IR 05000400/2013009
| ML13297A050 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 10/23/2013 |
| From: | Nease R NRC/RGN-II/DRS/EB1 |
| To: | Hamrick G Carolina Power & Light Co |
| References | |
| IR-13-009 | |
| Download: ML13297A050 (4) | |
Text
UNITED STATES ber 23, 2013
SUBJECT:
ERRATA 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 (ML13224A290)
Dear Mr. Hamrick:
The purpose of this erratum is to clarify the supporting information in the description for NCV 05000400/2013009-01 documented in Section 1R17 (report page 6) of Inspection Report 05000400/2013009. Please replace Section 1R17 (report page 6) of ML13224A290 with the page enclosed. I apologize for any inconvenience this clarification may have caused. If you have any questions, please contact me at (404) 997- 4530.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/readingrm/adams.html (The Public Electronic Reading Room).
Sincerely, RA Rebecca L. Nease, Chief Engineering Branch 1 Division of Reactor Safety Docket No.: 50-400 License No.: NPF-63 Enclosure: Corrected Page 6 Section 1R17 cc: Distribution via Listserv
Mr. George 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: ERRATA 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 (ML13224A290)
Dear Mr. Hamrick:
The purpose of this erratum is to clarify the supporting information in the description for NCV 05000400/2013009-01 documented in Section 1R17 (report page 6) of Inspection Report 05000400/2013009. Please replace Section 1R17 (report page 6) of ML13224A290 with the page enclosed. I apologize for any inconvenience this clarification may have caused. If you have any questions, please contact me at (404) 997- 4530.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/readingrm/adams.html (The Public Electronic Reading Room).
Sincerely, RA Rebecca L. Nease, Chief Engineering Branch 1 Division of Reactor Safety Docket No.: 50-400 License No.: NPF-63 Enclosure: Corrected Page 6 Section 1R17 cc: Distribution via Listserv xPUBLICLY AVAILABLE NON-PUBLICLY AVAILABLE SENSITIVE x NON-SENSITIVE ADAMS: x Yes ACCESSION NUMBER:____________x SUNSI REVIEW COMPLETE FORM 665 ATTACHED OFFICE RII DRS RII DRS RII DRS SIGNATURE RA RA RA NAME T. Fanelli R. Nease G. Hopper DATE 10/22/2013 10/23/2013 10/23/2013 10/ /2013 10/ /2013 10/ /2013 10/ /2013 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO OFFICIAL RECORD COPY DOCUMENT NAME: S:\DRS\ENG BRANCH 1\ERRATA\HARRISERRATAV.4.DOCX
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 when installed would result in an additional burden to control room operators because it would be a change to indicators in the control room. In addition, 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.
Letter to Ernest J. Kapopoulos, Jr. from Rebecca Nease dated October 23, 2013.
SUBJECT: ERRATA 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 (ML13224A290)
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