ML13224A290

IR 05000400-13-009, 04/01/2013 07/15/2013, Shearon Harris Nuclear Power Plant, Unit 1, Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications Baseline Follow-up

ML13224A290
Person / Time
Site:	Harris
Issue date:	08/12/2013
From:	Nease R L NRC/RGN-II/DRS/EB1
To:	Kapopoulos E J Carolina Power & Light Co
References
IR-13-009
Download: ML13224A290 (16)

See also: IR 05000400/2013009

Text

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION II 245 PEACHTREE CENTER AVENUE NE, SUITE 1200

ATLANTA, GEORGIA 30303

-1257 August 12, 2013

Mr. Ernest Kapopoulos, Jr. Vice President

Shearon Harris Nuclear Power Plant

Carolina Power and Light Company

P.O. Box 165, Mail Code: Zone 1

New Hill, NC 27562

-0165

SUBJECT: SHEARON HARRIS NUCLEAR POWER PLANT

UNIT 1 - NRC EVALUATION

OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT PLANT

MODIFICATIONS

BASELINE INSPECTION FOLLOW-UP REPORT 05000400/201300

9

Dear Mr. Kapopoulos

On July 1 5, 20 1 3, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Shearon Harris

Nuclear Power Plant , Unit 1. The enclosed inspection report documents the inspection results which were discussed on July 1 5 , 20 1 3, with you and other members of your staff. 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.

One NRC-identified

finding of very low safety significance (Green) was

identified during this inspection.

This finding

was determined to involve a violation

of NRC requirements.

Additionally, the NRC has determined that

a traditional enforcement Severity Level IV violation occurred with the associated finding. The NRC is treating this

violation as

a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.

If you contest the violation or significance of this NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear

Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555

-0001; with

copies to the Regional Administrator, Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555

-0001; and the NRC Resident Inspector at the Sh earon Harris facility. If you disagree with a cross

-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your

disagreement, to the Regional Administrator, Region II; and the NRC Resident Inspector at the Sh earon Harris facility.

E. Kapopoulos, Jr.

2 In accordance with 10 CFR 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 (PARS) component of

NRC's Agencywide Document Access and Management System

(ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading

-rm/adams.html

(the Public Electronic Reading Room).

Sincerely, RA

Rebecca Nease, Chief

Engineering Branch 1

Division of Reactor Safety

Docket No.: 50

-400 License No.: NPF-63 Enclosure:

Inspection Report 05000400/20

1 3009 Supplementa

ry Information

cc: (See page 3

)

_________________________

SUNSI REVIEW COMPLETE FORM 665 ATTACHED

OFFICE RII: DRS RII: DCI NRR: DE RII: DRS RII: DRP SIGNATURE RA VIA EMAIL VIA EMAIL RA RA NAME AAlen TFanelli JThorp RNease GHopper DATE 8/07/2013 8/07/2013 8/07/2013 8/ /2013

8/ /2013

E-MAIL COPY?

YES NO YES NO YES NO YES NO YES NO

E. Kapopoulos, Jr. 3

cc: Ernest Kapopoulos, Jr. Vice President

Duke Energy

Electronic Mail Distribution

John Dufner

Plant Manager

Duke Energy

Electronic Mail Distribution

Sean T. O'Connor

Manager, Support Services

Duke Energy

Electronic Mail Distribution

Frankie Womack

Manager, Operations

Duke Energy

Electronic Mail Distribution

R.J. Kidd

Manager, Nuclear Oversight

Duke Energy

Electronic Mail Distribution

David H. Corlett

Supervisor

Licensing/Regulatory Programs

Duke Energy

Electronic Mail Distribution

Terry Slake

Manager Nuclear Security

Duke Energy

Electronic Mail Distribution

Mark Grantham

Manager, Engineering

Duke Energy

Electronic Mail Distribution

John W. (Bill) Pitesa

Chief Nuclear Officer

Duke Energy

Electronic Mail Distribution

Benjamin C. Waldrep

Vice President

Corporate Governance & Operation Support

Duke Energy

Electronic Mail Distribution

Michael Annacone Vice President

Organizational Effectiveness and Regulatory Affairs Duke Energy

Electronic Mail Distribution

Joseph W. Donahue

Vice President

- Nuclear Oversight

Duke Energy

Electronic Mail Distribution

M. Christopher Nolan

Director, Regulatory Affairs

Duke Energy

Electronic Mail Distribution

Donna B. Alexander

Manager, Fleet Regulatory Affairs

Duke Energy

Electronic Mail Distribution

Carol Y. Barajas

General Manager, Nuclear Operations

Duke Energy

Electronic Mail

Distribution

Edward T. O'Neil

Director, Nuclear Protective Services

Duke Energy

Electronic Mail Distribution

Timothy J. Wadsworth

Security Specialist

Duke Energy

Electronic Mail Distribution

(cc w/encl. continued next page)

