IR 05000331/2016007

Evaluations of Changes, Tests and Experiments, and Permanent Plant Modifications Baseline Inspection Report 05000331/2016007 (AAS)

ML16083A482
Person / Time
Site:	Duane Arnold
Issue date:	03/22/2016
From:	Robert Daley Engineering Branch 3
To:	Vehec T NextEra Energy Duane Arnold
References
IR 2016007
Download: ML16083A482 (16)

Text

UNITED STATES rch 22, 2016

SUBJECT:

DUANE ARNOLD ENERGY CENTER, EVALUATIONS OF CHANGES, TESTS AND EXPERIMENTS, AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000331/2016007

Dear Mr. Vehec:

On January 29, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations of Changes, Tests, and Experiments, and Permanent Plant Modifications inspection at your Duane Arnold Energy Center. The enclosed inspection report documents the inspection results, which were discussed on February 29, 2016, with Mr. R. Murrell 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.

Two NRC-identified findings of very-low safety significance (Green) were identified during this inspection. The findings were determined to involve a violation of NRC requirements.

However, because of their very-low safety significance, and because the issues were entered into your Corrective Action Program, the NRC is treating the issues as Non-Cited Violations in accordance with Section 2.3.2 of the NRC Enforcement Policy.

If you contest the subject or severity of the Non-Cited-Violations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, Region III; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at Duane Arnold Energy Center.

In addition, if you disagree with the cross-cutting aspect assigned to any finding 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 III, and the NRC Resident Inspector at Duane Arnold Energy Center. In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public Inspections, Exemptions, Requests for Withholding, of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS)

component of the NRC's Agencywide Documents 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/

Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Docket No. 50-331 License No. DPR-49

Enclosure:

Inspection Report 05000331/2015001

REGION III==

Docket No: 50-331 License No: DPR-49 Report No: 05000331/2016007 Licensee: NextEra Energy Duane Arnold, LLC Facility: Duane Arnold Energy Center Location: Palo, IA Dates: January 11 - 29, 2016 Inspectors: A. Shaikh, Senior Reactor Inspector (Lead)

I. Khan, Reactor Inspector J. Bozga, Reactor Inspector Approved by: Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Enclosure

SUMMARY

Inspection Report 05000331/2016007; 01/11/2016 - 01/29/2016; Duane Arnold Energy Center;

Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications.

This report covers a 2-week announced baseline inspection on evaluations of changes, tests, and experiments, and permanent plant modifications. The inspection was conducted by Region III based engineering inspectors. Two findings of very-low safety significance were identified by the inspectors. Each violation was considered a Non-Cited Violation (NCV) of U.S. Nuclear Regulatory Commission (NRC) regulations. The significance of most findings is indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Cross-cutting aspects were determined using IMC 0310, Aspects within the Cross-Cutting Areas. Findings and/or violations for which the SDP does not apply may be Green, or be assigned a severity level after NRC management review. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy, dated July 9, 2013. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 5, dated February 201

NRC-Identified

and Self-Revealed Findings Cornerstones: Mitigating Systems

Green.

The inspectors identified a finding of very low safety significance (Green) and associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B,

Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to document the review performed to conclude that a 50.59 evaluation was not required.

Specifically, the licensee failed to document the reviews performed to determine that installation of portable electric heaters in battery rooms would not have an adverse effect on the safety related batteries.

The inspectors determined that the licensees failure to document the reviews performed to conclude that a 50.59 evaluation was not required was contrary to procedure EN-AA-203-1201, 10 CFR Applicability and 10 CFR 50.59 Screening Reviews, and was a performance deficiency (PD). The PD was determined to be more than minor, and a finding, because if left uncorrected, the PD would become a more significant safety concern. Specifically, installation of portable electric heaters in battery rooms may increase the probability of hydrogen ignition and challenge the ability of safety related batteries to perform their safety function. In accordance with IMC 0609,

Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 2 the inspectors determined the finding affected the Mitigating Systems cornerstone. As a result, the inspectors determined the finding could be evaluated using Appendix A, The Significance Determination Process (SDP) for Findings At-Power,

Exhibit 2 for the Mitigating Systems cornerstone. The finding screened as very-low safety significance (i.e. Green) because it did not result in the loss of operability or functionality of any structure, system, or component. Specifically, the licensee did not enter a condition that required the installation of portable electric heaters in the battery room per Procedure AOP 904. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current licensee performance.

