ML15261A576
| ML15261A576 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 09/17/2015 |
| From: | O'Brien K Division of Reactor Safety III |
| To: | Vitale A Entergy Nuclear Operations |
| References | |
| EA-15-171 IR 2015012 | |
| Download: ML15261A576 (21) | |
See also: IR 05000255/2015012
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION III 2443 WARRENVILLE RD. SUIT
E 210 LISLE, IL 60532
-4352 September 17, 2015
EA-15-171 Mr. Anthony Vitale
Vice President, Operations
Entergy Nuclear Operations, Inc.
Palisades Nuclear Plant
27780 Blue Star Memorial Highway
Covert, MI 49043
-9530 SUBJECT: PALISADES NUCLEAR PLANT NRC INSPECTION REPORT
Dear Mr. Vitale:
On August 19, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection consisting of an operability determination review at your Palisades Nuclear Plant. The enclosed report documents the results of this inspection, which were discussed on August 19, 2015, with members of your staff.
This inspection was an examination of activities conducted under your license as they relate to operability determinations and compliance with the Commission's rules and regulations and the conditions of your license. Within this area, the inspection involved examination of selected procedures, representative records and interviews with personnel.
The enclosed report presents the results of this inspection including an apparent violation which is being considered for escalated enforcement action in accordance with the NRC Enforcement Policy, which appears on the NRC's Web site at
-nrc/regulatory/enforcement/enforce
-pol.html. As described in Section 1R15 of this report, the apparent violation of 10 CFR 50.
9 , "Completeness and Accuracy of
Information
," relates to your failure to provide information to the NRC that was complete and accurate in all material respects in letter PNP 2014
-015, "Relief Request Number 4-18 - Proposed Alternative Use of Alternate ASME [American Society of Mechanical Engineers] Code Case N
-770-1 Baseline Examination
," submitted to the NRC on February 25, 2014. This issue resulted from an error in a calculation supporting the analysis results provided in your February 25, 2014
, letter, and
, once identified by your staff, was promptly reported to the NRC. This apparent violation is not a
current safety concern because your staff demonstrated an adequate basis for continued
operability of the nine affected primary coolant system welds.
Because the NRC has not made a final determination in this matter, no notice of violation is being issued for the apparent violation at this time. In addition, please be advised that the number and characterization of the apparent violation may change based on further NRC review. The NRC requires lasting and effective corrective actions for this issue and your corrective actions for the apparent violation and associated finding of very low safet
y significance were discussed with NRC staff at the inspection exit meeting held on
A. Vitale -2- August 19, 201 5. As a result, it may not be necessary to conduct a pre
-decisional enforcement conference (PEC) in order to enable the NRC to make an enforcement decision. In addition, since you identified the violation, and based on our understanding of your corrective actions, a civil penalty may not be warranted in accordance with Section 2.3.4 of the Enforcement Policy. The final decision will be based on you confirming on the license docket that the corrective actions previously described to the NRC staff have been or are being taken.
Before the NRC makes a final decision on this matter, you may choose to: (1) attend a PEC, where you can present
to the NRC your point of view on the facts and assumptions used to arrive at the apparent violation and assess its significance, or (2)
submit your position on the violation to the NRC in writing. If you request a PEC, it should be held within 30 days of
your receipt of this letter. Please contact Mr.
David Hills at (630) 829
-9733, and in writing, within
10 days from the issue date of this letter to notify the NRC of your intentions. If we have not heard from you within 10 days, we will continue with our enforcement decision.
If you choose to request a PEC, the conference will afford you the opportunity to provide your perspective on these matters and any other information that you believe the NRC should take into consideration before making an enforcement decision. The decision to hold a PEC does not mean that the NRC has determined that a violation has occurred or that enforcement action will be taken. This conference would be conducted to obtain information to assist the NRC in making an enforcement decision. The topics discussed during the conference may include information to determine whether a violation occurred, information to determine the significance
of a violation, information related to the identification of a violation, and information related to any corrective actions taken or planned. We encourage you to submit supporting documentation at least one week prior to the conference in an effort to make the conference more efficient and effective. If you choose to attend a PEC, it will be open for public observation. The NRC will issue a public meeting notice and press release to announce the conference.
If you decide to submit only a written response, it should be sent to the NRC within 30 days of your receipt of this letter. It should be
clearly marked as a "Response to An Apparent Violation in NRC Inspection Report (05000255/2015012; EA
-15-171)" and should include for the apparent violation:
(1) the reason for the apparent violation or, if contested, the basis for disputing the apparent violation; (2) the corrective steps that have been taken and the results achieved; (3) the corrective steps that will be taken; and (4) the date when full compliance will be achieved. Your response may reference or include previously docketed correspondence, if the correspondence adequately addresses the required response. If an adequate response is not received within the time specified or an extension of time has not been granted by the NRC, the NRC will proceed with its enforcement decision or schedule a PEC. In addition, based on the results of this inspection, one NRC
-identified finding of very low safety significance was identified. This finding involved a violation of NRC requirements. However, because of the very low safety significance and because the issue was entered into your
Corrective Action Program, the NRC is treating the violation as a Non-Cited Violation (NCV) in accordance with Section 2.3.2 of the NRC Enforcement Policy.
A. Vitale -3- If you contest the subject or severity of the NCV, 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 a copy to the Regional Administrator, Region III; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555
-0001; and the
NRC resident i
nspector at the Palisades Nuclear Plant. 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 the Palisades Nuclear Plant.
