome and ellipsoidal bottom which houses the reactor vessel, reactor coolant piping, and other safety-systems. The CV is completely enclosed by a reinforced concrete SB having a cylindrical shape with a shallow dome roof. An annular space is provided between the steel CV and the interior face of the concrete SB of approximately 4.5 feet

(ft) to permit construction operations and periodic visual inspection of the steel containment vessel. The SB has an inside radius of 69.5 ft and a height of 279.5 ft measured from the top of the foundation ring to the top of the dome. The thicknesses of the SB wall and the dome are approximately 2.5 ft and 2 ft, respectively, and the exterior SB wall has eight vertical cutouts, (called flutes) spaced 45 degrees apart. These flutes consist of shoulders that extend another 1.5 ft outward and gradually taper back to the outer cylindrical wall of the SB while reaching a point of tangency 17 ft 11 inches from the centerline of the flute, (Sketch 2). The CV and SB are supported on a concrete foundation founded on a firm rock structure. With the exception of the concrete under the CV, there are no structural ties between the CV and the SB above the foundation slab. The CV provides the primary means to contain the post-accident environment and is designed to withstand and hold against accident induced pressure. The design of the SB provides for: shielding from radiation sources within the SB, controlled release of annulus atmosphere under an accident condition, and environmental protection of the CV.

Sketch 1: Simplified Davis-Besse Shield Building and Steel Containment Vessel Sketch 2: Davis-Besse Shield Building Flute and Shoulder Details BACKGROUND AND OVERVIEW During construction of the SB access opening to replace the reactor vessel head in 2011 the licensee discovered subsurface cracking located near the outer rebar mat, which extended into areas of the SB that has not been modified since original construction. The licensee investigated and confirmed the extent of subsurface laminar cracking through the use of Impulse Response testing and core boring samples taken from the SB. Specifically, laminar subsurface concrete cracks were identified along the outer rebar mat in the SB flute shoulders, at the top of the SB near the junction with the roof, and at the SB main steam line penetrations.

The licensee was able to demonstrate that the SB remained structurally adequate for the controlling load cases and remained capable of performing its safety functions. The licensees analysis and associated NRC review regarding operability of the SB are discussed in NRC Inspection Report (IR) 05000346/2012007 (ADAMS Accession No. ML12128A443).

As also discussed in that NRC inspection report, the NRC inspectors reviewed the SB design and licensing basis with respect to laminar cracking in the cylindrical wall. 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 inspectors 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 in ACI 318-63, or another industry design standard, design criteria that addressed concrete reinforcement effectiveness in proximity to laminar cracking. Hence, it appeared to the inspectors that the original design codes were no longer applicable to the current condition of the shield building. Therefore, the inspectors questioned if laminar cracking in proximity to the outer face reinforcement was a condition not in conformance with the current design and licensing basis.

The licensee performed further evaluation and addressed the concern in the licensees root cause analysis report (RCR) (ADAMS Accession Nos. ML120600056 and ML12142A053). The associated NRC inspection and evaluation of the RCR are discussed in NRC IR 05000346/2012009 (ADAMS Accession No. ML12173A023) and IR 05000346/2012010 (ADAMS Accession No. ML12276A342). The licensees RCRs concluded that the SB, with the laminar cracking in its cylindrical wall, was operable but non-conforming to the current design and licensing basis in regard to:

• The Davis-Besse Updated Safety Analysis Report (USAR) Section 3.8.2.2.5 and Design Criteria Manual 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 contrast, in the initial condition assessment of the laminar cracking, licensee 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 Section II.H.2.5.1.5 defined the load combinations and allowed stresses for the SB design. Licensee Calculation C-CSS-099.20-056, generated to address the laminar cracking, 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 licensees RCR specified Extent of Condition Corrective Action No. 1: Additional Examination of the Shield Building Exterior Wall. Specifically, the licensee identified potentially cracked or un-cracked areas of the SB using an Impulse Response testing vendor and confirmatory method (core borings) approved by the licensee.

The licensees RCR specified Direct Cause corrective actions to re-establish the design and licensing basis for the SB:

• Direct Cause Corrective Action No. 1: Testing Program to Investigate the Steel Reinforcement Capacity Adjacent to Structural Discontinuities. Specifically, the licensee administered a testing program performed at Purdue University and the University of Kansas. The test procedure was developed and performed by the selected facilities to determine the effect of the structural discontinuities (i.e., laminar cracks) on adjacent steel reinforcement splices. This Test Program and the Deliverable Test Report were reviewed and approved by the licensee.