Severity Level IV. The inspectors identified a Severity Level IV, NCV of 10 CFR 50.59,

Changes, Tests, and Experiments, having very-low safety significance (Green) for the licensees failure to document the basis for making a change to Updated Final Safety Analysis Report (UFSAR) Table 15.0-2 to allow the use of RADTRAD Version 3.03 for all Chapter 15 Accidents. Specifically, the licensee failed to demonstrate that the change to UFSAR Table 15.0-2 did not constitute a Departure from a Method of Evaluation described in the UFSAR and would have never required prior NRC review and approval.

The inspectors determined that the failure to evaluate whether the change to UFSAR Table 15-0.2 constituted a Departure from a Method of Evaluation was contrary to 10 CFR 50.59(d)(1) and was a PD. The PD was determined to be more than minor, and a finding, because if left uncorrected, the PD had the potential to become a more significant safety concern. Specifically, the inspectors could not reasonably determine that use of RADTRAD version 3.03 for all UFSAR Chapter 15 Accidents would not have increased the control room dose value during accidents. In addition, the associated violation was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval. The inspectors determined that finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process. Using Attachment 0609.04, Initial Characterization of Findings, Table 2 the inspectors determined that the finding affected the Barrier Integrity cornerstone. As a result, the inspectors evaluated the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 3 for the Barrier Integrity cornerstone.

The inspectors answered Yes to question C.1 in Exhibit 3 - Barrier Integrity Screening Questions. Specifically, the inspectors determined the finding only represented a degradation of the radiological barrier function provided for the control room. In accordance with Section 6.1.d of the NRC Enforcement Policy this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very-low safety significance (i.e., green finding). In accordance with IMC 0612, Power Reactor Inspection Reports, Section 07.03.c, the inspectors did not assign a cross-cutting aspect to this violation because the violation and underlying technical finding was not indicative of current plant performance

Licensee-Identified Violations

No violations were identified.

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R17 Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications

.1 Evaluation of Changes, Tests, and Experiments

a. Inspection Scope

The inspectors reviewed one safety evaluation performed pursuant to Title 10, Code of Federal Regulations (CFR), Part 50.59 to determine if the evaluation was adequate and that prior U.S. Nuclear Regulatory Commission (NRC) approval was obtained as appropriate. The inspectors also reviewed twenty screenings and/or applicability determinations where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. The inspectors reviewed these documents to determine if:

the changes, tests, and experiments performed were evaluated in accordance with 10 CFR 50.59, and that sufficient documentation existed to confirm that a license amendment was not required; the safety issue requiring the change, tests or experiment was resolved; the licensee conclusions for evaluations of changes, tests, and experiments were correct and consistent with 10 CFR 50.59; and the design and licensing basis documentation was updated to reflect the change.

The inspectors used, in part, Nuclear Energy Institute 96-07, Guidelines for 10 CFR 50.59 Implementation, Revision 1, to determine acceptability of the completed evaluations, and screenings. The Nuclear Energy Institute document was endorsed by the NRC in Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, dated November 2000. The inspectors also consulted Part 9900 of the NRC Inspection Manual, 10 CFR Guidance for 10 CFR 50.59, Changes, Tests, and Experiments.

This inspection sample constituted one evaluation and twenty samples of screenings and/or applicability determinations as defined in Inspection Procedure 71111.17-04.

The inspectors could not review the minimum sample size of six evaluations because the licensee only performed one evaluation during the triennial sample period.

b. Findings

(1) Failure to Document Reviews Performed in 50.59 Screen for New Abnormal Operating Procedure

Introduction:

The inspectors identified a finding of very-low safety significance (Green), and an associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to document the review performed to conclude that a 50.59 Evaluation was not required. Specifically, the licensee failed to document the review performed to determine that installation of portable electric heaters in battery rooms would not have an adverse effect on the safety-related batteries.