In accordance with Title 10 of the Code of Federal Regulations
(CFR) 2.390 of the NRC's "Rules of Practice," a copy of this letter, its
enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records System (PARS) component of NRC's Agencywide Documents Access and Management System (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading
-rm/adams.html
(the Public Electronic Reading Room).
Sincerely, /RA David Curtis Acting for/
Kenneth G. O'Brien, Director
Division of Reactor Safety
Docket No. 50
-255 License No. DPR
-20 Enclosure:
cc w/encl: Distribution via LISTSERV
Enclosure U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No. 50-255 License No
. DPR-20 Report No:
Licensee: Entergy Nuclear Operations, Inc.
Facility: Palisades Nuclear Plant
Location: Covert, MI
Dates: March 23 through August 19, 2015
Inspectors:
M. Holmberg, Reactor Inspector
A. Nguyen, Senior Resident Inspector
Approved by:
David E. Hills, Chief
Engineering Branch 1
Division of Reactor Safety
TABLE OF CONTENTS
SUMMARY ................................
................................................................
................................
. 2 REPORT DETAILS
................................
................................................................
.....................
5 1. REACTOR SAFETY ................................
................................................................
....... 5 1R15 Operability Determinations and Functional Assessments (71111.15)
..................
5 4. OTHER ACTIVITIES
................................
................................................................
.....13 4OA6 Management Meetings
................................
......................................................
13 SUPPLEMENTAL INFORMATION
................................
.............................................................
1 Key Points of Contact
................................
................................................................
.............
1 List of Items Opened, Closed, and Discussed
........................................................................ 1 List of Acronyms Used
................................
................................................................
............
2 List o f Documents Reviewed
................................
................................................................
.. 2
2 SUMMARY Inspection Report (IR) 05000255/2015012, 03/23/2015
-08/19/2015; Palisades Nuclear Plant; Operability Determinations and Functional Assessments.
This report covers a 5
-month period of inspection by the
senior resident inspector for the Palisades Nuclear Plant and a regional inspector. An
apparent violation was
identified by the licensee. Additionally, one Green finding was identified by the inspectors.
The finding was considered a N
on-Cited Violation (NCV) of U.S. Nuclear Regulatory
Commission (NRC) regulations.
The significance of inspection findings is indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red) and determined using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process
(SDP)," dated April
29, 2015. Cross-cutting aspects are determined using IMC 0310, "Aspects Within the Cross
-Cutting Areas," dated December 4, 2014. All violations of NRC requirements are dispositioned in accordance with the NRC's Enforcement Policy dated July 9, 2013.
The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG
-1649, "Reactor Oversight Process," Revision 5, dated February 2014.
Cornerstone: Initiating Events
TBD. A n apparent violation (AV) of Title 10 of the Code of Federal Regulations
(CFR) 50.9 was identified by the licensee, related to a failure to provide information that was complete and accurate in all material respects to the NRC in letter PNP 2014
-015 , "Relief Request (RR) Number 4
-18 - Proposed Alternative Use of Alternate ASME [American Society of Mechanical Engineers] Code Case N
-770-1 Baseline Examination." Specifically, in this document the licensee stated
, "In the unlikely case that crack initiation were to occur, crack growth calculations considering primary water stress corrosion cracking (PWSCC) as the failure mechanism demonstrate that the hot leg drain nozzle weldment satisfies ASME Code acceptance criteria for 60 effective full power years [EFPY] for a circumferential flaw, and more than 34 years for an axial flaw." However, this statement was not correct or accurate in that, the ASME Code acceptance criteria were not satisfied for 60 EFPY for a circumferential flaw and 34 years for an axial flaw, where correct information was 20 EFPY for a circumferential flaw, and 11.3 years for an axial flaw. This AV was not an immediate safety concern because the licensee
demonstrated an adequate basis for continued operability of the nine affected primary coolant system (PCS) welds. The licensee corrective actions for this AV included
completion of an operability evaluation, submittal of a corrected analysis to the NRC, and entering this issue into the Corrective Action Program (CAP) (CR-PLP-2015-03441). If the NRC was provided
with the correct information in letter PNP 2014
-015, where the affected welds satisfied ASME Code acceptance criteria (i.e., 75 percent through
-wall) for only 20 effective full power years for a circumferential flaw, and 11.3 years for an
axial flaw, the NRC would not likely have approved RR 4
-18 and, as a minimum, would have requested additional supporting analysis (e.g
., required substantial further inquiry). Further, the need for substantial further inquiry was illustrated by the licensee's
subsequent decision in RR
4-21 to abandon the prior analytical approach used in
RR 4-18. The inspectors evaluated the underlying technical issue in accordance with the SDP to determine the risk significance of this AV. The issue of concern was of more than minor significance because it was similar to the "not minor if" aspect of Example 3j in IMC 0612, Appendix
E, "Example of Minor Issues." Specifically, the erroneous
3 information provided in letter PNP 2014
-015 resulted in a condition in which there was a reasonable doubt on the operability of the systems and components that were the subject of the evaluation and dissimilar from the
"minor because" aspect of this example since the impact of the error for the operability of nine PCS welds was not minimal. In addition, the performance deficiency was determined to be more than minor because it was associated with the Initiating Event Cornerstone attribute of Equipment Performance and adversely affected the Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions. The inspectors evaluated the finding in accordance with IMC 0609, "Significance Determination Process," Attachment 0609.04, "Phase 1
- Initial Screening and Characterization of Findings," Table 3, for the Initiating Events Cornerstone
, and IMC 0609, Appendix A, "The SDP for Findings At
-Power." Because the licensee was able to demonstrate operability of the nine PCS welds susceptible to PWSCC, the inspectors answered "No" to questions A.1 and A.2, of Exhibit 1 , "Initiating Events Screening Questions," identified in Appendix A of IMC 609 and, as a result, the finding screened as having very low safety significance (Green).
