IR 05000483/2016002

NRC Integrated Inspection Report 05000483/2016002 and Notice of Violation

ML16225A577
Person / Time
Site:	Callaway
Issue date:	08/12/2016
From:	Taylor N H NRC/RGN-IV/DRP/RPB-B
To:	Diya F Union Electric Co
Taylor N H
References
IR 2016002
Preceding documents:	ML16258A043 (14 September 2016)
Download: ML16225A577 (81)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION REGION IV 1600 E. LAMAR BLVD.

ARLINGTON, TX 76011-4511 August 12, 2016 Mr. Fadi Diya, Senior Vice President and Chief Nuclear Officer Union Electric Company P.O. Box 620 Fulton, MO 65251 SUBJECT: CALLAWAY PLANT - NRC INTEGRATED INSPECTION REPORT 05000483/2016002 AND NOTICE OF VIOLATION

Dear Mr. Diya,

On June 30, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Callaway Plant. On July 19, 2016 , the NRC inspectors discussed the results of this inspection with you and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report.

NRC inspectors documented five findings of very low safety significance (Green) in this report.

Four of these findings involved violations of NRC requirements

. The NRC evaluated these violation s in accordance Section 2.3.2.a of the NRC Enforcement Policy, which appears on the NRC's Web site at http://www.nrc.gov/about

-nrc/regulatory/enforcement/enforce

-pol.html. The NRC is treating three violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the NRC Enforcement Policy

. We determined that one violation did not meet the criteria to be treated as an NCV because compliance has not been restored within a reasonable period after the violation was originally identified. Specifically, NRC inspectors identified and documented a noncompliance in NRC Integrated

Inspection Report

05000483/2010006 dated December 17, 2010. This finding was a violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, for the failure to take timely corrective actions for water hammer transients and corrosion on essential service water system components. As of the end of this inspection (more than 65 months later), compliance had still not been restored. The inspectors determined that the licensee did not provide an adequate justification for the delay.

You are required to respond to this letter and should follow the instructions specified in the enclosed Notice of Violation (Notice) when preparing your response. If you have additional information that you believe the NRC should consider, you may provide it in your response to the Notice.

The NRC's review of your response to the Notice will also determine whether further enforcement action is necessary to ensure your compliance with regulatory requirements.

If you contest the NCVs or their significance you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S.

Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV, 1600 East Lamar Blvd., Arlington, Texas 76011

-4511; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the Callaway Plant. If you disagree with a cross

-cutting aspect assignment or a finding not associated with a regulatory requirement 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 IV; and the NRC resident inspector at the Callaway Plant

. In accordance with 10 CFR 2.390, "Public Inspections, Exemptions, Requests for Withholding,"

a copy of this letter, its enclosure, and your response will be available electronically for public inspection in the NRC

's Public Document Room or from the Publicly Available Records (PARS) component of the NRC's Agencywide Document s Access and Management System (ADAMS).

ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading

-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA David Proulx Acting for/

Nicholas H. Taylor, Branch Chief Project Branch B Division of Reactor Projects Docket No. 50

-483 License No. NPF

-30 Enclosure s: 1. Notice of Violation 2. Inspection Report 05000483/2016002 w/ Attachment 1: Supplemental Information Attachment 2: Request for Information cc w/ encl: Electronic Distribution

ML16225A577 SUNSI Review By: DLP ADAMS Yes No Non-Sensitive Sensitive Publicly Available Non-Publicly Available Keyword: NRC-002 OFFICE SRI/DRP/B RI/DRP/B C:DRS/OB C:DRS/PSB2 C:DRS/EB1 C:DRS/EB2 NAME THartman MLangelier VGaddy RDeese TFarnholtz SGraves SIGNATURE /RA/ /RA/ /RA/ /RA/ /RA/ /RA/ DATE 8/8/16 8/8/16 8/1/2016 8/1/2016 8/1/2016 8/1/2016 OFFICE C:DRS/IPAT SRI:DRS/EB2 SRI:DRP/D TL:ACES D:DRP C:DRP/B NAME THipschman JDrake JJosey JKramer TWPruett NTaylor SIGNATURE /RA/ /RA/ /RA/ /RA/ /RA/ /RA DProulx Acting, for/

DATE 8/1/2016 8/5/16 8/9/16 8/3/2016 8/12/16 8/12/16 Letter to Fadi Diya from Nicholas H. Taylor August 12, 2016 SUBJECT: CALLAWAY PLANT

- NRC INTEGRATED INSPECTION REPORT 05000483/2016002 AND NOTICE OF VIOLATION DISTRIBUTION

Regional Administrator (Kriss.Kennedy@nrc.gov)

Deputy Regional Administrator (Scott.Morris@nrc.gov) DRP Director (Troy.Pruett@nrc.gov)

DRP Deputy Director (Ryan.Lantz@nrc.gov)

DRS Director (Anton.Vegel@nrc.gov)

DRS Deputy Director (Jeff.Clark@nrc.gov)

Senior Resident Inspector (Thomas.Hartman@nrc.gov) Resident Inspector (Michael.Langelier@nrc.gov) Branch Chief, DRP/B (Nick.Taylor@nrc.gov) Senior Project Engineer, DRP/B (David.Proulx@nrc.gov

) Project Engineer, DRP/B (Steven.Janicki@nrc.gov)

Administrative Assistant (Dawn.Yancey@nrc.gov)

Public Affairs Officer (Victor.Dricks@nrc.gov

) Project Manager (John.Klos@nrc.gov) Team Leader, DRS/TSS (Thomas.Hipschman@nrc.gov)

RITS Coordinator (Marisa.Herrera@nrc.gov

) ACES (R4Enforcement.Resource@nrc.gov)

Regional Counsel (Karla.Fuller@nrc.gov

) Technical Support Assistant (Loretta.Williams@nrc.gov)

Congressional Affairs Officer (Jenny.Weil@nrc.gov)

RIV Congressional Affairs Officer (Angel.Moreno@nrc.gov) RIV/ETA: OEDO (Jeremy.Bowen@nrc.gov)

RIV RSLO (Bill.Maier@nrc.gov)

ACES (R4Enforcement.Resource@nrc.gov)

ROPreports.Resource@nrc.gov ROPassessment.Resource@nrc.gov

- 1 - Enclosure 1 NOTICE OF VIOLATION Union Electric Company Docket No. 50

-483 Callaway Plant License No. NPF

-30 During an NRC inspection conducted June 6-30, 2016, a violation of NRC requirements was identified. In accordance with the NRC Enforcement Policy, the violation is listed below:

10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," requires, in part, that conditions adverse to quality are promptly identified and corrected.

Contrary to the above, from November 2010 through June 2016, the licensee failed to promptly correct a condition adverse to quality. Specifically, the licensee failed to adequately resolve water hammer and corrosion issues which were previously identified by the NRC as non

-cited violation 05000483/2010006

-01. The failure to resolve these issues resulted in subsequent safety

-related equipment failures.

This violation is associated with a Green Significance Determination Process finding.

Pursuant to the provisions of 10 CFR 2.201, Union Electric Company is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, DC 20555-0001 with a copy to the Regional Administrator, Region IV, 1600 East Lamar Blvd., Arlington, Texas 76011

-4511 and a copy to the NRC Senior Resident Inspector at the facility that is the subject of this Notice, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a

"Reply to a Notice of Violation

," and should include

(1) the reason for the violation, or, if contested, the basis for disputing the violation or severity level, (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 previous docketed correspondence if the correspondence adequately addresses the required response. If an adequate reply is not received within the time specified in this Notice, an order or a Demand for Information may be issued as to why the license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time.

If you contest this enforcement action, you should also provide a copy of your response, with the basis for your denial, to the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001. Because your response will be made available electronically for public inspection in the NRC Public Document Room or from the NRC's Agencywide Documents Access and Management System (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading

-rm/adams.html, to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the public without redaction. If personal privacy or proprietary information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you must specifically identify the portions of your response that you seek to have withheld and provide in detail the bases for your claim of withholding (e.g., explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.390(b) to support a request for

- 2 - withholding confidential commercial or financial information).

If safeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.

In accordance with 10 CFR 19.11, you may be required to post this Notice within two working days of receipt.

Dated this 12 th day of August 2016

- 1 - Enclosure 2 U.S. NUCLEAR REGULATORY COMMISSION REGION IV Docket: 05000 483 License: NPF-30 Report: 05000483/2016002 Licensee: Union Electric Company Facility: Callaway Plant Location: Junction Highway CC and Highway O Steedman, MO Dates: April1 through June 30, 2016 Inspectors:

T. Hartman, Senior Resident Inspector M. Langelier, P.E., Resident Inspector J. Drake, Senior Reactor Inspector P. Hernandez, Health Physicist J. Josey, Senior Resident Inspector, Comanche Peak R. Kopriva, Senior Reactor Inspector J. O'Donnell, Health Physicist Approved By:

Nicholas H. Taylor Chief, Project Branch B Division of Reactor Projects

- 2 -

SUMMARY

IR 05000483/2016002; 04/01/2016

- 06/30/2016; Callaway Plant, Equipment Alignment, Heat Sink Performance , Operability Determinations and Functionality Assessments, Problem Identification and Resolution, Follow-up of Events and Notices of Enforcement Discretion

. The inspection activities described in this report were performed between April and June 30, 2016, by the resident inspectors at the Callaway Plant and inspectors from the NRC's Region IV office. Five finding s of very low safety significance (Green) are documented in this report.

Four of these findings involved violation s of NRC requirements. The significance of inspection findings is indicated by their color (Green, White, Yellow, or Red), which is determined using Inspection Manual Chapter 0609, "Significance Determination Process." Their cross-cutting aspects are determined using Inspection Manual Chapter 0310, "Aspects within the Cross-Cutting Areas." Violations of NRC requirements are dispositioned in accordance with the NRC Enforcement Policy.

The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG

-1649, "Reactor Oversight Process."

Cornerstone: Initiating Events

Green.

The inspectors reviewed a self-revealed finding for the licensee's failure to follow the plant procedure for foreign material exclusion. Specifically, after finding foreign material (broken cable ties) within the main generator excitation transformer, established as a foreign material exclusion Level 2 area, the licensee failed to determine the reason for the foreign material and enter the issue into the corrective action program for resolution as required by Procedure AP A-ZZ-00801, "Foreign Material Exclusion," Revision 32. The licensee's failure to follow the plant procedure for foreign material exclusion was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, after identifying several broken cable ties on the floor inside a foreign material exclusion Level area the licensee did not determine the reason for the foreign material nor enter the condition into the corrective action program as required by Procedure APA-ZZ-00801. Because the licensee failed to understand what caused the cable tie degradation, a subsequent cable tie failure resulted in a plant trip.

Using Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At-Power," dated June 19, 2012, the finding was determined to be of very low safety significance because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect of training in the human performance area because the organization did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, several groups within the licensee's organization were unaware the excitation transformer cabinet was classified as a foreign material exclusion Level area nor the requirements if foreign material is found within the foreign material exclusion area [H.9]. (Section 4OA 3)

Cornerstone: Mitigating Systems

Green.