E. Kapopulous, Jr.

4 cc w/encl. continued

David Black

Manager, Fleet Security

Duke Energy

Electronic Mail Distribution

Lara S. Nichols

Deputy General Counsel

Duke Energy

Electronic Mail Distribution

Kate Nolan

Associate General Counsel

Duke Energy

Electronic Mail Distribution

David A. Cummings

Associate General Counsel

Duke Energy

Electronic Mail Distribution

John H. O'Neill, Jr.

Shaw, Pittman, Potts & Trowbridge

2300 N. Street, NW

Washington, DC 200

37-1128

Chairman North Carolina Utilities Commission

Electronic Mail Distribution

Robert P. Gruber

Executive Director Public Staff

NCUC Electronic Mail Distribution

Joe Bryan

Chair Board of County Commissioners of Wake County P.O. Box 550

Raleigh, NC 27602

Walter Petty

Chair Board of County Commissioners of Chatham County

P.O. Box 1809

Pittsboro, NC 27312

Senior Resident Inspector

U.S. Nuclear Regulatory Commission

Shearon Harris Nuclear Power Plant

5421 Shearon Harris Rd

New Hill, NC 27562

-9998 W. Lee Cox, III

Chief, Division of Health Service Regulation, Radiation Protection Section

Electronic Mail Distribution

Letter to Ernest Kapopoulos

, Jr., from Rebecca Nease

dated August 12, 2013.

SUBJECT: SHEARON HARRIS NUCLEAR POWER PLANT UNIT 1

- NRC EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION FOLLOW

-UP REPORT 05000400/2013009

DISTRIBUTION

C. Evans, RII EICS (Part 72 Only)

L. Douglas, RII EICS (Linda Douglas)

OE Mail (email address if applicable)

RIDSNRRDIRS

PUBLIC RidsNrrPMShearonHarris Resource

Enclosure U. S. NUCLEAR REGULATORY COMMISSION

REGION II Docket No.: 50-400 License No.:

NPF-63 Report No.:

05000400/2013 009 Licensee: Carolina Power and Light Company Facility: Shearon Harris Nuclear Power Plant, Unit 1

Location: 5413 Shearon Harris Road

New Hill, NC 27562

Dates: April 1, 201 3 , through July 15 , 201 3 Inspectors:

A. Alen, Reactor Inspector

T. Fanelli, Construction Inspector

Approved by:

Rebecca Nease, Chief Engineering Branch 1

Division of Reactor Safety

SUMMARY IR 05000400/2013009; 04/01/2013 - 07/1 5/2013; Shearon Harris Nuclear Power

Plant, Unit

1; Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications

Baseline Follow-up. Two Nuclear Regulatory Commission (NRC) inspectors from Region II conducted the inspection. One Severity Level (SL) IV non-cited violation

(NCV) with an associated finding

was identified. The significance of inspection

findings is indicated by their color (i.e., greater than Green, or Green, White, Yellow, Re

d) and determined

using Inspector

Manual Chapter (IMC)

0609, "Significance Determination Process

(SDP)," dated 06/02/11.

All violations of NRC requirements are dispositioned in accordance with the NRC

's Enforcement Policy dated 1/28/13. The NRC's program

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

-1649, "Reactor Oversight Process," (ROP) Revision 4, dated December 2006.

A. NRC-Identified and Self

-Revealing Findings

Cornerstone: Mitigating Systems

SL IV: The inspectors identified a SL IV Green NCV

of 10 CFR 50.59

, "Changes, Tests, and Experiments," for the licensee's failure to obtain a license amendment before implementing a change that created the possibility of a malfunction of a system, structure, or component important to safety with a different result than previously evaluated. The licensee did not follow guidance in Nuclear Energy Institute document NEI 01-01, "Guidelines on Licensing Digital Upgrades," Rev. 1, (referenced in licensee Procedure EGR-NGGC-0157, "Engineering of Plant Digital Systems and Components," Rev. 7), which resulted in the licensee implementing a change that creat ed the possibility of common cause software malfunctions of the

reactor protection system

and engineered safety fe

atures actuation systems not previously evaluated in the Updated Final Safety Analysis Report. This failure to follow NEI guidance when implementing a change was a performance deficiency.

