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{{Adams
#REDIRECT [[IR 05000255/2015012]]
| number = ML15261A576
| issue date = 09/17/2015
| title = IR 05000255/2015012, on 03/23/2015 - 08/19/2015; Palisades Nuclear Plant; Operability Determinations and Functional Assessments. (Msh)
| author name = O'Brien K
| author affiliation = NRC/RGN-III/DRS
| addressee name = Vitale A
| addressee affiliation = Entergy Nuclear Operations, Inc
| docket = 05000255
| license number = DPR-020
| contact person =
| case reference number = EA-15-171
| document report number = IR 2015012
| document type = Inspection Report, Letter
| page count = 21
}}
See also: [[see also::IR 05000255/2015012]]
 
=Text=
{{#Wiki_filter:UNITED STATES
                            NUCLEAR REGULATORY COMMISSION
                                              REGION III
                                    2443 WARRENVILLE RD. SUITE 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 05000255/2015012
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 Commissions 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 NRCs Web site at http://www.nrc.gov/about-
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 safety
significance were discussed with NRC staff at the inspection exit meeting held on
 
A. Vitale                                        -2-
August 19, 2015. 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 inspector 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. OBrien, Director
                                                Division of Reactor Safety
Docket No. 50-255
License No. DPR-20
Enclosure:
IR 05000255/2015012
cc w/encl: Distribution via LISTSERV
 
          U.S. NUCLEAR REGULATORY COMMISSION
                          REGION III
Docket No.          50-255
License No.        DPR-20
Report No:          05000255/2015012
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
                                                        Enclosure
 
                                            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 of Documents Reviewed .................................................................................................. 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 Non-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
NRCs 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. An 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 licensees
        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
                                                  2
 
  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 licensees
  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
  licensees 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
                                            3
 
nonconforming condition associated with nine PCS welds susceptible to PWSCC
[IMC 310, Item P.2]. (Section 1R15)
                                      4
 
                                        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 licensees 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: An apparent violation (AV) of 10 CFR 50.9 was identified by the licensee,
      related to an apparent failure to provide information 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.
                                                5
 
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 licensees 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 - ADAMS Number ML14127A543 and NRC Regulatory
        Information Summary 2015-10 - ADAMS 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 licensees 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.
                                            6
 
* On September 4, 2014, the NRC issued a letter documenting the NRCs 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 licensees vendor provided a letter to the licensee which
  described the error in the vendors 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 licensees 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
  RR 4-21 (superseded RR 4-18) and provided corrections to the calculation and
  analysis that supported the original RR 4-18. Specifically, in Enclosure 2 of
                                    7
 
    *  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 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 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 licensees corrective actions
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
be 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 vendors 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 licensees 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.
                                          8
 
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 this 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.
An 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
                                          9
 
  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 licensees 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 licensees corrective actions for the AV discussed in
  the previous report section, the inspectors identified a separate performance deficiency
  and finding associated with the licensees 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 licensees 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 licensees 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
                                            10
 
[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
implemented 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 licensees 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-3600
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
                                          11
 
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 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, 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).
                                          12
 
    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
                                              13
 
                              SUPPLEMENTAL INFORMATION
                                  KEY POINTS OF CONTACT
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. OConnor, 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 DISCUSSED
Opened
05000255/2015012-01          AV      Inaccurate/Incomplete Information Submitted For Relief
                                      Request 4-18 (Section 1R15)
05000255/2015012-02          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
                                                                                      Attachment
 
                                    LIST OF ACRONYMS USED
ADAMS            Agencywide Documents 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
                                                    2
 
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 inspector 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. OBrien, Director
                                                                                Division of Reactor Safety
Docket No. 50-255
License No. DPR-20
Enclosure:
IR 05000255/2015012
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ADAMS Accession Number ML15261A576
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  OFFICE RIII                          RIII            RIII                                  RIII                                      RIII
  NAME          MHolmberg              DHills          MKunowski for EDuncan                  KLambert for RSkokowski DCurtis for KOBrien
  DATE          09/08/15              09/16/15          09/08/15                              09/16/15                                  09/16/15
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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. OBrien dated
SUBJECT: PALISADES NUCLEAR PLANT NRC INSPECTION REPORT 05000255/2015012
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