The inspector s identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the licensee's failure to account for the essential service water pipe stresses caused by pressure fluctuations of the known column closure water hammer phenomenon. The licensee failed to properly account for essential service water piping membrane stress and impact loads as required by the 1974 ASME Code,Section III, paragraphs ND-3112.4 and ND-3111. Specifically, the licensee's design calculations for the essential service water system did not account for the pressure fluctuations caused by a known column closure water hammer phenomenon that occurs during a loss of off-site power or load sequencer testing. The licensee completed a prompt operability determination assuring the system was operable under the current conditions and was completing engineering evaluations of the data collected to demonstrate the operability of the system under design conditions. The licensee entered this issued into the corrective action program as Callaway Action Request s 201603472 and 201603819. The inspectors determined that the licensee's failure to account for the pressure fluctuations caused by a known column closure water hammer phenomenon in the design calculations for the essential service water system was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

Using Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At-Power," dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding:

(1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2)did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4)does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with the licensee's maintenance rule program.

This finding has a cross-cutting aspect of conservative bias in the human performance area because the licensee failed to demonstrate that a proposed action was safe in order to proceed, rather than unsafe in order to stop.

Specifically, when the licensee recognized that the column separation water hammer phenomenon was occurring in the essential service water system, they only applied the forces to the containment coolers, not the entire system

[H.14]. (Section 1R04)

Green.

The inspectors identified a non-cited violation of 10 CFR 50.55a, "Codes and Standards," for the licensee's failure to repair various ASME Code Class 3 components in accordance with ASME C ode ,Section XI requirements.

Specifically, the licensee did not follow the applicable ASME Code requirements when making repairs to various components in the ASME Code Class 3 essential service water system

. The licensee reasonably determined the essential service water system remained operable, and completed the necessary repairs and testing to restore compliance with ASME Code. The licensee entered this issue into their corrective action program as Callaway Action Requests 201603640 and 201604282.

The inspectors determined that the programmatic failure to repair various ASME Code Class 3 components in the essential service water system in accordance with ASME Code was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At-Power," dated June 19, 2012, inspectors determined that this finding was of very low safety significanc e (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2)did not represent a loss of system and/or function, (3)did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4)does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with the licensee's maintenance rule program.

Specifically, the licensee performed a historical system health review and reasonably determined the essential service water system remained operable because periodic system walkdowns by the system owner and shiftly rounds by operations had not identified significant system leaks, and the appropriate repairs and testing were completed on the affected components. This finding has a cross-cutting aspect of training in the human performance area because the organization did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, the licensee failed to ensure training of the personnel was adequate to recognize that the repair of the leaks constituted repairs in accordance with ASME Code,Section XI and thus failed to include the necessary ASME testing requirements in the work performance packages to ensure adequate performance of an activity which affected testing of a safety-related modification/repair to risk-significant systems, and thereby ensure nuclear safety

[H.9]. (Section 1R07)

Green.

The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for the licensee's failure to perform an adequate operability assessment when a degraded or nonconforming condition was identified. Specifically, after the licensee identified that a severe water hammer transient would occur following a loss of off-site power, the licensee generated an operability evaluation that relied on judgement and inaccurate information which failed to establish a reasonable expectation of operability. Following questions from inspectors the licensee determined that this judgement was not correct and performed a new evaluation to ensure operability of the essential service water system. The licensee entered this issue into their corrective action program as Callaway Action Request 201605488.

The licensee's failure to properly assess and document the basis for operability when a sever e water hammer occurred in the essential service water system was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, severe water hammer transients in the essential service water system due to a loss of off-site power, result in a condition where structures, systems, and components necessary to mitigate the effects of accidents may not have functioned as required. Using Inspection Manual Chapter 0609, Appendix A, "The Significance

Determination Process (SDP) for Findings At-Power," dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding:

did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic event, and (1)was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2)did not represent a los s of system and/or function, (3)did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4)does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with the licensee's maintenance rule program. This finding has a cross-cutting aspect of conservative bias in the human performance area because the licensee failed to demonstrate that a proposed action was safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee's use of unsupported judgement and incorrect data resulted in an evaluation that failed to demonstrate a reasonable expectation of operability [H.14]. (Section 1R15)

Green.

The inspectors identified a cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," associated with the licensee's failure to take timely corrective action for a previously identified condition adverse to quality. Specifically, the licensee failed to adequately resolve water hammer and corrosion issues that were previously identified by the NRC as non-cited violation 05000483/2010006

-01 and the failure to resolve these issues resulted in subsequent safety-related equipment failures. The licensee performed an operability determination that established a reasonable expectation of operability pending implementation of corrective actions.

The licensee entered this issue into their corrective action program as Callaway Action Request 201604440. The licensee's failure to take timely and adequate corrective actions to correct a condition adverse to quality was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correct water hammer and corrosion issu e resulted in the licensee declaring safety-related room coolers and chillers inoperable until an analysis of system operability was completed. This affected their capability to respond to initiating events to prevent undesirable consequences Using Inspection Manual Chapter 0609, Appendix A, "The Significance Determinatio n Process (SDP) for Findings At-Power," dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2)did not represent a loss of system and/or function, (3)did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4)does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with the licensee's maintenance rule program. This finding has a cross-cutting aspect of resources in the human performance area because the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety. Specifically, by failing to address water hammer and corrosion issues, station management failed to ensure that the essential service water system was available and adequately maintained to respond during a loss of off-site power event

[H.1]. (Section 4OA2.3)

PLANT STATUS

Callaway began the inspection period at 86 percent power while coasting down at the end of t he operating cycle and on April 2, 2016, the licensee shut the plant down to start Refueling Outage 21. The reactor was restarted on May 9. On May 14, at approximately 90 percent power (during power ascension

), the plant reduced power to approximately 65 percent to address a main feedwater pump issue. The licensee repaired the feedwater pump on May 15 and recommenced power ascension. The plant returned to 100 percent power on May 16. The plant remained at full power for the remainder of the inspection period. REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity 1 R 01 Adverse Weather Protection (71111.01)

.1 Summer Readiness for Off

-site and Alternate AC Power Systems

a. Inspection Scope

On June 7, 2016 , the inspectors completed an inspection of the station's off

-site and alternate-ac power systems. The inspectors inspected the material condition of these systems, including transformers and other switchyard equipment to verify that plant features and procedures were appropriate for operation and continued availability of off-site and alternate

-ac power systems. The inspectors reviewed outstanding work orders and open Callaway action requests for these systems. The inspectors walked down the switchyard to observe the material condition of equipment providing off

-site power sources

.

The inspectors verified that the licensee's procedures included appropriate measures to monitor and maintain availability and reliability of the off

-site and alternate-ac power systems. These activities constituted one sample of summer readiness of off

-site and alternate

-ac power systems, as defined in Inspection Procedure 71111.01.

b. Findings

No findings were identified.

.2 Readiness for Impending Adverse Weather Conditions

a. Inspection Scope

On April 26, 2016, the inspectors completed an inspection of the station's readiness for impending adverse weather conditions. The inspectors reviewed plant design features, the licensee's procedures to respond to severe weather including thunderstorms, tornadoes and high winds , and the licensee's implementation of these procedures. The inspectors evaluated operator staffing and accessibility of controls and indications for those systems required to control the plant.

These activities constituted one sample of readiness for impending adverse weather conditions, as defined in Inspection Procedure 71111.01

b. Findings

No findings were identified.

1 R 04 Equipment Alignment (71111.04)

Partial Walk

-Down

a. Inspection Scope

The inspectors performed partial system walk

-downs of the following risk

-significant systems: May 24 , 2016, train A motor-driven auxiliary feedwater system June 2, 2016, train B class 1E switchgear June 8, 2016, train A essential service water June 9, 2016, train B essential service water The inspectors reviewed the licensee's procedures and system design information to determine the correct lineup for the systems. They visually verified that critical portions of the trains were correctly aligned for the existing plant configuration.

These activities constituted four partial system walk

-down samples as defined in Inspection Procedure 71111.04.

b. Findings

Introduction.

The inspectors identified a Green non

-cited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the licensee's failure to account for the essential service water pipe stresses caused by pressure fluctuations of the known column closure water hammer phenomenon.

Description.

With the current essential service water system design, every loss of off-site power at Callaway would result in a water column separation and subsequent re-pressurization by the loss of normal service water pumps and the sequencing start of the essential service water pumps. This phenomenon was not specifically described in the licensee's Updated Final Safety Analysis Report, however, it had been clearly identified in previous Callaway Action Requests 199800739, 199800740, 199800741, 200207750, 200404532, 200703197, 200703247, 200703257, 200703491, 200810348, 200810384, 200811050, 201003386, 201109846, 201303346, 201303370, 201303451, 201303502, 201303702, 201303736, 201407222, 201407245, 201407246, 201407248, 201602824, 201603472, 201603484, 201604058, and 201604063. This system characteristic was also described in Callaway's response to NRC Generic Letter 96-06, "Assurance of Equipment Operability and Containment Integrity during Design

-Basi s Accident Conditions," January 28, 1997. Additionally, there was external operating experience concerning water hammer phenomena and the impact on system piping.

Callaway is designed to ASME Code,Section III Nuclear Power Components, 1974 and 1974 winter addenda and ANSI B31.1 1973 piping code including the 1973 summer addendum. Piping analyses are performed to ensure that design Class II and III pipi ng systems perform their safety

-related functions during plant normal, upset , and faulted conditions. Pipes are subject to various loading conditions like pressures, d ead load, thermal, earthquake, and seismic/thermal anchor motions. The 1974 ASME Code,Section III , paragraph ND-3112.4, "Design Allowable Stress Values," part c states, in part, The wall thickness of a component computed by these rules shall be determined so that the maximum direct membrane stress due to any combination of loadings that are expected to occur simultaneously does not exceed the maximum allowable stress permitted at the temperature that is expected to be maintained in the metal under the condition of loading being considered.

Section III , paragraph ND-3111, "Loading Criteria," of the ASME Code, states in part, "The loading that shall be taken into account in designing a component shall include, but are not limited to, the following: -

(b) Impact loads, including rapidly fluctuating pressures."

Calculation 0096-020-CALC-01, Revision 0, "Callaway Water Hammer Load Calculation," Section 2.0 states in part, ... both Wolf Creek and Callaway are SNUPPS plants, many similarities exist. This calculation compares the conditions which can affect the impact velocity and the amount of air in the system, and adjusts the results from the Wolf Creek pressure vs. time data to account for those differences.

Even though Callaway recognized the similarities between Wolf Creek and their unit, they failed to reevaluate their essential service water when Wolf Creek recognized that their initial assumptions regarding water hammer phenomena were incorrect. WCN005-PR-0, a report from ENERCON, which addressed water hammer phenomena in the essential service water system, stated on page 6, The results shown in the Table in Section 5.1 of the ALTRAN Report 96225-TR02 were evaluated by an ENERCON structural expert. His opinion was that the loads shown were significant enough in every case to warrant further detailed analysis. This analysis requires the generation of a detailed FTH (Force Time History) that would result from the CCWH (column closure water hammer) generated in the ESW (essential service water) for a LOOP (loss of off

-site power) event. The report recommended that these FTH's would then be evaluated using a structural piping program and the results added to the existing stresses. Ultimately a new stress analysis of record would be generated. This would be a revision of the existing one. Modifications to supports may be required to qualify the system.