• Direct Cause Corrective Action No. 2: Engineering Plan to Re-Establish the Design and Licensing Basis for Shield Building. Specifically, the licensee developed a plan for re-establishing SB conformance to the Davis-Besse design and licensing basis:

i.

Upon the completion of the corrective action for inspection of the SB, the exact extent of laminar cracking would be more precisely established. Upon the completion of the corrective action for the Testing Program, the capacity of the reinforcing steel adjacent to the laminar cracking would be known. The steps for re-establishing SB design and licensing bases conformance were to be finalized and additional corrective actions were to be initiated, as required.

The licensee completed two corrective actions (CAs) to re-establish the design and licensing basis of the SB with laminar cracking:

• CA-2011-03346-15; Complete Owner's Acceptance of the Design Basis Calculation and the associated 50.59 documentation for the Shield Building. The calculation included the effects of laminar cracking and university research testing results that investigated the effects of laminar cracking.

• CA-2011-03346-16; Complete Owner's Acceptance, approval and processing of USAR Change Notice for a new appendix to the USAR that reflects the current Shield Building configuration inclusive of laminar cracking, university research test results, and design basis methodology.

4OA2 Identification and Resolution of Problems

.1 Shield Building Laminar Cracking Corrective Actions

a. Inspection Scope

On September 16, 2013, the NRC initiated an inspection related to the licensees corrective actions associated with re-establishment of the design and licensing basis of the SB with identified laminar cracking. The objectives of the inspection included verification that the new design calculation and testing relating to the laminar cracking were completed in conformance with requirements in the facility license, the applicable codes and standards, licensing commitments, and the regulations. Specifically, the inspectors reviewed the licensees testing that determined the extent of SB laminar cracking and the reinforcement (rebar) splice capacity within a SB laminar crack region.

In addition, the inspectors reviewed the licensees new design calculation that concluded: the calculation methodologies were consistent with the original design basis code; ACI Standard 318-63; Building Code Requirements for Reinforced Concrete; the SB met all design requirements specified in the USAR; and the SB will perform its USAR design functions.

The inspectors also reviewed the licensees 10 CFR 50.59 evaluation dated September 10, 2013, associated with the new design calculation. The inspectors used, in part, Nuclear Energy Institute (NEI) 96-07, Guidelines for 10 CFR 50.59 Implementation, Revision 1, to determine acceptability of the completed evaluation and screening. The NEI document was endorsed by the NRC in Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, dated November 2000.

Documents reviewed are listed in the Attachment to this report.

The inspectors review of this issue constituted a single follow-up inspection sample for in-depth review as defined in IP 71152-05.

b. Findings

(1) Methodology and Acceptance Criteria Utilized for Design and Licensing Basis of Shield Building with Laminar Cracking

Introduction:

The inspectors identified an unresolved item (URI) regarding the licensees actions to re-establish the design and licensing basis for the SB with identified laminar cracking. Specifically, the inspectors questioned whether the licensee's 10 CFR 50.59 evaluation provided appropriate rationale to support its licensing basis conclusion.

Description:

The licensee used a combination of testing and calculations to re-establish the design and licensing basis of the SB with laminar cracking. The licensee used additional Impulse Response testing and confirmatory core boring data to more precisely establish the extent of SB laminar cracking. The licensee also performed testing at selected university laboratories to determine rebar splice design capacity in laminar crack areas. Using these data as input, the licensee performed an evaluation, calculation C-CSS-099.20-063, Revision 0, Shield Building Design Calculation, to demonstrate the SB with laminar cracking had structural capacity to perform its design basis functions consistent with acceptance criteria specified in the design basis code, ACI 318-63, and standard ACI 307-69, referenced in the USAR. Calculation C-CSS 099.20-063 utilized computer software ANSYS to model the shield building and calculate concrete and rebar stress for design basis loading conditions.

