Latest revision as of 15:53, 4 November 2019

Text

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; NRC/RGN-II/DRS/EB1
To:	Kapopoulos E; Carolina Power & Light Co
References
	IR-13-009
	Download: ML13224A290 (16)
	v • d • e

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

05000400/2013009

Dear Mr. Kapopoulos:

On July 15, 2013, 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 15, 2013, 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 Commissions 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 Shearon 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

Shearon Harris facility.

E. Kapopoulos, Jr. 2

In accordance with 10 CFR 2.390 of the NRCs 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

NRCs 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 05000400/2013009

Supplementary 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. Benjamin C. Waldrep

Vice President Vice President

Duke Energy Corporate Governance & Operation Support

Electronic Mail Distribution Duke Energy

Electronic Mail Distribution

John Dufner

Plant Manager Michael Annacone

Duke Energy Vice President

Electronic Mail Distribution Organizational Effectiveness and Regulatory

Affairs

Sean T. O'Connor Duke Energy

Manager, Support Services Electronic Mail Distribution

Duke Energy

Electronic Mail Distribution Joseph W. Donahue

Vice President - Nuclear Oversight

Frankie Womack Duke Energy

Manager, Operations Electronic Mail Distribution

Duke Energy

Electronic Mail Distribution M. Christopher Nolan

Director, Regulatory Affairs

R.J. Kidd Duke Energy

Manager, Nuclear Oversight Electronic Mail Distribution

Duke Energy

Electronic Mail Distribution Donna B. Alexander

Manager, Fleet Regulatory Affairs

David H. Corlett Duke Energy

Supervisor Electronic Mail Distribution

Licensing/Regulatory Programs

Duke Energy Carol Y. Barajas

Electronic Mail Distribution General Manager, Nuclear Operations

Duke Energy

Terry Slake Electronic Mail Distribution

Manager

Nuclear Security Edward T. ONeil

Duke Energy Director, Nuclear Protective Services

Electronic Mail Distribution Duke Energy

Electronic Mail Distribution

Mark Grantham

Manager, Engineering Timothy J. Wadsworth

Duke Energy Security Specialist

Electronic Mail Distribution Duke Energy

Electronic Mail Distribution

John W. (Bill) Pitesa

Chief Nuclear Officer (cc w/encl. continued next page)

Duke Energy

Electronic Mail Distribution

E. Kapopulous, Jr. 4

cc w/encl. continued North Carolina Utilities Commission

David Black Electronic Mail Distribution

Manager, Fleet Security

Duke Energy Robert P. Gruber

Electronic Mail Distribution Executive Director Public Staff

NCUC

Lara S. Nichols Electronic Mail Distribution

Deputy General Counsel

Duke Energy Joe Bryan

Electronic Mail Distribution Chair

Board of County Commissioners of Wake

Kate Nolan County

Associate General Counsel P.O. Box 550

Duke Energy Raleigh, NC 27602

Electronic Mail Distribution

Walter Petty

David A. Cummings Chair

Associate General Counsel Board of County Commissioners of

Duke Energy Chatham County

Electronic Mail Distribution P.O. Box 1809

Pittsboro, NC 27312

John H. O'Neill, Jr.

Shaw, Pittman, Potts & Trowbridge Senior Resident Inspector

2300 N. Street, NW U.S. Nuclear Regulatory Commission

Washington, DC 20037-1128 Shearon Harris Nuclear Power Plant

5421 Shearon Harris Rd

New Hill, NC 27562-9998

Chairman

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

U. S. NUCLEAR REGULATORY COMMISSION

REGION II

Docket No.: 50-400

License No.: NPF-63

Report No.: 05000400/2013009

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, 2013, through July 15, 2013

Inspectors: A. Alen, Reactor Inspector

T. Fanelli, Construction Inspector

Approved by: Rebecca Nease, Chief

Engineering Branch 1

Division of Reactor Safety

Enclosure

SUMMARY

IR 05000400/2013009; 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.

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, Red) 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 NRCs 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 licensees 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 created the

possibility of common cause software malfunctions of the reactor protection system and

engineered safety features 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 program, 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 licensees 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 teams questioning, the

licensee performed a 10 CFR 50.59 evaluation and concluded the change did not

require a LAR prior to implementation. The inspectors reviewed the evaluation and

could not verify the licensees 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 licensees conclusion that they could eliminate

consideration and effects of software-based common cause failures (CCF) 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 licensees

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 the 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 replacements.

5

On April 16, 2013, the licensee provided additional information regarding the analyses

and testing of the boards. The NRC staff conducted an in-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 licensees 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 created 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 licensees 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 boards were digital CPLD-based boards

that required an application-specific software (data file) to configure the boards logic

functions. These data files placed in the boards 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 licensees 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 failures 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 provides 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 provides 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 states, in part, that additional measures are

appropriate for systems that are highly safety significant (e.g., the RPS and ESFAS) to

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 licensees 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, the

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 licensees 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 (SDP) 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 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).

Enforcement: Title 10 of the Code of Federal Regulations, 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 licensees

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 15, 2013, the team presented the inspection results to Mr. Ernest Kapopoulos,

Jr., Site Vice President, and other members of the licensees staff. The team verified

that no proprietary information was retained by the inspectors or documented in this

report.

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

05000400/2013009-01 NCV Failure to Submit a License Amendment Request for a

Digital Modification to the Solid State Protection

System (Section 1R17)

Closed

05000400/2013002-03 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 Attachment

2

ML13224A290: Difference between revisions

Latest revision as of 15:53, 4 November 2019

Text

Navigation menu

ML13224A290: Difference between revisions

Latest revision as of 15:53, 4 November 2019

Text

Navigation menu

Search