The analysis later stated, "To perform the reanalysis for the startup of the ESW pumps following a LOOP requires that Force Time Histories (FTH) be generated. These are required for the structural analysis."

The ALTRAN report referenced by ENERCON was report number 09-0223-TR-001 , Revision 0. This report, on page 6 of 14, stated in part, "The water hammer pressures calculated are to be used for preliminary structural assessment of the piping system's ability to withstand this loading and to determine if a more detailed force time history needs to be generated." On page 7 the report continued, "Experience has shown that the concerns resulting from water hammer events are:

(1) Over-pressure of pipes and components, e.g.

, ruptured tubes in heat exchangers, and

(2) Pipe and component nozzle stress due to bending moments created by the CCWH force time history (FTH)."

Despite the internal and external operating experience, the licensee only updated the design calculation for the containment coolers to include the pressures associated with the water hammer phenomena, but did not included these stresses in the design calculations for the remainder of the essential service water system. The basic engineering disposition written to address the potential effects of water hammer impact loads on the structural integrity of the pressure boundary did not include the pressure stresses induced in the pipe due to the water hammer phenomenon. It stated, in part, This Basic Engineering Disposition is to document that the potential effects of water hammer impact loads on the structural integrity of the pressure boundary have been evaluated for piping affected by pitting corrosion. Because water hammer pressure waves are of short duration and are self

-limiting (secondary) loads, assuring that the pitted pipe meets ASME Boiler and Pressure Vessel Code (Code) requirements for design loads is sufficient to conclude that the pressure boundary has sufficient margin to withstand impact from water hammer.

This engineering evaluation fail ed to meet the requirements of ASME Code Section III, paragraph ND-3111, "Loading Criteria,", which states in part, "The loading that shall be taken into account in designing a component shall include, but are not limited to, the following: ...

(b) Impact loads, including rapidly fluctuating pressures." In addition, operating experience at Callaway has consistently demonstrated that the pressure boundary lacks sufficient margin to withstand the impact from the water hammer as documented in the multiple Callaway action requests concerning system leaks after a water hammer event has occurred.

Although this was a deficiency affecting the design and qualification of the essential service water system, the licensee was able to demonstrate that the operability and function of the essential service water system had not been lost because the leaks that occurred were less than the allowable losses from the ultimate heat sink. The spray from the leaks did not adversely impact any other equipment, and the components affected maintained structural integrity.

Analysis.

The inspector s determined that the licensee's failure to account for the pressure fluctuations caused by a known column closure water hammer phenomenon in

the design calculations for the essential service water system was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

Using Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At

-Power," dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding:

(1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2)did not represent a loss of system and/or function, (3)did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out

-of-service for longer than their technical specification allowed outage time, and (4)does not represent an actual loss of function of one or more non

-technical specification trains of equipment designated as high safety significant for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with the licensee's maintenance rule program. This finding has a cross

-cutting aspect of conservative bias in the human performance area because the licensee failed to demonstrate that a proposed action was safe in order to proceed, rather than unsafe in order to stop. Specifically, when the licensee recognized that the column separation water hammer phenomenon was occurring in the essential service water system, they only applied the forces to the containment coolers, not the entire system [H.14].

Enforcement.

Title 10 CFR Part 50 Appendix B, Criterion III, "Design Control," states, in part, that for those structures, systems and components to which this appendix applies

, design control measures shall provide for verifying or checking the adequacy of designs. Contrary to the above, from June 4, 1985 , to the present, for the safety

-related essential service water system, to which 10 CFR Part 50 applies, the licensee failed to provide for verifying or checking the adequacy of designs. Specifically, the licensee did not include the pressures induced by the water hammer phenomenon in the design calculation for the essential service water system as required by the 1974 ASME Code, which the licensee is committed to follow.

The licensee performed a historical system health review and reasonably determined the essential service water system remained operable because periodic system walkdowns by the system owner and shiftly rounds by operations had not identified significant system leaks, and the appropriate repairs and testing were completed on the affected components. In addition, the licensee conducted an instrumented run of the system simulating a loss of off

-site power and collected data on the pressure spikes experienced by the system. Following the completion of the test the licensee conducted a system walkdown to inspection for indications of damage to the system. Based on the results of this evolution, the licensee completed a prompt operability determination assuring the system was operable under the current conditions, and was completing engineering evaluations of the data collected to demonstrate the operability of the system under design conditions.

This violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the Enforcement Policy because it was of very low safety significance

, and was entered into the licensee's corrective action program as Callaway Action Requests 201603472 and 201603819: NCV 05000483/2016002

-01, "Failure to Account for Water Hammer Stresses in Essential Service Water System Calculations."

1 R 05 Fire Protection (71111.05)

Quarterly Inspection

a. Inspection Scope

The inspectors evaluated the licensee's fire protection program for operational status and material condition. The inspectors focused their inspection on five plant areas important to safety:

May 12, 2016, train B battery and switchboard rooms (C-15) Jun e 2, 2016, train A electrical penetration room (A-18) June 3, 2016, boric acid tank rooms (A

-3) June 9 , 2016 , train A control room air conditioning room (A

-22) June 9, 2016, train A battery and switchboard rooms (C

-16) For each area, the inspectors evaluated the fire plan against defined hazards and defense-in-depth features in the licensee's fire protection program. The inspectors evaluated control of transient combustibles and ignition sources, fire detection and suppression systems, manual firefighting equipment and capability, passive fire protection features, and compensatory measures for degraded conditions.

These activities constituted five quarterly inspection samples, as defined in Inspection Procedure 71111.05.

b. Findings

No findings were identified.

1 R 07 Heat Sink Performance (71111.07)

a. Inspection Scope

The inspectors completed an inspection of the readiness and availability of risk-significant heat exchangers. The inspectors verified the licensee used the industry standard periodic maintenance method outlined in EPRI NP-7552 for the heat exchangers. Additionally, the inspectors walked down the heat exchanger s to observe the performance and material condition and/or verified that the heat exchangers were correctly categorized under the Maintenance Rule and were receiving the required maintenance.

April 3, 2016 , emergency core cooling system room coolers June 9, 2016, control room chillers These activities constitute d completion of two heat sink performance annual review sample s, as defined in Inspection Procedure 71111.07.

b. Findings

Introduction.

The inspectors identified a Green non-cited violation of 10 CFR 50.55a, "Codes and Standards," for the licensee's failure to repair various ASME Code Class components in accordance with ASME Code,Section XI requirements. Specifically, the licensee did not follow the applicable ASME Code requirements when making repairs to various components in the ASME Code Class 3 essential service water system.

Description.

The inspectors identified a programmatic issue with the licensee's inservice inspection and repair program because the engineering department personnel lacked adequate training and knowledge of the ASME Code to recognize activities that constituted repair activities per ASME Section XI. Specifically, the licensee had been repairing leaking tubes on various ASME Code Class 3 room coolers (SGL09B

- B Safety Injection Pump Room Cooler, SGL10A

- A Residual Heat Removal Pump Room Cooler, SGL10B

- B Residual Heat Removal Pump Room Cooler, and SGL13B

- B Containment Spray Pump Room Cooler) as a simple maintenance evolution, and failed to recognized that this constituted a repair activity per ASME Code,Section XI. The maintenance activities of concern were repairs to plug tube leaks which consisted of cutting a tube in order to remove a defect (pinhole), then mechanically installing (no brazing or welding) a Swagelok cap to plug the tube. Use of Swagelok caps to repair heat exchanger tube leaks is allowed by ASME Code and licensee procedures. These jobs were planned and performed as a maintenance activity in accordance with applicable licensee procedures.

Callaway is currently committed to the 2007 Edition/2008 Addenda of ASME Code,Section XI. ASME Code,Section XI, IWA-4120(b)(7)exempts ASME Class 2 and 3 mechanical tube plugging; however, the repairs to these components are considered an ASME Code,Section XI Repair/Replacement Activity. Per footnote 1 in IWA-4110 alterations are considered a repair/replacement activity per Section XI of ASME Code. This is because the tubes that had the Swagelok fittings installed still see system pressure: flow through the tube was not isolated. Therefore

, the pressure boundary was altered and the licensee is required to ensure it meets the requirements for ASME Code Class 3 pressure boundaries.

The physical work that was performed met the requirements of Section XI. Safety-related Swagelok caps were installed and ASME Code,Section III (the construction code) sections ND-3646 and ND-3674.1(e) allow the use of caps, so the repairs met the applicable construction code requirements

. The licensee did not consider the work as a repair activity per ASME Code,Section XI, therefore, requirements were not documented in the work packages and were not completed. These requirements were:

ANII notification Traceability of code pressure retaining parts Performance of require d pressure test

- VT-2 The licensee documented these deficiencies under Callaway Action Request 201603640, verified and documented the use of code pressure retaining parts, and completed the required VT

-2 pressure tests to correct these issues.

The repair performed on SGL13A (Containment Spray Pump A Room Cooler) utilized brazing to build up base metal of a pinhole leak. This resulted in a repair that was not an approved method by the ASME Code,Section XI. To correct this condition, the licensee generated Job 16002356-500, "Repair Tubing that was Improperly Repaired under Job 10506915." This job was completed in accordance with ASME Code requirements and a successful VT-2 was performed. In addition, the engineering department received training on ASME Code repair recognition and requirements.

Analysis.

The inspectors determined that the programmatic failure to repair various ASME Code Class 3 components in the essential service water system in accordance with ASME Code was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

Using Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At

-Power," dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding: (1)was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2)did not represent a loss of system and/or function, (3)did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out

-of-service for longer than their technical specification allowed outage time, and (4)does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant for greater than

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with the licensee's maintenance rule program.

Specifically, the licensee performed a historical system health review and reasonably determined the essential service water system remained operable because periodic system walkdowns by the system owner and shiftly rounds by operations had not identified significant system leaks, and the appropriate repairs and testing were completed on the affected components. This finding has a cross

-cutting aspect of training in the human performance area because the organization did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, the licensee failed to ensure training of the personnel was adequate to recognize that the repair of the leaks constituted repairs in accordance with ASME Code,Section XI and thus failed to include the necessary ASME testing requirements in the work performance packages to ensure adequate performance of an activity which affected testing of a safety

-related modification/repair to risk-significant systems, and thereby ensure nuclear safety

[H.9].