No cross-cutting aspect was assigned because this
Green finding was identified by the licensee. (Section 1R15)
Green. An NRC-identified finding of very low safety significance and an associated
NCV of Title 10 of the Code of Federal Regulations
(CFR), Part 50, Appendix B , Criterion V, "Instructions, Procedures and Drawings
," was identified for the licensee's failure to adhere to
the site procedure for performing operability determinations during the evaluation of a nonconforming condition associated with nine primary coolant system (PCS) welds susceptible to primary water stress corrosion cracking (PWSCC). The licensee's corrective actions for this finding included completion of an operability determination in accordance with the site operability procedure to include a new analysis which demonstrated the AMSE Code acceptance criteria would continue to be met for the affected welds during the remainder of the operating cycle. The licensee entered the failure to comply with the operability procedure into the CAP (CR-PLP-2015-03434). This finding was determined to be more than minor because it was similar to the "not
minor if" aspect of Example 3j in IMC 0612, Appendix E, "Example of Minor Issues
," because the errors in operability evaluation CA
-1 of CR-PLP-2015-01239 resulted in a condition in which there was a reasonable doubt on the operability of the systems and components that
were the subject of the evaluation and dissimilar from the "minor because" aspect of this example since the impact of the errors on the operability evaluation was not minimal. In addition, the performance deficiency was determined to be more than minor because it was associated with the Initiating Event Cornerstone attribute of Equipment Performance and adversely affected the Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions. The inspectors evaluated the finding in accordance with IMC 0609, "Significance Determination Process," Attachment 0609.04, "Phase 1
- Initial Screening and Characterization of Findings," Table 3, for the Initiating Events Cornerstone and
IMC 0609, Appendix A, "The SDP for Findings At
-Power." Because the licensee was able to demonstrate operability of the nine PCS welds susceptible to PWSCC, the inspectors answered "No" to questions A.1 and A.2, of Exhibit 1
, "Initiating Events Screening Questions," identified in Appendix A of IMC 609 and, as a result, the finding screened as having very low safety significance (Green). This finding has a cross
-cutting aspect in Evaluation for the Problem Identification and Resolution cross
-cutting area since the licensee failed
to thoroughly evaluate the impact on operability of a
4 nonconforming condition associated with nine PCS welds susceptible to
PWSCC [IMC 310, Item P.2].
(Section 1R15)
5 REPORT DETAILS
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R15 Operability Determinations and Functional Assessments
(71111.15)
a. Inspection Scope
The inspectors reviewed the following issue:
Calculation error affecting flaw evaluation of nine primary coolant system (PCS) welds susceptible to primary water stress corrosion cracking (PWSCC) submitted to the NRC in letter PNP 2014
-015 , "Relief Request (RR) Number 4
-18 - Proposed Alternative Use of Alternate American Society of Mechanical Engineers (ASME) Code Case N
-770-1 Baseline Examination."
The inspectors selected this operability issue based on the risk significance of the associated components and systems. The inspectors evaluated the technical adequacy of the evaluations to ensure that Technical Specification (TS) operability was properly justified and the subject component or system remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the
TS and the Updated Final Safety Analysis Report to the licensee's evaluations to determine whether the components or
systems were operable. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations. Additionally, the inspectors reviewed a sample of corrective action documents to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluation. Documents reviewed are listed in the Attachment to this report.
This operability inspection constituted one sample as defined in Inspection Procedure 71111.15-05. b. Findings .1 Inaccurate/Incomplete Information Submitted For Relief Request 4
-18 Introduction: A n apparent violation (AV) of 10
CFR 50.9 was identified by the licensee, related to an apparent
failure to provide informati
on that was complete and accurate in all material respects to the NRC in letter PNP
2014-015. Specifically, in this document the licensee stated
, "In the unlikely case that crack initiation were to occur, crack growth calculations considering PWSCC as the
failure mechanism demonstrate that the hot leg drain nozzle weldment satisfies ASME Code acceptance criteria for 60 effective full
power years (EFPY) for a circumferential flaw, and more than 34 years for an axial flaw." However, this statement was not correct or accurate in that, the ASME Code acceptance criteria were not satisfied for 60 EFPY for a circumferential flaw and 34 years for an axial flaw, where correct information was 20 EFPY for a circumferential flaw, and 11.3 years for an axial flaw. This AV was not an immediate safety concern because the licensee demonstrated an adequate basis for continued operability of the nine affected PCS welds.