Description:

Abnormal Operating Procedure (AOP) 904, Extreme Cold Weather

(<0 degrees Fahrenheit), was developed and issued by the licensee to provide actions that are required to be completed when there is a weather forecast for extreme cold weather. This procedures includes a step that directs the installation of portable electric heaters in the control building battery rooms to maintain battery temperatures above the Technical Specification 3.8.6 limit. The inspectors were concerned that installation of portable electric heaters in the battery rooms may have an adverse impact on safety-related batteries. Specifically, the inspectors were concerned that electric heaters may increase the probability of hydrogen ignition in the battery rooms. In addition, the inspectors were also concerned that portable electric heaters may result in spot heating, which may adversely impact battery cells and result in uneven heating of the battery room.

The inspectors reviewed the 10 CFR 50.59 screening performed by the licensee for this new procedure to determine if adverse impacts on safety related batteries were identified. The inspectors determined that the 50.59 screening performed for AOP 904 did not consider adverse effects of using portable electric heaters to maintain the temperature of safety-related batteries. Licensee Procedure EN-AA-203-1201, 10 CFR Applicability and 10 CFR 50.59 Screening Reviews, requires, in part, that reviews performed be documented if it is concluded that a 50.59 evaluation is not required. The inspectors determined that the licensee failed to document the reviews performed as required by procedure EN-AA-203-1201.

Analysis:

The inspectors determined that the licensees failure to document the reviews performed to conclude that a 50.59 evaluation was not required was contrary to Procedure EN-AA-203-1201, 10 CFR Applicability and 10 CFR 50.59 Screening Reviews, and was a performance deficiency (PD). Specifically, the 50.59 screening performed by the licensee did not contain any discussion which demonstrated that installation of portable electric heaters in battery rooms would not have an adverse effect on safety-related batteries. In addition, the licensee did not perform any other engineering analyses or evaluations that demonstrated that installation of portable electric heaters would not have an adverse effect on safety-related batteries.

The PD was determined to be more than minor, and a finding, because if left uncorrected, the PD would become a more significant safety concern. Specifically, installation of portable electric heaters in battery rooms may increase the probability of hydrogen ignition and challenge the ability of safety-related batteries to perform their safety function.

In accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 2 the inspectors determined the finding affected the Mitigating Systems cornerstone. As a result, the inspectors determined the finding could be evaluated using Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2 for the Mitigating Systems cornerstone. The finding screened as very-low safety significance (i.e., Green) because it did not result in the loss of operability or functionality of any structure, system, or component. Specifically, the licensee did not enter a condition that required the installation of portable electric heaters in the battery room per Procedure AOP 904.

The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current licensee performance.

Enforcement:

Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality be prescribed by documented procedures of a type appropriate to the circumstances and be accomplished in accordance with these procedures. The licensee established Procedure EN-AA-203-1201, Revision 6 as the implementing procedure for 10 CFR applicability and 10 CFR 50.59 screening reviews. Procedure EN-AA-203-1201, Footnote 3 states, If it is concluded that a 50.59 Evaluation is NOT required, document the reviews as indicated on this form or in accordance with local administrative procedures.

Contrary to the above, on December 10, 2007, the licensee failed to follow Footnote 3 of procedure EN-AA-203-1201. Specifically, the licensee failed to document the reviews performed to conclude that a 50.59 evaluation was not required for installation of portable electric heaters in battery rooms containing safety related batteries.

This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy because it was of very-low safety significance, and was entered into the licensees Corrective Action Program as Action Request 02102327. The licensees immediate corrective action was to sequester Procedure AOP 904 and provide guidance to the Operations Department to cease use of portable electric heaters per AOP 904.

(NCV 05000331/2016007-01, Failure to Document Reviews Performed in 50.59 Screen for New Abnormal Operating Procedure)

(2) Failure to Document 50.59 Evaluation for Updated Final Safety Analysis Report Change Concerning Radiological Dose Consequence Analysis Methodology

Introduction:

The inspectors identified a Severity Level IV, NCV of 10 CFR 50.59, Changes, Tests, and Experiments, having very-low safety significance (Green) for the licensees failure to document the basis for making a change to Updated Final Safety Analysis Report (UFSAR) Table 15.0-2 to allow the use of RADTRAD Version 3.03 for all Chapter 15 Accidents. Specifically, the licensee failed to demonstrate that the change to UFSAR Table 15.0-2 did not constitute a Departure from a Method of Evaluation described in the UFSAR and would have never required prior NRC review and approval.

