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| issue date = 08/12/2013 | | issue date = 08/12/2013 | ||
| title = 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 | | title = 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 | ||
| author name = Nease R | | author name = Nease R | ||
| author affiliation = NRC/RGN-II/DRS/EB1 | | author affiliation = NRC/RGN-II/DRS/EB1 | ||
| addressee name = Kapopoulos E | | addressee name = Kapopoulos E | ||
| addressee affiliation = Carolina Power & Light Co | | addressee affiliation = Carolina Power & Light Co | ||
| docket = 05000400 | | docket = 05000400 | ||
Line 14: | Line 14: | ||
| page count = 16 | | page count = 16 | ||
}} | }} | ||
See also: [[ | See also: [[see also::IR 05000400/2013009]] | ||
=Text= | =Text= | ||
{{#Wiki_filter: | {{#Wiki_filter: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. | 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, | 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 617061 and | |||
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 617061. (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 588797 | |||
Attachment | |||
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 617061 | |||
}} | }} |
Latest revision as of 15:53, 4 November 2019
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) | |
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
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,
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
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
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