6 Description: In March of 2015, the licensee notified NRC staff, that information provided to the NRC in letter PNP 2014
-015 requesting NRC approval to defer examination of nine PCS welds was not accurate because of an error made in a calculation used to support the analysis results documented in this letter. On March 23, 2015, the inspectors initiated a review of this issue to determine the impact of this error on the operability of the nine affected PCS welds and to assess the licensee's corrective actions. On February 25, 2014, the licensee submitted a letter PNP 2014
-015 to the NRC requesting approval to defer volumetric examination of nine PCS welds based in part on
the evaluations of postulated weld cracks that demonstrated ASME Code acceptance criteria were met. In this letter, the licensee stated that the ASME Code acceptance criteria would continue to be met for a postulated circumferential flaw for 60 EFPY and more than 34 EFPY for a postulated axial flaw. On February 26, 2015, the licensee was notified by its vendor of a nonconservative error in a calculation used to support this
analysis. Specifically, the vendor had erroneously applied the normal operating pressure load which introduced a bending moment into the hot leg pipe wall rather than
an expected radial and axial expansion loads typical of internally applied pressure in the piping. In particular, the induced bending moment created a compressive (i.e., less tensile) stress behavior in and around the inside of the nozzle
-to-pipe weld. As a result, the erroneously applied pressure load reduced the radial and hoop tensile stresses at
the weld inside diameter rather than increasing them. The net effect of this error on the analysis results was that the ASME Code acceptance criteria were met for only 20 EFPY
for a postulated circumferential and 11.3 EFPY for a postulated axial flaw. Palisades EFPY of operation had already exceeded both of these values.
The inspectors developed a timeline of activities related to this issue as discussed below. During the January 2014 refueling outage, the NRC identified nine PCS welds
susceptible to PWSCC which had not been volumetrically examined by the licensee as required by NRC regulations (reference NRC Inspection Report 05000255/2014002
- ADAM S Number ML14127A543 and NRC Regulatory Information Summary 2015
-10 - ADAM S Number ML15068A131).
On February 25, 2014, the licensee submitted a letter PNP 2014
-015 "Relief Request Number RR 4
-18 - Proposed Alternative, Use of Alternate ASME Code Case N-770-1 Baseline Examination" to the NRC. In this letter, the licensee requested the NRC to approve deferral of volumetric examinations on nine PCS
welds based in part, on evaluations of postulated weld cracks that demonstrated
that ASME Code acceptance criteria would be maintained.
On March 6, 2014, in letter PNP 2014
-028, the licensee submitted vendor calculations to the NRC that were used to support the licensee's conclusions documented in RR 4
-18 including calculation 1200895.306, Revision 0.
On March 12, 2014, the NRC granted verbal approval of RR 4
-18 until the next refueling outage scheduled for the fall of 2015.
7 On September 4, 2014, the NRC issued a letter documenting the NRC's basis for approval of RR 4
-18 (e.g., NRC safety evaluation).
On February 26, 2015, the
licensee was notified by its vendor that an error was made in a vendor calculation supporting RR 4
-18. On February 27, 2015, the licensee documented in CR
-PLP-2015-0928 that an error was made in a vendor calculation supporting RR 4
-18 and notified the Palisades Senior Resident Inspector.
On March 3 and March 19, 2015, during routine licensing conference calls with NRC staff, the licensee notified the NRC Project Manager for Palisades in the Office of Nuclear Reactor Regulation (NRR) that an error was made in a vendor calculation supporting RR 4
-18. On March 6, 2015, the licensee's vendor provided a letter to the licensee which
described the error in the vendor's calculation and the impact on the analysis results discussed in RR 4
-18. On March 23, 2015, in CA
-4 of CR-PLP-2015-0928, the licensee identified five vendor documents submitted to the NRC that contained errors and assigned an action to interface with the NRC to determine which of these corrected documents were to be resubmitted to the NRC.
On March 23, 2015, the inspectors and staff in the Office of NRR conducted a
tele-conference meeting with the licensee to determine the impact of the vendor calculation error supporting RR 4
-18 and to evaluate the licensee's planned corrective actions. The licensee reported that the error in the vendor calculation was nonconservative because a corrected analysis resulted in a reduction in the time (by approximately a factor of two) until a postulated PWSCC would reach
75 percent through
-wall. On March 24, 2015, the NRC concerns from the March 23, 2015
, call, prompted the licensee to initiate CA
-1 of CR-PLP-2015-1239 to document a basis for operability of the nine PCS welds affected by the calculation error. The inspectors identified that the licensee had not previously completed an operability evaluation for this condition because it was not recognized as a nonconformance with the license basis (see next report section). On March 31, 2015, the licensee completed an operability evaluation CA
-1 of CR-PLP-2015-1239 for this issue and determined that the affected welds were operable.
On May 22, 2015, the licensee submitted a letter PNP 2015
-037 , "Relief Request Number RR 4
-21 - Proposed Alternative, Use of Alternate ASME Code Case
N-770-1 Baseline Examination," to the NRC. In this letter, the licensee identified that a discrepancy was discovered in a calculation that supported relief request
RR 4-18, requested approval for an alternative analysis/basis as described in
-18) and provided corrections to the calculation and analysis that supported the original RR 4
-18. Specifically, in Enclosure 2
of
8 PNP 2015-037, the licensee stated
, "The erroneously applied pressure caused an unbalanced pressure load, which introduced a bending moment into the hot leg pipe wall rather than an expected radial and axial expansion typical of internally applied pressure in the piping. In particular, the induced moment tended to create a compressive (i.e., less tensile) stress behavior in and arou
nd the inside of the nozzle
-to-pipe weld. As a result, the erroneously applied pressure reduced the radial and hoop tensile stresses at the weld inside diameter
rather than increase them." And "In the unlikely case that crack initiation were to occur, crack growth calculations considering PWSCC as the failure mechanism demonstrate that the hot leg drain nozzle weldment satisfies ASME Code
acceptance criteria (i.e., 75 percent through
-wall) for 20 EFPY for a circumferential flaw, and 11.3 years for an axial flaw." The licensee entered the failure to provide complete and accurate information to the NRC as part of RR 4
-18 into the Corrective Action Program (CAP) (CR
-PLP-2015-03441) and
initiated an apparent cause evaluation. The licensee's corrective action
s completed for this issue included an operability evaluation, and submittal of a corrected analysis to the NRC.