Description:

In UFSAR Table 15.0-2 lists the Computer Codes & Methods of Evaluation used in UFSAR Chapter 15 Accident

Analysis.

As a part of the implementation of License Amendment Request (LAR) 261 the licensee updated UFSAR Table 15.0-2 to reflect the approval of RADTRAD Version 3.03 for Radiological Dose Consequence

Analysis.

The LAR 261 specifically approved the use of RADTRAD Version 3.03 only for the Control Rod Drop Accident (CRDA). However, the update to UFSAR Table 15.0-2 that was performed allowed the use of RADTRAD Version 3.03 for all accident analyses.

The inspectors reviewed the 10 CFR 50.59 screening performed for the update to UFSAR Table 15.0-2 and determined that the 10 CFR 50.59 screening did not address the use of RADTRAD Version 3.03 for accidents other than CRDA. Specifically, the 10 CFR 50.59 screening for the UFSAR update stated that LAR 261 bounded all of the changes made to the UFSAR. However, LAR 261 did not approve the use of RADTRAD Version 3.03 for accidents other than CRDA. The inspectors determined that the change to UFSAR Table 15.0-2 to allow use of RADTRAD Version 3.03 for all Chapter 15 Accidents was a change to an element of a method of evaluation described in the UFSAR. The 50.59 screening performed by the licensee failed to evaluate whether this change constituted a Departure from a Method of Evaluation described in the UFSAR. Therefore, the inspectors could not reasonably determine that this change was not a Departure from a Method of Evaluation described in the UFSAR that would have never required prior NRC review and approval.

Analysis:

The inspectors determined that the failure to evaluate whether the change to UFSAR Table 15-0.2 constituted a Departure from a Method of Evaluation was contrary to 10 CFR 50.59(d)(1) and was a PD. Specifically, the change to UFSAR Table 15.0-2 allowed the use of RADTRAD 3.03 for all UFSAR Chapter 15 Accidents based on the approval in LAR 261. However, LAR 261 only approved use of RADTRAD 3.03 for the CRDA. The licensee failed to evaluate whether using RADTRAD 3.03 on accidents other than the CRDA constituted a Departure from a Method of Evaluation described in the UFSAR.

The PD was determined to be more than minor, and a finding, because if left uncorrected, the PD had the potential to become a more significant safety concern.

Specifically, the inspectors could not reasonably determine that use of RADTRAD version 3.03 for all UFSAR Chapter 15 Accidents would not have increased the control room dose value during accidents.

In addition, the associated violation was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the significance determination process (SDP) because they are considered to be violations that potentially impede or impact the regulatory process. This violation is associated with a finding that has been evaluated by the SDP and communicated with an SDP color reflective of the safety impact of the deficient licensee performance. The SDP, however, does not specifically consider the regulatory process impact. Thus, although related to a common regulatory concern, it is necessary to address the 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 determined that finding could be evaluated using the SDP in accordance with Inspection Manual Chapter 0609, Significance Determination Process. Using 0609.04, Initial Characterization of Findings, Table 2 the inspectors determined that the finding affected the Barrier Integrity cornerstone. As a result, the inspectors evaluated the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 3 for the Barrier Integrity cornerstone.

The inspectors answered Yes to question C.1 in Exhibit 3 - Barrier Integrity Screening Questions. Specifically, the inspectors determined the finding only represented a degradation of the radiological barrier function provided for the control room.

In accordance with Section 6.1.d of the NRC Enforcement Policy this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very-low safety significance (i.e., green finding).

In accordance with Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Section 07.03.c, the inspectors did not assign a cross-cutting aspect to this violation because the violation and underlying technical finding was not indicative of current plant performance

Enforcement:

Title 10 CFR Part 50.59, Changes, Tests, and Experiments, Section (d)(1) requires the licensee to maintain records of changes in the facility, of changes in procedures, and of tests and experiments made pursuant 10 CFR 50.59(c).

These records must include a written evaluation which provides the bases for the determination that the change, test, or experiment does not require a license amendment.