Enforcement

. Title 10 CFR 50.55a, "Codes and Standards," requires, in part, that safety-related pressure vessels, piping, pumps and valves, and their supports must meet the requirements applicable to components that are classified as ASME Code Class 3. Contrary to the above, as of April 18, 2016, the licensee failed to ensure that safety-related pressure vessels, piping, pumps and valves, and their supports must meet the requirements applicable to components that are classified as ASME Code Class 3. Specifically, the licensee failed to complete repairs to various ASME Code Class 3 components in the essential service water system because the engineering department did not recognize that correcting tube leakage constituted a repair activity per ASME Code,Section XI. The licensee has completed the applicable testing requirements for the repairs as part of the planned corrective actions. The licensee implemented immediate correction actions to enter this issue into the corrective action program for resolution. The licensee also completed the necessary repairs and testing to restore compliance with ASME Code. This violation is being treated as a non

-cited violation, consistent with Section 2.3.2.a of the Enforcement Policy because it was of very low safety significance, and was entered into the licensee's corrective action program as Callaway Action Requests 201603640 and 201604282: NCV 05000483/2016002

-02, "Failure to Meet Applicable ASME Code Requirements for Repairs to Components in the Essential Service Water System." 1 R 08 Inservice Inspection Activities (71111.08)

The activities described below constitute completion of two inservice inspection sample s , as defined in Inspection Procedure 71111.08.

.1 Non-destructive Examination Activities and Welding Activities

a. Inspection Scope

The inspectors directly observed the following nondestructive examinations:

SYSTEM WELD IDENTIFICATION EXAMINATION TYPE Auxiliary Feedwater System Report Number 5 010-16-0057 Condensate Storage Tank to Auxiliary Feedwater Header Isolation Valve , Field Weld-25 (Component ALV0202

) Magnetic Particle Auxiliary Feedwater System Report Number 5010 005 8 Condensate Storage Tank to Auxiliary Feedwater Header Isolation Valve

, Field Weld-2 6 (Component ALV0202

) Magnetic Particle Auxiliary Feedwater System Report Number 5010 005 9 Condensate Storage Tank to Auxiliary Feedwater Header Isolation Valve

, Field Weld-2 7 (Component ALV0202

) Magnetic Particle Auxiliary Feedwater System Report Number 5010 00 60 Condensate Storage Tank to Auxiliary Feedwater Header Isolation Valve

, Field Weld-28 (Component ALV0202

) Magnetic Particle Auxiliary Feedwater System Report Number 5010 00 61 Condensate S torage Tank to Auxiliary Feedwater Header Isolation Valve

, Field Weld-2 9 (Component ALV0202

) Ma gnetic Particle SYSTEM WELD IDENTIFICATION EXAMINATION TYPE Safety Injection System Report Number 5000 001 0 Safety Injection Accumulator D Outlet, Upstream Check Valve Test Line Isolation Valve, Field Weld

-01 (Component EPHV8877D)

Penetrant Safety Injection System Report Number 5000 0011 Safety Injection Accumulator D Outlet, Upstream Check Valve Test Line Isolation Valve, Field Weld

-02 (Component EPHV8877D)

Penetrant Safety Injection System Report Number 5000 001 2 Safety Injection Accumulator D Outlet, Upstream Check Valve Test Line Isolation Valve, Field Weld

-03 (Component EPHV8877D)

Penetrant Reactor Coolant System Record Number 5030 012 Fabricated Pipe Spool Piece Including Valve BBV0007 Reactor Coolant System Loop 1 Hot Leg to Nuclear Sample System Isolation Valve, Job Number 16001742-405 (Weld Joint s 16001742-405-FW-05 and 06) Radiograph Reactor Coolant System Record Number 5030 014 Reactor Coolant System Pressurizer Chemical and Volume Control System Auxiliary Spray Supply Drain (Component BBV0400)

Radiograph Reactor Coolant System Record Number UT 024 Reactor Pressure Vessel Stud Number 1 (Component 2

-CH-STUD-01) Ultrasonic Reactor Coolant System Record Number UT 025 Reactor Pressure Vessel Stud Number 2 (Component 2

-CH-STUD-02-R1) Ultrasonic Reactor Coolan t System Record Number UT 026 Reactor Pressure Vessel Stud Number 3 (Component 2

-CH-STUD-03) Ultrasonic SYSTEM WELD IDENTIFICATION EXAMINATION TYPE Reactor Coolant System Record Number UT 050 Reactor Pressurizer Safety Nozzle A Inner Radius Area Examination (Component 2-BB03-10A-A-IR, Exam Angle 55° + 38°) Ultrasonic Reactor Coolant System Record Number UT 050 Reactor Pressurizer Safety Nozzle A Inner Radius Area Examination (Component 2-BB 03-10 A-A-IR, Exam Angle 55° - 38°) Ultrasonic Reactor Coolant System Record Number UT-16-052 Reactor Pressurizer Safety Nozzle B Inner Radius Area Examination (Component 2-BB03-10B-B-IR, Exam Angle 55° + 38°) Ultrasonic Reactor Coolant System Record Number UT 052 Reactor Pressurizer Safety Nozzle B Inner Radius Area Examination (Component 2-BB 03-10 B-B-IR, Exam Angle 55

° - 38°) Ultrasonic Reactor Coolant System Record Number UT 053 Reactor Pressurizer Safety Nozzle B

to Top Head Weld (Component 2

-TBB03-10B-B-W, Exam Angle 55° - 38°) Ultrasonic Reactor Coolant System Acquisition Log No. DM/Pipe 22-1 Reactor Outlet Nozzle (Hot Leg) 22

° (Nozzle to Safe

-End Dissimilar Metal Weld 2-RV-301-121-A and Safe

-End to Pipe Weld 2-BB-01-F103) Ultrasonic Reactor Coolant System Acquisition Log No. DM/Pipe 158-1 Reactor Outlet Nozzle (Hot Leg) 158

° (Nozzle to Safe

-End Dissimilar Metal Weld 2-RV-301-121-B and Safe

-End to Pipe Weld 2-BB-01-F203) Ultrasonic SYSTEM WELD IDENTIFICATION EXAMINATION TYPE Reactor Coolant System Acquisition Log No. DM/Pipe 202

-1 Reactor Outlet Nozzle (Hot Leg) 202

° (Nozzle to Safe

-End Dissimilar Metal Weld 2-RV-301-121-C and Safe

-End to Pipe Weld 2-BB-01-F303) Ultrasonic Reactor Coolant System Acquisition Log No. DM/Pipe 338-1 Reactor Outlet Nozzle (Hot Leg) 338

° (Nozzle to Safe

-End Dissimilar Metal Weld 2-RV-301-121-D and Safe

-End to Pipe Weld 2-BB-01-F403) Ultrasonic Safety Injection System Report Number 5041-16-0020 Safety Injection Pumps

- Crosstie to Cold Leg Loops Numbers 1, 2, 3, and 4 (Component Location P049)

Visual Reactor Coolant System Report Number 5041-16-0021 Reactor Pressure Vessel Head (Component RBB01)

Visual Essential Service Water System Record Number 5042 003 5 Essential Service Water System Support

(Component EF0 2 C003142) Visual Essential Service Water System Record Number 5042-16-003 6 Essential Service Water System Support Hanger (Component EF0 3 C0 34134) Visual Essential Service Water System Record Number 5042-16-0037 Essential Service Water System Support

(Component EF01C012311

) Visual Emergency Diesel Generator Record Number 5042 003 8 Diesel Generator A Jacket Water Heat Exchanger Supports (Component EKJ06A) Visual Emergency Diesel Generator Record Number 5042 003 9 Diesel Generator A Jacket Water Heat Exchanger Supports (Component EJH06A) Visual SYSTEM WELD IDENTIFICATION EXAMINATION TYPE Chemical and Volume Control System Report Number 5042-16-0056 Chemical and Volume Control System Pipe Support (Component BG23H004231)

Visual The inspectors reviewed records for the following nondestructive examinations:

SYSTEM IDENTIFICATION EXAMINATION TYPE Condensate System Report Number 5010-16-0040 High Pressure Condensate Main Steam Dump Valve Low Point Drain Steam Trap Bypass Valve (Component ABV0184)

Magnetic Particle Auxiliary Feedwater System Report Number 5010-16-0042 Condensate Storage Tank to Auxiliary Feedwater Pump Suction Check Valve (Component ALV0217)

Magnetic Particle Auxiliary Feedwater System Report Number 5010-16-0048 Auxiliary Feedwater System 3-inch Tee to 3-inch Spool Piece (Job Number 15001243, Field Weld FW-16) Magnetic Particl e Auxiliary Feedwater System Report Number 5010-1-0049 Hardened Condensate Storage Tank

to Auxiliary Feedwater Pump Header Isolation Valve (Component ALV0202, Job Number 15000069, Field Weld FW-30) Magnetic Particle Safety Injection System Report Number 5000-16-0008 Safety Injection Pump B Loop 4 Hot Leg Test Line Isolation HV (Component EMHV8889D)

Penetrant Safety Injection System Report Number 5000 0010 Safety Injection Accumulator D Outlet Upstream Check Valve Test Line Isolation (Component EPHV8877D, Downstream Side of Valve)

Penetrant SYSTEM IDENTIFICATION EXAMINATION TYPE Safety Injection System Report Number 5000-16-0011 Safety Injection Accumulator Outlet Upstream Check Valve Test Line Isolation (Component EPHV8877D, Upstream Side of Valve)

Penetrant Chemical and Volume Control System Report Number 5000 0018 Chemical and Volume Control System Letdown Throttle Valve B (Component BGV0002) Penetrant Reactor Coolant System Record Number 5030 010 Fabricated Pipe Spool Piece Including Valve BBV0007

-Reactor Coolant System Loop 1 Hot Leg to Nuclear Sample System Isolation Valve (Job Number 16001742-400, Field Weld Joint 16001742-400-FW-01) Radiograph Reactor Coolant System Record Number 5030 011 Fabrica ted Pipe Spool Piece Including Valve BBV0007

-Reactor Coolant System Loop 1 Hot Leg to Nuclear Sample System Isolation Valve (Job Number 16001742

-400 , Field Weld Joint 16001742 FW-02) Radiograph Reactor Coolant System Report Number 5042 028 Reactor Pressure Vessel Head (Component RBB01, Second Inspection

) Visual During the review and observation of each examination, the inspectors observed whether activities were performed in accordance with the ASME Code requirements and applicable procedures. The inspectors also reviewed the qualifications of all nondestructive examination technicians performing the inspections to determine whether they were current.