The licensee entered this issue into their corrective action progr

am, performed an evaluation that provided a reasonable expectation of operability, and initiated development of a

license amendment request.

The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and

capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Additionally, in accordance with the guidance in the NRC Enforcement Manual, the 10 CFR 50.59 violation was more than minor because there was reasonable likelihood

that the change would require

NRC approval prior to implementation.

The inspectors evaluated the significance of the finding using IMC 0609, "The Significance Determination Process,"

and determined the finding was of very

3 low safety significance (Green). In accordance with the Enforcement Policy, the violation of 10 CFR 50.59 was determined to be a SL IV violation because it resulted in a condition evaluated as having very low safety significance

(i.e., Green) by the SDP. The finding had

a cross-cutting aspect in the "Decision Making" component of the "Human Performance

" area because the most significant causal factor of the performance deficiency was that the licensee failed to oversee the work activities of vendors such that nuclear safety was supported H.4(c).

(Section 1R17)

B. Licensee-Identified Violations

None

REPORT DETAILS

1. REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R17 Evaluations of Changes, Tests, and

Experiments and Permanent Plant Modifications

(Closed) Unresolved Item (URI)

05000400/2013002

-03, "Solid State Protection System Digital Modification." (ML13120A340)

a. Inspection Scope

During the 2013

, baseline inspection performed in accordance with Inspection Procedure 71111.17, "

Evaluations of Changes, Tests, and

Experiments and Permanent Plant Modifications

," the team identified

a URI related to the licensee's implementation of

a permanent plant change

that replaced the solid state protection system (SSPS) control circuit boards with digital complex programmable

logic device (CPLD)

-based boards. As referenced

in site procedures

, the licensee reviewed the plant change

in accordance with the guidance and process described in Nuclear Energy Institute (NEI) 96

-07, "Guidelines for 10 CFR 50.59 Implementation," Rev. 1. The licensee determined

the change could be implemented without performing

a formal 10 CFR 50.59 evaluation to determine if a license amendment request (LAR) was required to be submitted to the

Nuclear Regulatory

Commission (NRC)

prior to implementation. The licensee failed to recognize that the software used in the replacement

boards had the potential to adversely affect the design functions of the SSPS

therefore, erroneously concluded that the change could be implemented without performing a formal 10 CFR 50.59 evaluation, and without obtaining a license amendment. Subsequent

to the team's

questioning, the licensee performed a

10 CFR 50.59

evaluation an

d concluded the change did

not require a LAR prior to implementation. The inspectors reviewed the evaluation

and could not verify the

licensee's bases for concluding that the change did not meet the 10 CFR 50.59 (c)(2)(vi) criterion for requiring a license amendment. Specifically, the inspectors could not confirm the licensee's conclusion that they could eliminate consideration and effects of software

-based common cause failures (C

CF) by meeting the Standard Review Plan (SRP) criteria contained in Branch Technical Position (BTP)

7-19 , "Guidance for Evaluation of

Diversity and Defense

-in-Depth in Digital Computer

-Based I&C Systems," Rev. 6

. This item was unresolved pending further inspection to determine if the licensee's performance constituted a violation of 10 CFR 50.59, "Evaluation of Changes, Tests, and Experiments.

" The team determined that additional

information from the licensee

and consultation with the Office of Nuclear Regulation (NRR) was warranted before reaching a final disposition of th

e URI.

On April 5, 2013

, the NRC staff conducted a meeting with the licensee and vendor of the replacement boards (Westinghouse)

to discuss the design, development, qualification, testing, and implementation

of the SSPS circuit board replacement

s.