Analysis: The inspectors determined that the failure to provide information to the NRC that was complete and accurate in all material respects in letter PNP 2014
-015 requesting NRC approval to defer examination of nine PCS welds that appears not
to b e in accordance with
10 CFR 50.9 and a performance deficiency. Additionally, the inspectors determined that
the licensee had reasonable opportunity to foresee and correct the inaccurate/incomplete information discussed above during owner acceptance review of the vendor's calculations prior to submitting this information to the NRC.
The inspectors reviewed this issue in accordance with IMC 0612, Appendix B , "Issue Screening ," dated September 7, 2012. Because the apparent failure to provide complete and accurate information to the NRC had the potential to impede or impact the regulatory process, the finding was evaluated in accordance with NRC Enforcement Policy for traditional enforcement items and the underlying technical issue was evaluated using the SDP to determine the risk significance of this issue. Specifically, this AV 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, or actual consequences. Thus, although related to a common regulatory concern, it is necessary to address the apparent violation and finding using different processes to correctly reflect both the regulatory importance of the apparent violation and the safety significance of the associated finding.
If the NRC was provided with the correct information in letter PNP 2014
-015, where the affected welds satisfied ASME Code acceptance criteria (i.e., 75 percent through
-wall) for only 20 EFPY for a circumferential flaw, and 11.3 years for an axial flaw, the NRC would not likely have approved RR 4
-18 and, as a minimum, would have requested additional supporting analysis (e.g.
, required substantial further inquiry). The need for substantial further inquiry was illustrated by the licensee's subsequent decision in RR 4
-21 to abandon the prior analytical approach used in RR 4
-18 that relied on a closed form analysis (e.g., SmartCrack Software Program) and instead changed to a more sophisticated finite element analysis approach using an ANSYS software program to model crack growth behavior in evaluation of the structural and leakage integrity at the limiting weld.
9 The inspectors evaluated the underlying technical issue in accordance with the SDP to determine the risk significance of this AV. The issue of concern was of more than minor significance because it was similar to the "not minor if" aspect of Example 3j in IMC 0612, Appendix E, "Example of Minor Issues." Specifically, the erroneous information provided in letter PNP 2014
-015 resulted in a condition in which there was a reasonable doubt on the operability of the systems and components that were the subject of the evaluation and dissimilar from the "minor because" aspect of this example since the impact of the error for the operability of nine PCS welds was not minimal. In addition, the performance deficiency was determined to be more than minor because it was associated with the Initiating Event Cornerstone attribute of Equipment Performance
and adversely affected the Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions. The inspectors evaluated the finding in accordance with IMC 0609, "Significance Determination Process," Attachment 0609.04, "Phase 1
- Initial Screening and Characterization of Findings,"
Table 3, for the Initiating Events Cornerstone and IMC 0609, Appendix A, "The SDP for Findings At
-Power," dated June 19, 2012. Because the licensee was able to demonstrate operability of the nine PCS welds susceptible to PWSCC, the inspectors answered "No" to questions A.1 and A.2, of Exhibit 1
, "Initiating Events Screening Questions," identified in Appendix A of IMC 609 and, as a result, the finding screened as having very low safety significance (Green). No cross
-cutting aspect was assigned because thi
s Green finding was identified by the licensee.
Enforcement: Title 10 of the Code of Federal Regulations (10 CFR) 50.9(a), "Completeness and Accuracy of Information," requires that "Information provided to the Commission by an applicant for a license or by a licensee or information required by statute or by the Commission's regulations, orders, or license conditions to be maintained by the applicant or the licensee shall be complete and accurate in all material respects."
In Attachment 1
, "Relief Request Number RR 4
-18 Proposed Alternative" of letter PNP
2014-015 "RR Number 4
-18 - Proposed Alternative Use of Alternate ASME Code Case
N-770-1 Baseline Examination" in the section titled "Structural Evaluation," the licensee stated , in part, "ASME Code acceptance criteria are satisfied for 60 EFPY for a circumferential flaw, and more than 34 years for an axial flaw assuming crack initiates at day one. Using hot leg crack growth rate and temperature."
In Attachment 3
, "Structural Integrity Associates, Inc. Memorandum - Evaluation of the Palisades Nuclear Plant Hot Leg Drain Nozzle for Primary Water Stress Corrosion Cracking" of letter PNP 2014
-015 "RR Number 4
-18 - Proposed Alternative Use of Alternate ASME Code Case N
-770-1 Baseline Examination," in
the section titled "Conclusions" the licensee stated
, in part , "In the unlikely case that crack initiation were to occur, crack growth calculations considering PWSCC as the failure mechanism demonstrate that the hot leg drain nozzle weldment satisfies ASME Code acceptance criteria for 60 EFPY for a circumferential flaw, and more than 34 years for an axial flaw."