The inspectors directly observed a portion of the following welding activities:

SYSTE M WELD IDENTIFICATION WELD TYPE Reactor Coolant System Valve BBV-0400, Reactor Coolant System Pressurizer Chemical and Volume Control System Auxiliary Spray Supply Drain (Job 15001126-500, ASME Code Class 2, Field Weld FW-03) Manual Gas Tungsten Arc Welding Chemical and Volume Control System Valve BGV-0003, CVCS Letdown Orifice A Outlet Throttle Valve Piping (Job 13005673-510, ASME Code Class 2, Field Weld FW

-03 , -04 and -05) Manual Gas Tungsten Arc Welding Chemical and Volume Control System Valve BGV-0002, CVCS Letdown Orifice A Outlet Throttle Valve Piping (Job 13005672

-510, ASME Code Class 2, Field Weld FW

-01, -02, and -03) Manual Gas Tungsten Arc Welding Auxiliary Feedwater System Hardened Condensate Storage Tank Re-Circulation Line An d Tie-In to Existing Auxiliary Feedwater System Piping (Job 15001243

-500, Field Welds FW-11, -12, -13, -14, -15, and -16) Manual Gas Tungsten Arc Welding The inspectors reviewed records of the following welding activities

SYSTEM WELD IDENTIFICATION WELD TYPE Chemical and Volume Control System Valve BGV-0001, CVCS Letdown Orifice A Outlet Throttle Valve Piping

(Job 13005670

-510, ASME Code Class 2, Field Weld FW

-03, -04, and -05) Manual Gas Tungsten Arc Welding Chemical and Volume Control System Valve BGV-0001, CVCS Letdown Orifice A Outlet Throttle Valve Piping (Job 13005670

-010, ASME Code Class 2, Field Weld FW

-01, and -02) Manual Gas Tungsten Arc Welding Chemical and Volume Control System Valve BGV-0002, CVCS Letdown Orifice A Outlet Throttle Valve Piping (Job 13005672

-010, ASME Code Class 2, Field Weld FW

-04, and -05) Manual Gas Tungsten Arc Welding The inspectors reviewed whether the welding procedure specifications and the welders had been properly qualified in accordance with ASME Code

,Section IX requirements. The inspectors also determined whether essential variables were identified, recorded in the procedure qualification record, and formed the bases for qualification of the welding procedure specifications

.

b. Findings

No findings were identified.

.2 Vessel Upper Head Penetration Inspection Activities

a. Inspection Scope

The inspectors reviewed the results of the licensee's bare metal visual inspection of the reactor vessel upper head penetrations to determine whether the licensee identified any evidence of boric acid challenging the structural integrity of the reactor head components and attachments.

The inspectors also verified that the required inspection coverage was achieved and limitations were properly recorded. The inspectors reviewed whether the personnel performing the inspection were certified examiners to their respective nondestructive examination method.

b. Findings

The licensee replaced the reactor head during the last refueling outage, R F-20, during the fall 2014, and elected to do a visual inspection of the reactor head at the completion of the first inservice cycle. Some items of interest were identified requiring further inspection. The licensee concluded that there was no leakage associated with any of the reactor vessel closure head penetrations which was documented in Callaway Action Request 201603166. The inspectors witnessed the inspection, discussed the concern with the individuals that had performed the inspection, reviewed the photographs of the areas of concern, and agreed with the licensee's conclusion.

No findings were identified.

.3 Boric Acid Corrosion Control Inspection Activities

a. Inspection Scope

The inspectors reviewed the licensee's implementation of its boric acid corrosion control program for monitoring degradation of those systems that could be adversely affected by boric acid corrosion. The inspectors reviewed the documentation associated with the licensee's boric acid corrosion control walkdown as specified in Procedure EDP-ZZ-01004 , "Boric Acid Corrosion Control Program," Revision 18. The inspectors reviewed whether the visual inspections emphasized locations where boric acid leaks could cause degradation of safety significant components and whether engineering evaluation used corrosion rates applicable to the affected components and properly assessed the effects of corrosion induced wastage on structural or pressure boundary integrity.

The inspectors observed whether corrective actions taken were consistent with the ASME C ode and 10 CFR Part 50, Appendix B requirements.

The inspectors reviewed licensee boric acid evaluations where boric acid deposits were found on reactor coolant system piping components and other components:

COMPONENT NUMBER DESCRIPTION CALLAWAY ACTION REQUEST BBHV8002A and BHV8002B Reactor H ead Vent Valve Tailpieces on Top of the Reactor H ead 201406993 EEJ01A Residual Heat Removal (RHR) System H eat Exchanger A

- Flange 201406827 EEJ01 B Residual Heat Removal (RHR) System H eat Exchanger B - Flange 201406 528 BB10-C503 Hangar BB10

-C503 (Adjacent Valve BBHV8141C, RCP C SEAL # 1 SEAL WTR OUT ISO HV Experienced Packing Leakage) 201407170 EMHV8923A Refueling Water Storage Tank to Safety Injection Pump A Suction Isolation Valve 201407454 EPV0124 Downstream Isolation Valve for Test Header Valve EPHV8879D 201407589 EMV0179 ENV0123 Safety Injection Pump A from Residual Heat Removal Heat Exchanger A Suction Vent Valve B Containment Spray Pump Casing and Seal Housing Vent Valve 201408130 EJ8842 Residual Heat Removal Trains A&B Safety Injection System Hot Leg Recirculation Supply Header Pressure Relief Valve 201409218 BBHV8351A Reactor Coolant Pump A Seal Water Supply Isolation Valve 201500874 BGFCV0110A BGPIS0141 Blending Tee Flow Control Valve and Seal Water Injection Filter B 201503867 BGV0551 Chemical and Volume Control System Seal Water Injection Filter B Outlet Drain Valve (Bolted Blind Flange Assembly Downstream of Valve) 201504450 EPHV8877B Safety Injection System Upstream Check Test Line Isolation Valve 201505362 EMHV8923A Refueling Water Storage Tank to Safety Injection Pump A Suction Isolation Valve 201600224

b. Findings

No findings were identified.

.4 Steam Generator Tube Inspection Activities

a. Inspection Scope

The inspectors reviewed the steam generator tube eddy current examination scope and expansion criteria to determine whether these criteria met technical specification requirements, EPRI guidelines, and commitments made to the NRC. The inspectors also reviewed whether the eddy current examination inspection scope included areas of degradations that were known to represent potential eddy current test challenges such as the top of tubesheet, tube support plates, and U

-bends. The inspectors confirmed that repairs were required at the time of the inspection

. Steam Generator Inspection The inspectors verified that the number and sizes of steam generator tube flaws/degradation identified were consistent with the licensee's previous outage operational assessment predictions.

The inspectors verified that steam generator eddy current examination scope and expansion criteria met technical specification requirements.

The inspectors verified that eddy current probes and equipment configurations used to acquire data from the steam generator tubes were qualified to detect the known/expected types of steam generator tube degradation in accordance with Appendix H, "Performance Demonstration for Eddy Current Examination of EPRI Document 1013706." Eddy current bobbin probe examinations all four steam generators (100 percent of all inservice tubes, full length tube

-end to tube

-end) was performed.

Eddy current array probe examinations (all four steam generator s) was performed.

The inspectors reviewed the licensee's identification of the following tube degradation mechanisms

All inservice 1R18 tube support plate multi-land wear indications, including the following:

o Steam Generator C (8 lands) o Steam Generator D (4 lands) Anti-vibration bar (AVB) wear All cold leg tubes having non

-nominal tubesheet drill hole diameters 20 percent of hot leg tubes with sludge from the 1R18 sludge analysis Tube Repair The inspectors verified that the licensee implemented repair methods which were consistent with the repair processes allowed in the plant technical specification requirements and to determine if qualified depth sizing methods were applied to degraded tubes accepted for continued service. The licensee repaired a total of 25 tubes. The following repairs were made.

Steam Generator A

- 9 tubes plugged Steam Generator C

- 14 tubes plugged Steam Generator D

- 2 tubes plugged Secondary Side Inspections The inspectors observed and reviewed secondary side inspection results and verified the licensee took corrective actions in response to the observed degradation.

Inspections performed were:

Top of tubesheet water lancing on all four steam generator s: o Prior to water lancing, a pre

-look visual inspection was performed to examine the sludge piles in two steam generator s.

Foreign object search and retrieval (FOSAR)

Visual inspections of steam drums in steam generator A and steam generator D Visual Examinations The inspectors observed and reviewed the visual examination inspection results. Inspections performed were:

As-found and as

-left visual examination of primary channel heads (both hot leg and cold leg)

Nuclear Safety Advisory Letter 12-1 (and Information Notice 2013-20) primary bowl inspections

b. Findings

No findings were identified.

.5 Identification and Resolution of Problems

a. Inspection scope

The inspectors reviewed 22 Callaway action request reports which dealt with inservice inspection activities and found the corrective actions for inservice inspection issues were appropriate. From this review the inspectors concluded that the licensee has an appropriate threshold for entering inservice inspection issues into the corrective action program and has procedures that direct a root cause evaluation when necessary. The licensee also has an effective program for applying industry inservice inspection operating experience.

b. Findings

No findings were identified.

.6 Essential Service Water System Inspection

a. Inspection Scope

Inspectors performed a focused baseline inspection of the essential service water system due to concerns with system reliability as a result of ongoing corrosion and water hammer issues. The scope of the inspection included system walkdowns as well as review of design calculations, Callaway actio n requests, operability determinations, and testing and surveillances associated with the essential service water system.

b. Findings

A finding of very low safety significance was identified and is discussed in Section

1R07 , Heat Sink Performance

.

1R11 Licensed Operator Requalification Program and Licensed Operator Performance

(71111.11)

.1 Review of Licensed Operator Requalification

a. Inspection Scope

On May 31, 2016, the inspectors observed an evaluated simulator scenario performed by an operating crew. The inspectors assessed the performance of the operators and the evaluators' critique of their performance. The inspectors also assessed the modeling and performance of the simulator during the activities.

These activities constituted completion of one quarterly licensed operator requalification program sample, as defined in Inspection Procedure 71111.11.

b. Findings

No findings were identified.

.2 Review of Licensed Operator Performance

a. Inspection Scope

On April 2, 2016, the inspectors observed the performance of on

-shift licensed operators in the plant's main control room. At the time of the observations, the plant was in a period of heightened activity due to shutdown activities for Refueling Outage 21, including the main turbine overspeed trip testing.

In addition, the inspectors assessed the operators' adherence to plant procedures, including Procedure ODP-ZZ-00001, "Operations Department

- Code of Conduct,"

Revision 97, and other operations department policies.

These activities constitute d completion of one quarterly licensed operator performance sample, as defined in Inspection Procedure 71111.11.

b. Findings

No findings were identified.

1 R 12 Maintenance Effectiveness (71111.12)

a. Inspection Scope

On March 24, 2016, the inspectors reviewed the emergency core cooling system room coolers for instances of degraded performance or condition of safety

-related structures , systems, and components.

The inspectors reviewed the extent of condition of possible common cause structure, system, and component failures and evaluated the adequacy of the licensee's corrective actions. The inspectors reviewed the licensee's work practices to evaluate whether

these may have played a role in the degradation of the structures, systems, and components. The inspectors assessed the licensee's characterization of the degradation in accordance with 10 CFR 50.65 (the Maintenance Rule), and verified that the licensee was appropriately tracking degraded performance and conditions in accordance with the Maintenance Rule.

These activities constituted completion of one maintenance effectiveness sample

, as defined in Inspection Procedure 71111.12.

b. Findings

A finding of very low safety significance was identified and is discussed in Section

1R07 , Heat Sink Performance

.