5 On April 16, 2013

, the licensee provided additional information regarding the analyses and testing of the boards. The NRC staff conducted an

i n-office review of additional information provided by the licensee and vendor. b. Findings Introduction:

The inspectors identified a

SL IV Green NCV

of 10 CFR 50.59

, "Changes, Tests, and Experiments," for the licensee's

failure to obtain a license amendment before implementing a change that created the possibility of a malfunction of a system, structure, or component important to safety with a different result than previously evaluated. The licensee did not follow guidance in Nuclear Energy Institute document NEI 01-01, "Guidelines on Licensing Digital Upgrades," Rev. 1, (referenced in licensee Procedure EGR

-NGGC-0157, "Engineering of Plant Digital Systems and Components," Rev. 7), which resulted in the licensee

implementing a change that creat ed the possibility of common cause software malfunctions of the

reactor protection system (RPS) and engineered safety features actuation systems (ESFAS) not

previously evaluated in the Updated Final Safety Analysis Report (UFSAR). The licensee's failure to follow NEI guidance when implementing this change was a performance deficiency.

Description:

The SSPS circuit boards provide the coincidence logic to produce trip signals for the RPS and actuation signals for the ESFAS. Engineering

Change 78484, "Replace SSPS boards with new Westinghouse design boards," Rev. 6, examined a digital modification to the existing SSPS circuit boards. Unlike the original circuit boards, which used fixed logic

devices, the replacement board

s were digital CPLD

-based boards that require

d an application

-specific software (data file) to configure the board's logic functions. These data files placed in the board's CPLD memory perform a specified design basis safety function in the SSPS. Because

potential software related failures represent a new failure mode, and could

occur on each of the redundant SSPS safety trains, there is a potential increase in the likelihood of

software common cause failure

(CCF) of the safety function performed by the CPLDs and ultimately, the SSPS.

Licensee procedure EGR

-NGGC-0157, "Engineering of Plant Digital Systems and Components," Rev. 7, described the licensee's process for complying with the requirements of 10 CFR 50.59 when implementing

modifications of instrumentation and control systems employing digital equipment technology. The procedure referenced the use of guidelines contained in NEI 01

-01, "Guideline on Licensing Digital Upgrades," Rev. 1 , to evaluate digital modifications against the 10 CFR 50.59 (c)(2)(i - viii) criteria in order to determine if a LAR was required to be submitted to the NRC prior to implementation.

Section 4.4.6

, "Does the activity create a possibility for a malfunction of an SSC important to safety with a different result?" of NEI 01-01, provided guidance

on evaluating digital modifications against criterion (c)(2)(vi) of 10 CFR 50.59 with respect to software CCFs. This section stated that engineering evaluations of the quality and design processes should determine if there is reasonable assurance that the likelihood of failure s due to software

(including software CCF

), are sufficiently low and whether or not they should be considered further in the 10 CFR 50.59 evaluation process. These

6 evaluations are described further in

Sections 5.1

, "Failure Analysis," and 5.3, "Assessing Digital System Dependability

," of NEI 01

-01. Section 5.1 provide

s guidance to analyze potential failures and consequences of the digital equipment and associated software to determine if they represent an acceptable risk level. Section 5.3 provide

s guidance to evaluate the dependability of the digital equipment and its associated software. A highly dependable digital device that is developed (including its software) in accordance with

a defined life

-cycle process and complies with applicable industry standards and regulatory guidance discussed in Section 5.3.3, "Digital System Quality

," of NEI 01

-01, should provide reasonable assurance of quality and low likelihood of failures. In addition to the evaluations of the quality and design processes, Section 3.2.2

, "Software Common Cause Failures," of NEI 01

-01 state s, in part , that additional measures are appropriate for systems that are highly safety significant (e.g.

, the RPS and ESFAS) t

o achieve an acceptable level of risk. For digital modifications to such systems, defense

-in-depth and diversity (D3) in the overall plant design are analyzed (in accordance with

Section 5.2, "Defense

-in-Depth and Diversity Analysis," of NEI 01

-01) in order to assure that where there are vulnerabilities to software CCF, the plant has adequate capability to

cope with vulnerabilities

to software CCF

. The inspectors reviewed the licensee's 10 CFR 50.59 evaluation, in action request (AR)

588797, design documentation, and additional information provided by Westinghouse

(the CPLD boards' vendor

) and identified that the licensee failed to recognize the CPLD boards used software to control their safety functions and the human system interface (HSI) used by operations and maintenance. As a result, the licensee did not perform the engineering evaluations and analyses (described in

Sections 5.1 and 5.3 of NEI 01

-01) to evaluate the digital device quality and design processes. In addition, the licensee did not perform the D3 analysis (described in