The inspectors reviewed licensee 10 CFR 50.59 Evaluation 13-00918, related to calculation C-CSS-099.20-063, using NEI 96-07 as guidance. As the licensee described in this evaluation, calculation C-CSS-099.20-063 provided the new evaluation of the shield building, including the effects of laminar cracking for the shield wall, dome, and spring line areas of the building. The calculation included the results of laboratory testing performed to determine the effect of laminar cracking on the structural behavior and strength of the structure. The calculation included a change in methodology.

Licensee Evaluation 13-00918 concluded that the licensees use of the ANSYS computer program does not involve a departure from the method of evaluation described in the USAR, because the planned use of ANSYS was considered approved by the NRC for the intended application. Specifically, the licensee compared their use of ANSYS for analytical evaluation of the SB with a similar application of ANSYS reviewed by the NRC and documented in an NRC memorandum dated December 15, 2011, Subject: U.S EPR Design Certification Application - Safety Evaluation with Open Items for Portions of Chapter 3, Design of Structures, Components, Equipment and Systems (ADAMS Accession Nos. ML092860252 and ML113081431). The licensee further concluded that a license amendment was not required prior to implementation of the change.

The inspectors noted that 10 CFR 50.59 allows a licensee to make changes in the facility as described in its USAR without obtaining a license amendment pursuant to 10 CFR 50.90 only if:

(1) a change to the technical specification incorporated in the license is not required, and

(2) the change does not meet any of eight criteria specified in that regulation. One of these criteria is specified in 10 CFR 50.59(c)(2)(viii) as result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design basis or in the safety analysis. The inspectors also noted that NEI 96-07, Section 4.3.8, in providing detailed guidance on evaluating changes against that specific criterion, states In general, licensees can make changes to elements of a methodology without first obtaining a license amendment if the results are essentially the same as, or more conservative than, previous results. Similarly, licensees can also use different methods without first obtaining a license amendment if those methods have been approved by the NRC for the intended application. Further, Section 4.3.8.2 in discussing changing from one method of evaluation to another, states that A new method is approved by the NRC for intended application if it is approved for the type of analysis being conducted, and applicable terms, conditions, and limitations for its use are satisfied. Further, licensees are specifically allowed to apply methods that have been reviewed and approved by the NRC, or that have been otherwise accepted as part of another plants licensing basis, without prior NRC approval. That section also provides detailed guidance for determining whether a particular application of a different method is technically appropriate for the intended application, within the bounds of what has been found acceptable to the NRC, and does not require prior NRC-approval.

In reviewing Evaluation 13-00918, the inspectors compared the previously approved application referenced by the licensee against the licensees analysis for the shield building laminar cracking. The inspectors agreed that the ANSYS software was capable of accurately modeling the laminar cracking as opposed to the original licensing basis methodology. However, given that the licensees referenced NRC-approved application did not involve modeling of laminar cracking in the structure, the inspectors questioned whether the licensees entire methodology was within the bounds of what has been found acceptable to the NRC. The NEI 96-07 specifies that it is incumbent upon the users of a new methodology to ensure they have a thorough understanding of the methodology in terms of its existing application and conditions/limitations on its use and should document in the 10 CFR 50.59 evaluation the basis for determining it is approved for use in the intended application. In particular, the inspectors questioned whether the application to the shield building laminar cracking, given its uniqueness in the nuclear industry, was sufficiently similar to the referenced NRC-approved application to consider the licensees methodology as NRC-approved or otherwise applied appropriately with respect to the following:

• It was unclear to the inspectors whether the licensee used an appropriate reference on which to base its conclusion, under provisions of NEI 96-07, that ANSYS was considered "approved by the NRC for the intended application."

Specifically, the referenced SER was issued pursuant to an interim phase of the U.S. EPR design certification review process, and hence was not considered a final SER.

• The inspectors could not identify a licensing action in which the NRC had approved the use of Impulse Response testing and confirmatory core borings to validate the design condition (extent of laminar cracking) that was modeled in the analysis.

• The inspectors could not identify a licensing action in which the NRC had approved similar licensee laboratory testing to establish/validate rebar splice capacity within laminar crack areas assumed in the analysis. In addition, the licensees test report did not demonstrate the rebar capacity acceptance criteria based on licensee tests were essentially the same as, or more conservative than rebar splice criteria specified in ACI 318-63. The inspectors noted that provisions to check the adequacy of design by use of calculation methods or suitable testing program was a key component of 10 CFR Part 50, Appendix B, Criterion III, Design Control.