1 R 13 Maintenance Risk Assessments and Emergent Work Control (71111.13)

a. Inspection Scope

The inspectors reviewed three risk assessments performed by the licensee prior to changes in plant configuration and the risk management actions taken by the licensee in response to elevated risk:

April 4, 2016, yellow risk for reduced reactor coolant system inventory to support reactor vessel head assembly removal for refuel April 19, 2016 , yellow risk for train B spent fuel cooling system out

-of-service and train B electrical switchgear work in progress May 6, 2016, risk evaluation in accordance with Technical Specification 3.0.4.b for the atmospheric steam dumps , feedwater regulating valves, and turbine-driven auxiliary feedwater pump inoperable for moving from Mode 4 to Mode 3 The inspectors verified that these risk assessment were performed timely and in accordance with the requirements of 10 CFR 50.65 (the Maintenance Rule) and plant procedures. The inspectors reviewed the accuracy and completeness of the licensee's risk assessments and verified that the licensee implemented appropriate risk management actions based on the result of the assessments

. The inspectors also observed portions of two emergent work activities that had the potential to affect the functional capability of mitigating systems:

April 12, 2016, train A emergency diesel generator pump seals installed backwards June 21, 2016, loose bolts on train B control room air conditioning system The inspectors verified that the licensee appropriately developed and followed a work plan for these activities. The inspectors verified that the licensee took precautions to minimize the impact of the work activities on unaffected structures, systems, and components

. These activities constitute d completion of five maintenance risk assessments and emergent work control inspection sampl es, as defined in Inspection Procedur e 71111.13.

b. Findings

No findings were identified.

1 R 15 Operability Determinations and Functionality Assessments (71111.15)

a. Inspection Scope

The inspectors reviewed six operability determinations and functionality assessments that the licensee perf ormed for degraded or nonconforming structure s, systems, or components

April 11, 2016, operability determination of safety related instrument bus inverters April 14, 2016 , operability determination of leaks identified during tra in B engineering safety feature actuation system testing April 17, 2016, operability determination of containment electrical penetrations May 24, 2016, functionality assessment of the emergency off

-site facility with no air conditioning and no off

-site power May 31, 2016, power-operated relief valve block valve closed June 28, 2016, operability determination for train A emergency diesel generator due to jacket water heater not cycling off The inspectors reviewed the timeliness and technical adequacy of the licensee's evaluations. Where the licensee determined the degraded structures, systems, or com ponents to be operable or functional, the inspectors verified that the licensee's compensatory measures were appropriate to provide reasonable assurance of operability or functionality. The inspectors verified that the licensee had considered the effect of other degraded conditions on the operability or functionality of the degraded structure, system, or com p onent. These activities constitute d completion of six operability and functionality review samples , as defined in Inspection Procedure 71111.15.

b. Findings

Introduction.

The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," associated with the licensee's failure to perform adequate operability assessments when a degraded or nonconforming condition was identified.

Specifically, after the licensee identified that a severe water hammer transient would occur following a loss of off

-site power, the licensee generated an operability evaluation that relied on judgement and inaccurate information which failed to establish a reasonable expectation of operability.

Description.

On April 4, 2016 , the licensee identified that during a loss of off

-site power event the essential service water system will experience a column separation that results in a severe water hammer transient that could subject portions of the system to transient pressures and dynamic forces in excess of current station analyses. In response to this

, the licensee initiated Callaway Action Request 201603472 to capture the issue in the station's corrective action program. The licensee subsequently documented a prompt operability determination for the essential service water system.

Inspectors subsequently reviewed the licensee's prompt operability determination. During their review, the inspectors noted that the licensee had based their operability determination on the results of a special test conducted on April 27, 2016, to simulate system response to a loss of off

-site power event. Specifically, the licensee had collected data during the test associated with the strength of the system pressure wave, which was used to estimate pipe and support loads, and performed system walkdowns following the test and did not note any system damage.

Inspectors noted the following concerns with the licensee's determination:

The special test was run with the essential service water system at 68 degrees - the temperature had not been corrected to 95 degrees (design basis temperature of the ultimate heat sink). This resulted in a non

-conservative result since water hammer transients are more severe at elevated temperatures.

Due to the location of monitoring equipment, the measured strength of the system pressure wave was not representative of the peak pressure seen in the system. Therefore, the use of the measured peak pressure was non-conservative.

The testing lineup did not have all system components in their accident lineup which resulted in a non

-conservative damping of the severity of the water hammer transient.

Based on this, the inspectors determined that although the licensee's evaluation provided a reasonable expectation of operability under the current plant conditions, it failed to establish a reasonable expectation of operability for the identified condition at worst case design conditions for the system. Inspectors informed the licensee of their concerns and the licensee initiated Callaway Action Request 201605488.

The licensee performed a new operability evaluation

, and based on engineering judgement

, determined that the leaks that had previously been identified would not prevent the system from providing sufficient cooling to safety

-related components or challenge the required essential service water system inventory.

Analysis.

The licensee's failure to properly assess and document the basis for operability when a severe water hammer occurred in the essential service water system was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, severe water hammer transients in the essential service water system due to a loss of off

-site power result in a conditio n where structures, systems, and components necessary to mitigate the effects of accidents may not have functioned as required.

Using Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At

-Power," dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding: did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic event, and (1)was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2)did not represent a loss of system and/or function, (3)did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out

-of-service for longer than their technical specification allowed outage time, and (4)does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with the licensee' s maintenance rule program. This finding has a cross

-cutting aspect of conservative bias in the human performance area because the licensee failed to demonstrate that a proposed action was safe in order to proceed, rather than unsafe in order to stop.

Specifically, the licensee's use of unsupported judgement and incorrect data resulted in an evaluation that failed to demonstrate a reasonable expectation of operability

[H.14].

Enforcement.

Title 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," requires, in part, that activities affecting quality shall be accomplished in accordance with instructions, procedures, or drawings of a type appropriate to the circumstances. Callaway Procedure ODP-ZZ-00001, Addendum 15, "Operability and Functionality Determinations," an Appendix B quality related procedure, provides instructions for performing operability determinations. Procedure ODP-ZZ-00001, Addendum 15, step 3.2.2 states, in part, "The SM should ENSURE an appropriate level of questioning and challenging of assumptions occurs to ensure that a sound basis for operability exists throughout the OD process." Contrary to the above, on April 14, 2016, the licensee failed to ensure an appropriate level of questioning and challenging of assumptions occurred to ensure that a sound basis for operability existed throughout the operability determination process. Specifically, after the licensee identified that a severe water hammer transient would occur following a loss of off

-site power, the licensee generated an operability evaluation that relied on judgement and inaccurate information which failed to establish a reasonable expectation of operability. The licensee implemented immediate correction actions to enter this issue into the corrective action program for resolution. The licensee also performed an operability determination which established a reasonable expectation of operability pending implementation of corrective actions. This violation is being treated as a non

-cited violation, consistent with Section 2.3.2.a of the Enforcement Policy because it was of very low safety significance, and was entered into the licensee's corrective action program as Callaway Action Request s 201605488: NCV 05000483/2016002

-03, "Failure to Adequately Evaluate Operability for a Degraded Condition."

1 R 18 Plant Modifications (71111.18)

Permanent Modifications

a. Inspection Scope

The inspectors reviewed three permanent plant modification s that affected risk significant structures, systems, and components

May 19, 2016, modification that tied in the newly built hardened condensate storage tank to the auxiliary feedwater system (Modification Package 13-0033) June 10, 2016, modification that installed new check valves in the service water supply lines to the essential service water system (Modification Package 10-0003) June 10, 2016, modification that revised sequencer operation of EFHV0037 and EFHV0038 (Modification Package 10-0004)

The inspectors reviewed the design and implementation of the modification s. The inspectors verified that work activities involved in implementing the modification s did not adversely impact operator actions that may be required in response to an emergency or other unplanned event. The inspectors verified that post

-modification testing was adequate to establish the operability and functionality of the structures, systems, or components as modified.

These activities constitute d completion of three sample s of permanent modifications, as defined in Inspection Procedure 71111.18.

b. Findings

No findings were identified.

1 R 19 Post-Maintenance Testing (71111.19)

a. Inspection Scope

The inspectors reviewed five post-maintenance testing activities that affected risk-significant structures, systems, or components

March 24, 2016, train A residual heat removal room cooler leak April 13, 2016, train A emergency diesel generator maintenance window April 14, 2016, containment recirculation sump to train A residual heat removal pump suction isolation valve June 8, 2016, spring cans supporting the essential service water piping to t he component cooling water heat exchanger June 20, 2016, letdown heat exchanger outlet pressure control valve repairs The inspectors reviewed licensing

- and design-basis documents for the structures, systems, and components and the maintenance and post-maintenance test procedures. The inspectors observed the performance of the post

-maintenance tests to verify that the licensee performed the tests in accordance with approved procedures, satisfied the established acceptance criteria, and restored the operability of the affected structures, systems, and component s. These activities constitute d completion of five post-maintenance testing inspection samples, as defined in Inspection Procedure 71111.19.

b. Findings

No findings were identified.

1 R 20 Refueling and Other Outage Activities (71111.20)

a. Inspection Scope

During the station's refueling outage that concluded on May 10, 2016, the inspectors evaluated the licensee's outage activities. The inspectors verified that the licensee considered risk in developing and implementing the outage plan, appropriately managed personnel fatigue, and developed mitigation strategies for losses of key safety functions. This verification included the following:

Review of the licensee's outage plan prior to the outage Review and verification of the licensee's fatigue management activities Monitoring of shut

-down and cool

-down activities Verification that the licensee maintained defense

-in-depth during outage activities Observation and review of reduced

-inventory activities Observation and review of fuel handling activities Monitoring of heat

-up and startup activities These activities constitute d completion of one refueling outage sample, as defined in Inspection Procedure 71111.20.

b. Findings

No findings were identified.

1 R 22 Surveillance Testing (71111.22)

a. Inspection Scope

The inspectors observed three risk-significant surveillance tests and reviewed test results to verify that these tests adequately demonstrated that the structures, systems, and component s were capable of performing their safety functions:

Inservice tests:

April 6, 2016, emergency core cooling system full flow test

Other surveillance tests:

April 14, 2016, train B engineering safety feature actuation system testing June 29, 2016, train B emergency diesel generator slow start and 1

-hour run The inspectors verified that these tests met technical specification requirements, that the licensee performed the tests in accordance with their procedures, and that the results of the test satisfied appropriate acceptance criteria. The inspectors verified that the licensee restored the operability of the affected structures, systems, and component s following testing.

These activities constitute d completion of three surveillance testing inspection sample s , as defined in Inspection Procedure 71111.22.

b. Findings

No findings were identified.

RADIATION SAFETY

Cornerstones: Public Radiation Safety and Occupational Radiation Safety 2 RS 1 Radiological Hazard Assessment and Exposure Controls (71124.01)

a. Inspection Scope

The inspectors evaluated the licensee's performance in assessing the radiological hazards in the workplace associated with licensed activities. The inspectors assessed the licensee's implementation of appropriate radiation monitoring and exposure control measures for both individual and collective exposures. The inspectors walked down various portions of the plant and performed independent radiation dose rate measurements. The inspectors interviewed the radiation protection manager, radiation protection supervisors, and radiation workers. The inspectors reviewed licensee performance in the following areas:

Radiological hazard assessment, including a review of the plant's isotopic mix and isotopic percent abundance, hard

-to-detect radionuclides and potential alpha hazards. The inspectors also reviewed the licensee's evaluations of changes in plant operations and radiological surveys to identify and detect dose rates, neutron hazards, hot particle exposures, severe dose gradients, airborne radioactivity monitoring, and surface contamination levels.