Section 5.2 of NEI 01

-01) to demonstrate that D3 in the overall plant design was adequate to cope with the possibility of software CCFs. Specifically, the inspectors identified that the failure modes and effects analysis

performed by Westinghouse

did not analyze potential software failures. Additionally, th e development of the CPLD boards was outsourced to commercial vendors who used commercial software design practices and tools to design and program the CPLD boards

which did not meet the quality identified in

Section 5.3.3, "Digital System Quality," of NEI

01-01. The inspectors also identified that the new software

-based HSI for the CPLD boards resulted in an additional burden to control room operators because it resulted in changes to indicators in the control room. Specifically, a warning in the Westinghouse vendor manuals advised

of a new possible software failure mode for the HSI when maintenance personnel interfaced

with the communication port on the safeguards driver CPLD board. The inspectors could not find any evidence that the licensee had performed an evaluation of this warning.

The licensee's evaluation of criterion (c)(2)(vi) of 10 CFR 50.59 used guidance contained in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: Light Water Reactor Edition

," to evaluate software CCF for the CPLD boards. Specifically, the licensee concluded that the 'Testability' criteria in

Section 1.9, "Design Attributes to Eliminate Consideration of CCF," of

BTP 7-19, "Guidance for Evaluation of

Diversity and Defense

-in-Depth in Digital Computer

-Based I&C Systems," Rev. 6, could be used to eliminate consideration of software CCF

7 because of the hardware functional testing performed by Westinghouse. Following consultation with NRR, the inspectors

determined that the criteria in the BTP was intended to provide guidance to NRC staff in performing reviews of operating license applications (including LARs) and not as criteria to implement digital modifications under the 10 CFR 50.59 process

without prior NRC review and approval. As a result, the inspectors determined that the lack of engineering evaluations of the quality and design processes did not provide reasonable assurance that the replacement CPLD boards did not create the possibility of a software CCF of the SSPS, which

was a malfunction not previously evaluated in the UFSAR. Additionally, in failing to perform a D3 analysis the licensee did not demonstrate the capability to mitigate the effects of a software CCF, as specified by NEI 01-01, for highly safety significant

systems. The licensee entered this issue into their corrective action program as AR 617061617061and initiated development of a LAR. In addition, the licensee performed an operability evaluation. Based on the functional testing performed by the vendor and satisfactory surveillance testing, the licensee determined the SSPS was operable.

This determination, along with the boards' operating experience, provided

a reasonable

expectation that the system was operable.

Analysis: The licensee's failure to follow the guidance in NEI 01

-01 (referenced in licensee Procedure EGR

-NGGC-0157), which resulted in the licensee implementing

a change that created the possibility of common cause software malfunctions of RPS and ESFAS not previously evaluated in the UFSAR was a

performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, implementation of the new design CPLD boards affected the objective of ensuring the availability, reliability, and capability of the SSPS because the CPLD boards created the possibility of common cause software failures that were outside the current licensing bases of the SSPS.

Additionally, in accordance with the guidance in the NRC Enforcement Manual, the 10 CFR 50.59 violation was more than minor because there was reasonable likelihood that the change would require

NRC review and approval prior to implementation.

The finding was screened using the traditional enforcement process because violations

of 10 CFR 50.59 are considered to be violations that potentially impede or impact the regulatory process. Although this traditional enforcement violation is associated with a

finding that can be evaluated and communicated with a Significance Determination Process (S

DP) color reflective of the safety impact of the deficient licensee performance, the SDP does not specifically consider the regulatory process impact. Thus, although related to a common regulatory concern, it is necessary to address the traditional violation and finding using different processes to correctly reflect both the regulatory importance of the violation and the safety significance of the associated finding.

The inspectors used Inspection Manual Chapter (IMC) 0609, "Significance Determination Process," dated 6/2/11, to determine the safety significance of the finding.

8 Using IMC 0609, Attachment 4, "Initial Characterization of Findings," dated 6/19/12, Table 2, the inspectors determined that the finding affected the Mitigating Systems cornerstone. The inspectors then evaluated the finding using IMC 0609, Appendix A, "The Significance Determination Process for Findings At

-Power," dated 6/19/12, Exhibit

2 , for the Mitigating Systems Cornerstone. The inspectors

determined the finding was of very low safety significance (Green) because the deficiency affected the design of the

SSPS and was confirmed not to result in loss of operability of the system. In accordance with the NRC Enforcement Policy, Section 6.0, "Violation Examples," dated 1/28/13, a traditional enforcement violation of 10 CFR 50.59 that results in conditions evaluated as having very low safety significance (i.e., Green) by the SDP is considered a SL

IV violation (Section 6.1.d). The finding ha

d a cross-cutting aspect in the

"Decision Making" component of the "Human Performance

" area because the most significant causal factor of the performance deficiency was that the licensee failed to oversee the work activities of vendors such that nuclear safety was supported H.4(c).