• As noted in NRC IR 05000346/2013004 (ADAMS Accession No. ML13308A283),pursuant to testing conducted for the licensees shield building monitoring program, the licensee in August/September 2013 identified new crack indications some of which may be evidence of crack growth. As a result, the licensee was performing additional testing and analysis with the resulting licensee evaluation currently expected to be completed in mid-2014. Depending on those results, specifically whether cracks were in fact growing and the reason for that growth, it was not clear to the inspectors whether specific monitoring requirements or acceptance criteria with respect to extent of cracking needed to be approved by the NRC for the current operating license.

• As documented in NUREG-0136, Safety Evaluation Report related to operation of Davis-Besse Nuclear Power Station Unit 1, dated December 1976, the NRC reviewed and accepted ACI 318-63 code provisions that were used in the Davis-Besse safety analysis. Excerpts from NUREG-0136 related to ACI 318-63 include:

i. The major code used in the design of concrete seismic Category I structures is ACI 318-63.

ii. The design and analysis procedures that were used for these seismic Category I structures are the same as those approved on previously licensed applications and are in accordance with procedures delineated in ACI 318-63 and are acceptable.

iii. The various seismic Category I structures are designed and proportioned to remain within the limits established by the staff for the various load combinations. These limits are acceptable based on the ACI 318-63 Code modified as appropriate for load combinations that are considered extreme.

iv.

The criteria used in the analysis, design, and construction of all seismic Category I structures to account for anticipated loadings and postulated conditions that may be imposed upon each structure during its service lifetime are in conformance with established criteria, codes, standards, and specifications acceptable to the regulatory staff.

The licensee believed its analysis demonstrated the ACI design standard remained valid and that the SB design remained consistent with the standard despite the laminar cracking. However, since the ACI standard did not anticipate or contain provisions to govern evaluation of laminar cracking, the inspectors questioned whether the standard remained applicable/valid for the current condition. Further, it was not clear to the inspectors whether provisions in NEI 96-07 for adopting a different NRC-approved methodology could be used to supplant an industry standard specifically referenced in an NRC SER as the basis for NRC acceptance of the SB design.

The inspectors noted a provision in both ACI 318-63 and ACI 349-06, Code Requirements for Nuclear Safety-Related Concrete Structures, endorsed in current regulatory guidance documents, that would appear to allow the licensee to submit its tests and calculation supporting structural adequacy of the shield building with laminar cracking to the NRC for review and approval. This is a process that appears to be distinct from the license amendment process pursuant to 10 CFR 50.90. However, as of the end of this inspection, the licensee had not availed itself of either process to request NRC review and approval of the licensees decision to permanently accept the laminar cracking condition in the shield building wall with respect to the design and licensing basis.

This issue is considered an unresolved item pending further review and evaluation by the NRC staff to establish a position on whether the licensee's 10 CFR 50.59 evaluation provided appropriate rationale to support its licensing basis conclusion (URI 05000346/2013009-01, Methodology and Acceptance Criteria Utilized for Design and Licensing Basis of Shield Building with Laminar Cracking).

Despite the above design and licensing basis unresolved item, as noted in NRC IR 05000346/2013004, the inspectors continued to believe that the shield building laminar cracking condition remained bounded by the licensees 2011 operability evaluation, and there continued to be reasonable assurance that the shield building remained capable of performing its safety functions.

4OA6 Management Meetings

.1 Exit Meeting Summary

On March 28, 2014, the inspectors presented the inspection results to the Site Vice President, Mr. R. Lieb, and other members of the licensee staff. The licensee acknowledged the issues presented. Proprietary information reviewed by the inspectors was identified and controlled in accordance with applicable NRC policy and procedures regarding sensitive unclassified information.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

R. Lieb, Site Vice President

T. Henry, Design Engineering

J. Hook, Manager, Design Engineering

J. Sturdavant, Regulatory Compliance

G. Wolf, Supervisor, Regulatory Compliance

Nuclear Regulatory Commission

D. Kimble, Senior Resident Inspector

LIST OF ITEMS

Opened

05000346/2013009-01 URI Methodology and Acceptance Criteria Utilized for Design and Licensing Basis of the Shield Building with Laminar Cracking (Section 4OA2)

Closed and

Discussed

None Attachment