Contrary to the above, on April 10, 2007, the licensee failed to document the basis for changing UFSAR Table 15.0-2 to allow the use of RADTRAD Version 3.03 for Radiological Dose Consequence. Specifically, the licensee did not evaluate whether use of RADTRAD Version 3.03 for accidents other than the CRDA was a Departure from a Method of Evaluation described in the UFSAR that would have required a license amendment.

This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy because it was a Severity Level IV violation and was entered into the licensees Corrective Action Program as Action Request 02105462. The licensees immediate corrective actions included performing a gap analysis between RADTRAD Version 3.02 & Version 3.03 to determine if any significant differences exist and to demonstrate that the radiological dose consequence using RADTRAD Version 3.03 would provide essentially the same results as Version 3.02. In addition, the licensee intends to update the UFSAR table 15.0-2 to accurately describe which RADTRAD Version is applicable to each accident analyzed in the UFSAR under Chapter 15.

(NCV 05000331/2016007-02; Failure to Document 50.59 Evaluation for Updated Final Safety Analysis Report Change Concerning Radiological Dose Consequence Analysis Methodology).

.2 Permanent Plant Modifications

a. Inspection Scope

The inspectors reviewed fifteen permanent plant modifications that had been installed in the plant during the last three years. This review included in-plant walk-downs for portions of the modified Emergency Service Water piping and pipe supports and the Residual Heat Removal Service Water piping and pipe supports in the pump house.

The modifications were selected based upon risk significance, safety significance, and complexity. The inspectors reviewed the modifications selected to determine if:

the supporting design and licensing basis documentation was updated; the changes were in accordance with the specified design requirements; the procedures and training plans affected by the modification have been adequately updated; the test documentation as required by the applicable test programs has been updated; and post-modification testing adequately verified system operability and/or functionality.

The inspectors also used applicable industry standards to evaluate acceptability of the modifications. The list of modifications and other documents reviewed by the inspectors is included as an Attachment to this report.

This inspection constituted fifteen permanent plant modification samples as defined in Inspection Procedure 71111.17-04.

b. Findings

No findings were identified.

4OA2 Problem Identification and Resolution

.1 Routine Review of Condition Reports

a. Inspection Scope

The inspectors reviewed several corrective action process documents that identified or were related to 10 CFR 50.59 evaluations and permanent plant modifications.

The inspectors reviewed these documents to evaluate the effectiveness of corrective actions related to permanent plant modifications and evaluations of changes, tests, and experiments. In addition, corrective action documents written on issues identified during the inspection were reviewed to verify adequate problem identification, and incorporation of the problems into the corrective action system. The specific corrective action documents that were sampled and reviewed by the inspectors are listed in the to this report.

b. Findings

No findings were identified.

4OA6 Management Meetings

.1 Exit Meeting Summary

The inspectors presented the inspection results to Mr. R. Murrell and other members of the licensee staff on February 29, 2016. The licensee personnel acknowledged the inspection results presented, and did not identify any proprietary content.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

T. Vehec, Site Vice President

S. Brown, Director, Site Engineering

B. Preston, Manager, Engineering Design

M. Davis, Licensing Manager

T. Weaver, Senior Licensing Engineer

L. Swenzinksi, Licensing Engineer

B. Murrell, Licensing Engineer

J. Swales, Design Engineering

U.S. Nuclear Regulatory Commission

C. Norton, Senior Resident Inspector

J. Steffes, Resident Inspector

LIST OF ITEMS

Opened and Closed

05000331/2016007-01 NCV Failure to Document Reviews Performed in 50.59 Screen for New Abnormal Operating Procedure (Section 1R17.1.b.(1))

05000331/2016007-02 NCV Failure to Document 50.59 Evaluation for UFSAR Change Concerning Radiological Dose Consequence Analysis Methodology (Section 1R17.1.b.(2))

Discussed

None LIST OF ACRONYMS USED ADAMS Agencywide Documents Access and Management System AOP Abnormal Operating Procedures CFR Code of Federal Regulations CRDA Control Rod Drop Accident IMC Inspection Manual Chapter LAR License Amendment Request NCV Non-Cited Violation NRC U.S. Nuclear Regulatory Commission PARS Public Available Records System PD Performance Deficiency UFSAR Updated Final Safety Analysis Report