Instructions to workers, including labeling or marking containers of radioactive material, radiation work permits, actions for electronic dosimeter alarms, and changes to radiological conditions.

Contamination and radioactive material control including release of potentially contaminated material from the radiologically controlled area, radiological survey performance, radiation instrument sensitivities, material control and release criteria, procedural guidance, and control and accountability of sealed radioactive sources.

Radiological hazards control and work coverage including field observations of job performance and adequacy of radiological controls. During walk downs of the facility and job performance observations, the inspectors evaluated ambient radiological conditions, radiological postings, adequacy of radiological controls, radiation protection job coverage, and contamination controls. The inspectors also evaluated the use of electronic dosimeters in high noise areas, dosimetry selection and placement, implementation of effective dose equivalent for external exposures (EDEX), and the application of dosimetry to effectively monitor exposure for work in areas with significant dose rate gradients. The inspectors examined the licensee's controls for highly activated or contaminated materials (non-fuel) stored within spent fuel and other storage pools and evaluated airborne radioactive controls and monitoring.

High radiation area and very high radiation area controls including posting and physical controls for high radiation areas and very high radiation areas. During plant walk downs, the inspectors verified the adequacy of posting and physical controls, including for areas of the plan with the potential to become risk-significant high radiation areas.

Radiation worker performance and radiation protection technician proficiency with respect to radiation protection work requirements. The inspectors determined if workers were aware of the significant radiological conditions in their workplace, radiation work permit controls/limits in place, and were aware of their electronic alarming dosimeter dose and dose rate set points. The inspectors observed radiation protection technician job performance, including the performance of radiation surveys.

Problem identification and resolution for radiological hazard assessment and exposure controls. The inspectors reviewed audits, self

-assessments, and corrective action program documents to verify problems were being identified and properly addressed for resolution.

These activities constituted completion of the seven required samples of radiological hazard assessment and exposure control program, as defined in Inspection Procedure 71124.01.

b. Findings

No findings were identified.

2 RS 3 In-plant Airborne Radioactivity Control and Mitigation (71124.03)

a. Inspection Scope

The inspectors evaluated whether the licensee controlled in

-plant airborne radioactivity concentrations consistent with as low as reasonably achievable (ALARA) principles and that the use of respiratory protection devices did not pose an undue risk to the wearer. During the inspection, the inspectors interviewed licensee personnel, walked down various areas in the plant, and reviewed licensee performance in the following areas:

Engineering controls, including the use of permanent and temporary ventilation systems to control airborne radioactivity. The inspectors evaluated installed ventilation systems, including review of procedural guidance, verification the systems were used during high

-risk activities, and verification of airflow capacity, flow path, and filter/charcoal unit efficiencies. The inspectors also reviewed the use of temporary ventilation systems used to support work in contaminated areas such as high

-efficiency particulate air/charcoal negative pressure units. Additionally, the inspectors evaluated the licensee's airborne monitoring protocols, including verification that alarms and set points were appropriate.

Use of respiratory protection devices and evaluation of the licensee's respiratory protection program including use, storage, maintenance, and quality assurance of National Institute for Occupational Safety and Health

-certified equipment, air quality and quantity for supplied

-air devices and self-contained breathing apparatus (SCBA) bottles, qualification and training of personnel, and user performance.

Self-contained breathing apparatus for emergency use including the licensee's capability for refilling and transporting SCBA air bottles to and from the control room and operations support center during emergency conditions, hydrostatic testing of SCBA bottles, status of SCBA staged and ready for use in the plant including vision correction, mask sizes, etc., SCBA surveillance and maintenance records, and personnel qualification, training, and readiness.

Problem identification and resolution for airborne radioactivity control and mitigation. The inspectors reviewed audits, self

-assessments, and corrective action documents to verify problems were being identified and properly addressed for resolution.

These activities constituted completion of the four required samples of in

-plant airborne radioactivity control and mitigation program, as defined in Inspection Procedure 71124.03.

b. Findings

No findings were identified

OTHER ACTIVITIES

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Security 4OA 1 Performance Indicator Verification (71151)

.1 Safety System Functional Failures (MS05)

and Mitigating Systems Performance Index:

Heat Removal Systems (MS08)

a. Inspection Scope

For the period of second quarter 2015 through first quarter 2016 , the inspectors reviewed licensee event reports, maintenance rule evaluations, and other records that could indicate whether safety system functional failures had occurred. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7, and NUREG-1022, "Event Reporting Guidelines:

10 CFR 50.72 and 50.73," Revision 3, to determine the accuracy of the data reported.

These activities constituted verification of the safety system functional failures performance indicator and the mitigating system performance ind ex performance indicator, as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.2 Reactor Coolant System Identified Leakage (BI02)

a. Inspection Scope

The inspectors reviewed the licensee's records of reactor coolant system identified leakage for the period of second quarter 2015 through first quarter 2016 to verify the accuracy and completeness of the reported data. The inspectors reviewed the performance of Procedure OSP-BB-00009, "RCS Inventory Balance," Revision 37, conducted on May 12, 2016. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7, to determine the accuracy of the reported data. These activities constituted verification of the reactor coolant system leakage performance indicator, as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.3 Occupational Exposure Control Effectiveness (OR01)

a. Inspection Scope

The inspectors verified that there were no unplanned exposures or losses of radiological control over locked high radiation areas and very high radiation areas during the period of October 1, 2015 , through March 31, 2016. The inspectors reviewed a sample of radiologically controlled area exit transactions showing exposures greater than

100 mrem. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7, to determine the accuracy of the reported data.

These activities constituted verification of the occupational exposure control effectiveness performance indicator as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

.3 Radiological Effluent Technical Specifications/Off

-site Dose Calculation Manual Radiological Effluent Occurrences (PR01)

a. Inspection Scope

The inspectors reviewed corrective action program records for liquid or gaseous effluent releases that occurred between October 1, 2015 , and March 31, 2016 , and were reported to the NRC to verify the performance indicator data. The inspectors used definitions and guidance contained in Nuclear Energy Institute Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revisio n 7, to determine the accuracy of the reported data.

These activities constituted verification of the radiological effluent technical specifications/off

-site dose calculation manual radiological effluent occurrences performance indicator as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

4OA 2 Problem Identification and Resolution (7115 2)

.1 Routine Review

a. Inspection Scope

Throughout the inspection period, the inspectors performed daily reviews of items entered into the licensee's corrective action program and periodically attended the licensee's condition report screening meetings. The inspectors verified that licensee personnel were identifying problems at an appropriate threshold and entering these problems into the corrective action program for resolution. The inspectors verified that the licensee developed and implemented corrective actions commensurate with the significance of the problems identified. The inspectors also reviewed the licensee's problem identification and resolution activities during the performance of the other inspection activities documented in this report.

b. Findings

No findings were identified.

.2 Semiannual Trend Review

a. Inspection Scope

To verify that the licensee was taking corrective actions to address identified adverse trends that might indicate the existence of a more significant safety issue, the inspectors reviewed corrective action program documentation associated with the following licensee-identified trends:

Negative trend on essential service water leaks from safety related room coolers (Callaway Action Request 201 6 0 2658) Negative trend involving leaks on plant equipment as a result of train B engineering safety feature actuation system testing (Callaway Action Request 201 603472) These activities constitute completion of one semiannual trend review sample, as defined in Inspection Procedure 71152.

b. Observations and Assessments The inspectors' review of the possible trends noted above produced the following observations and assessments:

During the period of March 23 to May 3, 2016, the licensee had twelve leaks across eight safety

-related room coolers serviced by essential service water. The licensee considered this a negative trend and performed a root cause evaluation in Callaway Action Request 201602658 to determine the causes for the negative trend. The licensee determined the equipment reliability process did not adequately address the long

-standing equipment issues associated with safety related copper

-nickel heat exchangers.

To address the issue, the licensee replaced several room coolers during the recent refueling outage and has a plan to replace all but the containment coolers during the current online cycle. The containment coolers are planned for replacement during the next refueling outage. The inspectors evaluated the licensee's response to the negative trend and determined the actions were appropriate

.

Since April 2007, the Callaway plant has experienced leaks on plant equipment as a result of engineering safety feature actuation system testing. These leaks did not occur during every test, but several components have had repetitive failures and a leak had occurred on a component every refueling outage since 2013. The licensee considered this a negative trend and performed a root cause evaluation in Callaway Action Request 201603472 to determine the causes for the negative trend.

The licensee determined the original design of the system did not appropriately account for water column separation and collapse during functional operation and the corrective action process did not adequately drive the organization to correct the condition

. To address the issue, the licensee "hardened" several components during the recent refueling outage and has hired an external company to evaluate the pressures expected during a design

-based accident. The licensee will address the results of the analysis when it becomes available. The inspectors evaluated the licensee's response to the negative trend and determined the actions were appropriate.

c. Findings

A finding associated with these trends is documented in Section 4OA2.3.

.3 Annual Follow

-up of Selected Issues

a. Inspection Scope

The inspectors selected one issue for an in-depth follow

-up: On June 10, 2016, the inspectors reviewed Callaway Action Request 201010634 associated with Callaway's response to a non

-cited violation that was issued in Inspection Report 05000483/2010006 (ML103540576)

. The inspectors assessed the licensee's problem identification threshold, cause analyses, extent of condition reviews and compensatory actions. The inspectors identified that the licensee failed to appropriately prioritize the corrective actions and that these actions were not adequate to correct the condition

.

These activities constitute d completion of one annual follow

-up sample as defined in Inspection Procedure 71152.

b. Findings

Introduction.

Inspectors identified a Green cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," associated with the licensee's failure to take timely corrective action for a previously identified condition adverse to quality.

Specifically, the licensee failed to adequately resolve water hammer and corrosion issues that were previously identified by the NRC as non

-cited violation 05000483/2010006

-01 and the failure to resolve these issues resulted in subsequent safety

-related equipment failures.

Description.

Inspectors reviewed licensee's actions taken to address Non

-cited Violation 05000483/2010006

-01, "Failure to Correct Degraded Condition in Essential Service Water System in a Timely Manner," which was documented in Callaway Action Request 201010634. This non

-cited violation was issued because the licensee had been experiencing water hammer events which had caused leaks in safety-related joints and when coupled with system corrosion issues had resulted in leaks in heat exchanger tubes, fittings, and other components.

Inspectors reviewed the licensee's corrective actions taken in response to Non

-cited Violation 05000483/2010006-01. Inspectors noted that the licensee had implemented modifications to the station, Modification Packages 10-0003 and 10-0004, which installed check valves in the service water supply lines to the essential service water system and changed the timing sequence for valve operation in the essential service water system. The purpose of these modifications was to reduce the pressure transient imposed on the essential service water system from water hammer events caused by column separation. Inspectors determined that the licensee had not implemented corrective actions to address the corrosion issues that were also identified in the non

-cited violation and Callaway Action Request 201010634 was closed.