A n AV of Code of Federal Regulations (10 CFR) 50.9(a), "Completeness and Accuracy of Information," has been identified, as it appears that the information in
letter PNP 2014
-015 provided to the Commission on February 25, 2014, was not complete and accurate in all material respects because the ASME Code acceptance criteria would not have been met for 60 EFPY for a circumferential flaw and 34 years for an axial flaw, where correct information was 20 EFPY for a circumferential flaw, and 11.3 years for an axial
10 flaw. This change in the analysis results represented a significant reduction in the time to reach the ASME Code acceptance criteria limits and as such, was information considered material to the NRC in the review and approval of RR 4
-18. This was not an immediate safety concern because the licensee demonstrated an adequate basis for
continued operability of the affected welds. The licensee corrective actions for this issue included; completion of an operability evaluation, submittal of a corrected analysis to the NRC, and entering this issue into the CAP (CR-PLP-2015-03441). (AV 05000255/2015012
-01; Inaccurate/Incomplete Information Provided For Relief
Request 4-18). .2 Operability Evaluation Not Performed in Accordance with Station Procedure
Introduction
- The inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR 50, Part 50, Appendix B
, Criterion V
, "Instructions, Procedures and Drawings
," for the licensee's failure to adhere to the site procedure for performing operability determinations during the evaluation of a nonconforming condition
associated with nine PCS welds susceptible to PWSCC.
Description
- During review of the licensee's corrective actions for the AV discussed in the previous report section, the inspectors identified a separate performance deficiency and finding associated with the licensee's failure to follow the site procedure for evaluating the operability of nine PCS welds.
On February 27, 2015, the licensee was notified by its vendor of a nonconservative error in a vendor calculation 1200895.306 which determined the residual stress profile in the limiting PCS weld susceptible to PWSCC. This calculation had been submitted to the NRC on March 6, 2014, in support of a RR 4
-18 (discussed in the previous
section) and was used to support the licensee's conclusion that a postulated axial crack would not reach 75 percent through
-wall until more than 34 EFPY and 60 EFPY for a postulated circumferential crack. In calculation 1200895.306, the licensee's vendor erroneously applied the normal operating pressure load creating an unbalanced pressure load, which introduced a bending moment into the hot leg pipe wall model rather than an expected radial and axial expansion load typical of internally applied pressure in the piping. In particular, the induced moment tended to create a compressive (i.e., less tensile) stress behavior in and around the inside of the nozzle
-to-pipe weld. As a result, the erroneously applied pressure reduced the radial and hoop tensile stresses at the weld inside diameter rather than increasing them. The licensee evaluated the effect of this
non-conservative vendor calculation error in CR-PLP-2015-00928 and documented this issue as administrative in nature with proposed corrective actions to revise the affected
calculation and update the associated engineering change package. However, the licensee had not assigned an action to complete an operability evaluation of the nine PCS welds susceptible to PWSCC that had not been volumetrically examined to determine the extent of cracking within these welds. Because
the corrected flaw growth evaluation of a postulated PWSCC resulted in a time to reach a through
-wall leakage condition that was less than the current accumulated EFPY of operation, the inspectors were concerned for the lack of a basis to demonstrate that it was acceptable to continue operation with the nine PCS welds at risk for leakage or failure induced by PWSCC.
Procedure EN
-OP-104 "Operability Determination Process" defined an operability evaluation as a "Technical analysis and associated conclusions, including a prescriptive
description of any required Compensatory Measures, regarding Operability of a TS SSC
11 [structure system or component].
" The operability determination process is an activity affecting quality and the licensee identified procedure
EN-OP-104 as "quality related" which is a procedure required by the Entergy Quality Assurance Program Manual (QAPM). The QAPM is implemented through the use of approved procedures (e.g. policies, directives, procedures, instructions, or other documents) which provide written guidance for the control of quality related activities
and provide for the development of documentation to provide objective evidence of compliance.
In the QAPM the licensee stated that "Procedures that implement the QAPM are approved by the management
responsible for the applicable quality function. These procedures are to reflect the QAPM and work is to be accomplished in accordance with them."
Step 5.5.5.f of EN
-OP-104 required the licensee to identify the applicable current license basis (CLB) requirements for the SSC including review of other CLB documents such as
safety evaluations. In CR
-PLP-2015-00928 the licensee appropriately identified the CLB requirement for the nine PCS welds which included the NRC safety evaluation approving
RR 4-18 (reference: NRC Letter dated September 4, 2014, ADAMS Number
ML14223B226). However, the licensee incorrectly assumed that the NRC had not relied
on the results of the vendor calculations submitted in the review and approval of this safety evaluation. Therefore, the licensee did not identify this issue as a nonconformance with the CLB and hence did not properly accomplish Step 5.5.6.a of
EN-OP-104 which stated , "Evaluate component and system conformance with applicable requirements of the CLB."
On March 23, 2015, the inspectors reviewed CR
-PLP-2015-00928 and identified that the licensee staff failed to recognize the vendor calculation error as a nonconforming condition with respect to the CLB for the nine affected PCS welds. Step 3.16 of EN
-OP-104 defined a nonconforming condition as "A condition of a SSC that involves a failure to meet the CLB." In this case, the nonconservative calculation error shortened the time available until a PWSCC could reach 75 percent through
-wall which adversely effected the CLB for the nine PCS welds as evaluated by the NRC during review of RR 4-18. Consequently, the licensee had not complied with step 5.3 of EN
-OP-104 , which stated that "Operability should be determined immediately upon discovery (i.e., Immediate Determination) without delay and in a controlled manner using the best information available." The inspectors requested that the licensee identify the basis for operability
of the nine affected PCS welds which did not conform to the CLB as established in
RR 4-18. The inspectors' concern prompted the licensee to document this issue in CR-PLP-2015-01239 and complete an immediate operability evaluation. The licensee also implement ed a corrective action to document additional supporting evaluations/analysis in a prompt operability evaluation in accordance with procedure EN
-OP-104. On March 31, 2015, the licensee completed the prompt operability evaluation under
CA-1 of CR-PLP-2015-01239. However, the inspectors identified that the licensee had not established an adequate basis for a prompt operability that would cover the remaining operating cycle. Specifically, the licensee's operability evaluation relied on a leak-before-break type of analysis without identification of margins to prevent thru
-wall leakage and did not follow the ASME Code Section XI methods
(e.g., Article IWB
-36 00 "Analytical Evaluation of Flaws") to quantify factors of safety (e.g., margins) to protect against a sudden/rapid failure (e.g.