Enforcement:

Title 10 of the Code of Federal Regulation

s , Part 50.59(c)(2) states, in part, that the licensee shall obtain a license amendment prior to implementing a proposed change, if the change would create a possibility of a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR. Contrary to this, the licensee failed to obtain a license amendment prior to implementing a change that created a possibility of a malfunction of the SSPS with a different result than previously evaluated in the UFSAR. Specifically, since the spring of

2012 (when the CPLD boards were installed), the licensee implemented a change to the SSPS circuit boards which created a possibility of common cause

software malfunctions of the RPS and ESFAS not previously evaluated in the UFSAR. After the team identified this issue, the licensee performed an operability evaluation and determined the SSPS was operable. Additionally, at the time of the inspection

, the licensee

had initiated development of

a LAR. This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.

The violation was entered into the licensee's corrective action program as AR 617061617061

(NCV 05000400/2013009, Failure to Submit a License Amendment Request for a Digital Modification to the Solid State Protection System)

4OA6 Management Meetings

.1 Exit Meeting Summary

On July 1 5, 2013, the team presented the inspection results to Mr. Ernest Kapopoulos , Jr., Site Vice President, and other members of the licensee's staff. The team verified that no proprietary information was retained by the inspectors or documented in this report.

Attachment

SUPPLEMENTARY

INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

D. Corlett, Supervisor, Licensing/Regulatory Programs

J. Caves, Site Licensing NRC personnel

J. Thorp, Chief, Instrumentation & Controls (I&C) Branch , Division of Engineering, NRR N. Carte, Senior Electronics Engineer, I&C Branch , Division of Engineering, NRR S. Arndt, Senior Technical Advisor for Digital I&C, Division of Engineering, NRR

J. Austin, Shearon Harris Senior Resident Inspector

P. Lessard, Shearon Harris Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000400/2013009

-0 1 NCV Failure to Submit a License Amendment Request for a Digital Modification to the Solid State Protection System (Section 1R17) Closed 05000400/2013002

-0 3 URI Solid State Protection System Digital Modification (Section 1R17)

LIST OF DOCUMENTS REVIEWED

Section 1R17:

Evaluations of Changes, Tests, and

Experiments and Permanent Plant Modifications

Engineering Change

EC 78484, Digital Modification to SSPS Control Boards, Rev. 6 Basis Documents

Technical Specifications, Current

Updated Final Safety Analysis Report, Current

Condition Reports Reviewed

AR 588797588797

2 Other Documents

Branch Technical Position 7

-19 (NUREG-0800), Guidance for Evaluation of Diversity and Defense-in-Depth in Digital Computer

-Based Instrumentation and Control Systems, Rev.6

MDES-EDS-A-418A Eng. Data Sheet Universal Logic Board Configuration Settings

MDES-EDS-A-511A Eng. Data Sheet Safeguards Driver Boards Configuration Settings

MDES-EDS-A-515A Eng. Data Sheet Under voltage Output Board Configuration Settings

Nuclear Energy Institute, NEI

01-01, Guideline on Licensing Digital Upgrade

- EPRI TR-102348, Rev.1

Nuclear Energy Institute, NEI 96-07, Guidelines for 10 CFR 50.59 Implementation, Rev.1

WCAP-16769-P, WEC SSPS Universal Logic Board Replacement Summary Rpt, Rev. 2

WCAP-16770-P, WEC SSPS Safeguards Driver Board Replacement Summary Rpt, Rev. 0

WCAP-16771-P, WEC SSPS Under voltage Driver Board Replacement Summary Rpt, Rev. 1

WNA-TR-02644-SCP, SSPS New Design Circuit Boards Final Logic Test Rpt, Rev. 0

Z05R0 Questions to Westinghouse

(EC 70350)

Z20R5 Westinghouse Email on Frozen MCB

(EC 70350)

Westinghouse Electric Co. letter to John Caves, Duke Energy

- Reg. Affairs, March 7, 2013

Westinghouse Electric Co. letter to John Caves, Duke Energy

- Reg. Affairs, April 16, 2013

Action Requests Written as a Result of the Inspection

AR 617061617061