, structural integrity). Without application of the ASME Code methods, the operability evaluation was not consistent with procedure EN-OP-104 step 5.5.6(d) which required evaluation of the SSC against the applicable codes and standards requirements for operability and step 5.11.17 "ASME Class 1, 2, 3 Piping
12 Flaw Evaluation and Resolution
," which stated that "When Flaws are acceptable per the ASME Code acceptance standards, then structural integrity is assured and the SSC is OPERABLE." Additionally, the operability evaluation was not consistent with the NRC policy for operation with flawed piping as identified in Appendix C.11
, "Flaw Evaluation" of IMC 0326
Operability Determinations and Functionality Assessments for Conditions Adverse to Quality or Safety" which stated
, "Satisfaction of Code acceptance standards is the minimum necessary for operability of Class 1 pressure boundary components because of the importance of the safety function being performed." The licensee staff
stated that they had not followed the operability procedure for flawed piping welds because they did not have any known flaws. However, the nine PCS welds were susceptible to PWSCC and the CLB as established in the NRC safety evaluation of
RR 4-18 required the licensee to presume the presence of flaws (e.g., cracks) because volumetric examinations had not been completed to identify the extent of cracking
present in these welds.
On June 3, 2015, the licensee completed a revision to
operability evaluation CA
-1 of CR-PLP-2015-01239 to correct errors previously
identified by the inspectors and established an adequate basis for prompt operability for the remaining portion of the operating cycle. Specifically, in the revised operability evaluation, the licensee assumed PWSCC were present in the affected welds and
documented a new analysis which demonstrated the ASME Code acceptance criteria would continue to be met for the affected welds during the remainder of the operating cycle. The licensee entered the failure to comply with the operability procedure into the
CAP (CR-PLP-2015-03434). Analysis: The inspectors determined that the failure to adhere to the site procedure for performing operability determinations during the evaluation of a nonconforming condition
associated with nine PCS welds susceptible to PWSCC
was contrary to 10 CFR 50, Part 50, Appendix B, Criterion V
, and a performance deficiency.
This finding was determined to be more than minor because it was similar to the "not minor if" aspect of Example 3j in IMC 0612, Appendix E, "Example of Minor Iss
ues ," because the errors in Operability Evaluation CA
-1 of CR-PLP-2015-01239 resulted in a condition in which there was a reasonable doubt on the operability of the systems and
components that were the subject of the evaluation and dissimilar from the "minor because" aspect of this example since the impact of the errors on the Operability Evaluation was not minimal. In addition, the performance deficiency was determined to be more than minor because it was associated with the
Initiating Event Cornerstone
attribute of Equipment Performance and adversely affected the Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions.
The inspectors evaluated the finding in accordance with IMC 0609, "Significance Determination Process," Attachment 0609.04, "Phase 1
- Initial Screening and Characterization of Findings," Table 3, for the Initiating Events Cornerstone and IMC 0609, Appendix A, "The SDP for Findings At
-Power," dated June 19, 2012. Because
t he licensee was able to demonstrate operability of the nine PCS welds susceptible to PWSCC, the inspectors answered "No" to questions A.1 and A.2, of Exhibit 1
, "Initiating Events Screening Questions," identified in Appendix A of IMC 609 and, as a result, the finding screened as having very low safety significance (Green).
13 This finding has a cross
-cutting aspect in Evaluation
for the Problem Identification and Resolution cross
-cutting area since the licensee failed to thoroughly evaluate the impact
on operability of a nonconforming condition associated with nine PCS welds susceptible to PWSCC [IMC 310, Item P.2].
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion V
, "Instructions, Procedures and Drawings" requires, in part, that activities affecting
quality be prescribed and accomplished by procedures.
The operability determination process (an activity affecting quality) was described in procedure EN OP 104 "Operability Determination Process" and the licensee identified this procedure as "quality related" which is a
procedure required by the QAPM.
Procedure EN
-OP-104 , Step 5.5.6.a stated
, "Evaluate component and system conformance with applicable requirements of the CLB."
Procedure EN
-OP-104 Step 5.5.6.d stated
, "Evaluate the SSC condition against the applicable codes and standards requirements for operability." And Step 5.11.17
, "ASME Class 1, 2, 3 Piping Flaw Evaluation and Resolution
," stated , in part , "When Flaws are acceptable per the ASME Code acceptance standards, then structural integrity is assured and the SSC is OPERABLE."
Contrary to the above, on March 31, 2015, in CA
-1 of CR-PLP-2015-01239 the licensee failed to evaluate these welds (e.g., components) for conformance with the CLB as described in the NRC safety evaluation approving RR
4-18 (reference ADAMS Number ML14223B226) and failed to evaluate these welds against the applicable ASME
Code for operability. Corrective actions for this finding included completion of an operability
determination on June 3, 2015 in accordance with the site operability procedure to include a new analysis which demonstrated the ASME Code acceptance criteria would continue to be met for the affected welds during the remainder of the operating cycle. Because this violation was of very low safety significance, was corrected on June 3, 2015, and entered into the CAP (CR
-PLP-2015-03434), this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000255/2015012
-02; Operability Evaluation Not Performed in Accordance with Station Procedure).
4. OTHER ACTIVITIES
4OA6 Management Meetings
.1 Exit Meeting Summary
On August 19, 2015, the inspectors presented the inspection results to Mr. R
. Craven, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors confirmed that none of the potential report input discussed was considered proprietary.
ATTACHMENT: SUPPLEMENTAL INFORMATION
Attachment
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTAC
T Licensee D. Corbin, Operations Manager
R. Craven, Acting General Manager Plant Operations
T. Davis, Regulatory Assurance
J. Hardy, Regulatory Assurance Manager
D. Mannai, Fleet Regulatory Assurance Senior Manager
D. Nestle, Radiation Protection Manager
K. O'Connor, Design Engineering Manager
B. Sova, Engineering Supervisor
U.S. Nuclear Regulatory Commission
E. Duncan, Chief, Reactor Projects Branch 3
J. Collins, Senior Materials Engineer, Division of Engineering, Office of Nuclear Reactor Regulation
LIST OF ITEMS OPENED, CLOSED, AND DISCUS
SED Opened 05000255/2015012
-01 AV Inaccurate/Incomplete Information Submitted For Relief Request 4-18 (Section 1R1
-02 NCV Operability Evaluation Not Performed in Accordance with Station Procedure (Section 1R15)
Closed 05000255/2015012
-02 NCV Operability Evaluation Not Performed in Accordance with Station Procedure (Section 1R15)
Discussed None
2 LIST OF ACRONYMS USE
D ADAMS Agencywide Document
s Access and Management System
ASME American Society of Mechanical Engineers
AV Apparent Violation
CAP Corrective Action Program
CFR Code of Federal Regulations
CLB Current License Basis
EFPY Effective Full Power Years
IMC Inspection Manual Chapter
NCV Non-Cited Violation
NRC U.S. Nuclear Regulatory Commission
NRR Office of New Reactor Regulation
PARS Publicly Available Records System
PCS Primary Coolant System
PEC Pre-Decisional Enforcement Conference
PWSCC Primary Water Stress Corrosion Cracking RR Relief Request
SDP Significance Determination Process
SSC Structure, System, or Component
TBD To Be Determined
TS Technical Specification
LIST OF DOCUMENTS REVIEWED The following is a partial list of documents reviewed during the inspection. Inclusion on this list does not imply that the NRC inspector reviewed the documents in their entirety, but rather that selected sections or portions of
the documents were evaluated as part of the overall inspection effort. Inclusion of a document on this list does not imply NRC acceptance of the document or any part of it, unless this is stated in the body of the inspection report.
1R15 Operability Determinations and Functionality Assessments
- CR-PLP-2015-03441, dated August 18, 2015
- CR-PLP-2015-03434, dated August 18, 2015
- CR-PLP-2015- 00928, dated February 27, 2015
- CR-PLP-2015-02427, dated June 11, 2015
- CR-PLP-2015- 01239, Corrective Action 1 Operability Evaluation, dated March 31, 2015
- CR-PLP-2015-01239, Corrective Action 1 Operability Evaluation, dated June 3, 2015
- Letter PNP 2014
-015 , RR Number 4
-18 - Proposed Alternative Use of Alternate ASME Code Case N-770-1 Baseline Examination, dated February 25, 2014. - Letter PNP 2015
-037 , Relief Request Number RR 4
-21 - Proposed Alternative, Use of Alternate ASME Code Case N
-770-1 Baseline Examination, dated May 22, 2015.
- Procedure EN
-OP-104 , Operability Determination Process, Revision 8
A. Vitale -3- If you contest the subject or severity of the NCV, 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 a copy to the Regional Administrator, Region III; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555
-0001; and the
NRC resident i
nspector at the Palisades Nuclear Plant. 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 the Palisades Nuclear Plant.
In accordance with Title 10 of the Code of Federal Regulations
(CFR) 2.390 of the NRC's "Rules of Practice," a copy of this letter, its
enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records System (PARS) component of NRC's Agencywide Documents Access and Management System (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading
-rm/adams.html
(the Public Electronic Reading Room).
Sincerely, /RA David Curtis Acting for/
Kenneth G. O'Brien, Director Division of Reactor Safety
Docket No. 50
-255 License No. DPR
-20 Enclosure:
cc w/encl: Distribution via LISTSERV
DISTRIBUTION
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ADAMS Accession Number ML15261A576
Publicly Available
Non-Publicly Available
Sensitive Non-Sensitive To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copy
OFFICE RIII RIII RIII RIII RII I NAME MHolmberg DHills MKunowski for EDuncan KLambert for RSkokowski
DCurtis for KO'Brien DATE 09/08/15 09/16/15 09/08/15 09/16/15 09/16/15 OFFICIAL RECORD COPY
OE Concurrence provided via e
-mail from Kyle Hanley on 09/15/15
. NR Concurrence provided via e
-mail from Nestor Feliz
-Adorno on 09/15/15
.
Letter to Mr. Anthony Vitale from Mr. Kenneth G. O'Brien dated
SUBJECT: PALISADES NUCLEAR PLANT NRC INSPECTION REPORT
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