Inspectors performed a subsequent review of the licensee's corrective action program documents and noted that water hammer events continued to occur when the essential service water system was operated during simulated accident conditions (engineering safety feature actuation system testing). Inspectors identified 28 instances where water hammer events and corrosion issues had damaged safety

-related components since Non-cited Violation 05000483/2010006

-01 had been issued. Examples include:

November 17, 2011, train B component cooling water heat exchanger tube side relief valve and the inlet tube side drain valve were found the be leaking by following engineering safety feature actuation system testing December 6, 2011, train A motor driven auxiliary feedwater pump room cooler tube leak April 12, 2012

, train A centrifugal charging pump room cooler tube leak April 29, 2012, train B component cooling water room cooler gasket leak following engineering safety feature actuation system testing May 1, 2013, train B motor driven auxiliary feedwater pump room cooler tube leak following engineering safety feature actuation system testing October 17, 2014, train A centrifugal charging pump room cooler tube leak, B motor driven auxiliary feedwater pump room cooler tube leak, B control room air conditioning condenser endbell gasket leak, and B emergency diesel generator intercooler expansion joint leak following engineering safety feature actuation system testing Additionally, from March 23 to May 3, 2016, the licensee had identified twelve leaks across eight safety-related room coolers serviced by essential service water and damaged gaskets on the safety

-related control room chiller (Licensee Event Report 2016-001-00).

Based on this, inspectors determined that the modifications, Modifications Packages10-000 3 and 10-0004 that were implemented by the licensee were not adequate to mitigate the effects of a water hammer transient. Specifically, system corrosion issues and column separation/water hammer events continued to occur, and these events continued to cause damage to safety related components.

Based on this, inspectors determined that the licensee had failed to take timely and adequate corrective actions to correct the water hammer and corrosion issues in the essential service water system.

Inspectors informed the licensee of their observations and the licensee initiated Callaway Action Request 201604440 to capture this issue in the station's corrective action program. The licensee also generated an operability determination

, and based on engineering judgement

, determined that though water hammer transients had caused leaks in the system, the leaks that had previously been identified would not prevent the system from providing sufficient cooling to safety

-related components or challenge the required essential service water system inventory.

Analysis.

The licensee's failure to take timely and adequate corrective actions to correct a condition adverse to quality was a performance deficiency. The performance deficiency i s more than minor, and therefore a finding, because it i s associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correct water hammer and corrosion issu e resulted in the licensee declaring safety-related room coolers and chillers inoperable until an analysis of system operability was completed. This affected their capability to respond to initiating events to prevent undesirable consequences.

Using Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At

-Power," dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding

(1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2)did not represent a loss of system and/or function, (3)did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out

-of-service for longer than their technical specification allowed outage time, and (4)does not represent an actual loss of function of one or more non

-technical specification trains of equipment designated as high safety

-significant for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with the licensee's maintenance rule program. This finding has a cross-cutting aspect of resources in the human performance area because the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety. Specifically, by failing to address water hammer and corrosion issues, station management failed to ensure that the essential service water system was available and adequately maintained to respond during a loss of off

-site power event

[H.1].

Enforcement.

Title 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," requires, in part, that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. Contrary to the above, from November 2010 through June 2016, for quality related components associated with the essential service water system, to which 10 CFR Part 50, Appendix B applies, the licensee failed to assure that conditions adverse to quality were promptly identified and corrected. Specifically, the licensee failed to adequately resolve water hammer and corrosion issues which were previously identified by the NRC as Non

-cited Violation 05000483/2010006

-01 and the failure to resolve these issues resulted in subsequent safety

-related equipment failures. The licensee implemented immediate correction actions to enter this issue into the corrective action program for resolution. The licensee also performed an operability determination that established a reasonable expectation of operability pending implementation of corrective actions. The violation was entered into the licensee's corrective action program as Callaway Action Request 201604440. This violation is being treated as a cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy, because the licensee did not restore compliance (or demonstrate objective evidence of plans to restore compliance) within a reasonable period of time (i.e.

, in a time frame commensurate with the significance of the violation) after the violation was identified. A Notice of Violation is documented in Enclosure 1

VIO 05000483/2016002-04 , "Failure to Promptly Correct Conditions Adverse to Quality."

4OA 3 Follow-up of Events and Notices of Enforcement Discretion (71153)

(Closed) Licensee Event Report 201 4-00 6-0 0 , "Main Generator Excitation Transformer Faulted to Ground, Causing Reactor Trip

"

a. Inspection Scope

On December 3, 2014 , a turbine and reactor trip occurred, when the main generator excitation transformer faulted to ground. The reactor trip was classified as uncomplicated and all safety systems performed as designed at the onset of the plant trip. However, during recovery the valve providing flow from the motor-driven auxiliary feedwater pump B to steam generator D (ALHV0005) failed to throttle closed. The problems with ALHV0005 were the subject of a special inspection and were dispo sitioned in NRC Inspection Report 05000483/2015009 (ADAMS Accession Number ML16013A021

). Repair of the excitation transformer was completed and the plant returned to power operations on December 6, 2014.

The construction of the excitation transformer includes high voltage jumper cables between termination points inside its protective enclosure and the winding taps of the transformer coils. The jumper cables are routed above the iron core of the transformer and are supported by insul ating boards and restrained by nylon cable ties. The fault to ground was caused when a jumper cable dropped onto the iron transformer core after failure of the nylon cable ties. The cable ties were an original part of the transformer installed in 2007.

The licensee determined the root cause of the transformer failure was inadequate design (routing cables above the transformer core) and material selection (use of nylon cable ties) during the manufacture of the transformer.

Corrective actions included replacing the nylon cable ties with Tefzel cable ties, which are designed for higher temperatures and longer life expectancy, as well as adding lacing to supplement the Tefzel cable ties. The inspectors reviewed the licensee's submittal along with corrective action documents and determined that the licensee adequately documented the event, including the potential safety consequences and necessary corrective actions. A finding related to a failure to follow the licensee's foreign material exclusion procedure is documented in this section. This licensee event report is closed.

b. Findings

Introduction.

Inspectors reviewed a Green, self

-revealed finding for the licensee's failure to follow the plant procedure for foreign material exclusion. Specifically, after finding foreign material (broken cable ties) within the main generator excitation transformer, established as a foreign material exclusion Level 2 area, the licensee failed to determine the reason for the foreign material and enter the issue into the corrective action program for resolution as required by Procedure AP A-ZZ-00801, "Foreign Material Exclusion," Revision 32.

Description.

On December 3, 2014, an unexpected turbine and reactor trip occurred. The licensee's investigation determined the direct cause of the event was nylon cable tie wraps used to restrain a critical vendor cable failed allowing the cable to fall onto the hot transformer core, where the cable insulation degraded quickly resulting in a phase-to-ground short. The nylon cable ties became brittle from the environmental conditions inside the cabinet.

The licensee's root cause of the event was inadequate design and material selection during the manufacture of the transformer. This transformer was installed in April 2007 to update old and obsolete main generator exciters. The transformer was manufactured and installed by the vendor as a single component. The design used low

-grade nylon cable ties to restrain high voltage jumper cables on insulating boards located above the transformer core. No p reventive maintenance strategy was provided by the transformer manufacturer nor identified by the licensee's engineering personnel.

In July 2013, while the plant was off

-line, the licensee performed an inspection inside the excitation cabinet. The cabinet was identified as a foreign material exclusion Level 2 (FME-2) area and was considered a "standard risk" area. These areas require boundaries and cleanliness controls. While inside the cabinet, an engineer identified several cable ties on the floor of the transformer. The cable ties were very brittle and disintegrated in his hand when he picked them up off of the floor. The engineer was unaware the transformer cabinet was being controlled as a FME

-2 area and did not consider the broken cable ties as foreign material. The engineer notified the engineering "war room" of the issue. The licensee took no further action.

Licensee Procedure APA-ZZ-0080 1, defines foreign material as "Any material that is NOT part of a system or component as designed." Section 4.8 of the procedure also directs individuals that enter an FME

-2 area to Inspect for the presence of any "As

-Found" foreign material WHEN the system or component is initially breached. IF present, retrieve the foreign material in accordance with an approved recovery plan or document the review and approval of system operation with the foreign material in the system. Try to determine the source of, and the reason for, the foreign material. Report the loss of FME integrity in the corrective action request system.

The licensee determined the source of the foreign material, but did not determine the reason for the foreign material nor enter the loss of foreign material exclusion integrity into their corrective action program. As a result, the licensee did not evaluate the condition related to the degradation of nylon cable ties inside the cabinet.

The licensee addressed the issue in Callaway Action Request 201606129. Corrective actions included reminding employees about the importance of foreign material and adherence to the foreign material exclusion procedure.

Analysis.

The licensee's failure to follow the plant procedure for foreign material exclusion was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, after identifying several broken cable ties on the floor inside a FME-2 area the licensee did not determine the reason for the foreign material nor enter the condition into the corrective action program as required by Procedure APA-ZZ-00801. Because the licensee failed to understand what caused the cable tie degradation, a subsequent cable tie failure resulted in a plant trip.

Using Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) for Findings At Power," dated June 19, 2012, the finding was determined to be of very low safety significance because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect of training in the human performance area because the organization did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, several groups within the licensee's organization was unaware the excitation transformer cabinet was classified as an FME

-2 area nor the requirements if foreign material is found within the foreign material exclusion area [H.9].

Enforcement.

Inspectors did not identify a violation of regulatory requirements associated with this finding. Because this finding does not involve a violation and is of very low safety significance, it is identified as:

FIN 05000483/2016002

-05 , "Failure to Follow Plant Foreign Material Exclusion Procedure." These activities constitute d completion of one event follow-up sample, as defined in Inspection Procedure 71153.

4OA 6 Meetings, Including Exit

Exit Meeting Summary

On April 15, 2016, regional inspectors presented the radiation safety inspection results to Mr. T. Hermann, Site Vice President, and Mr.

B. Cox, Senior Director

, Nuclear Operations, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.

On April 22, 2016 , regional inspectors presented the inservice inspection results to Mr. F. Diya, Senior Vice President and Chief Nuclear Office r, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors acknowledged review of proprietary material during the inspection which had been or will be returned to the licensee.

On July 1 9, 2016, the resid ent inspectors presented the inspection results to Mr.

F. Diya, Senior Vice President and Chief Nuclear Officer, and other members of the licensee staff. The licensee acknowledged the issues presented. The licensee confirmed that any proprietary information reviewed by the inspectors had been returned or destroyed.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

K. Blair, Engineer, Steam Generators

B. Cox, Senior Director, Nuclear Operations

D. Davis, Non

-Destructive Testing, Level III

F. Diya, Senior Vice President and Chief Nuclear Officer

T. Elwood, Supervising Engineer, Regulatory Affairs/Licensing

G. Forster, Non

-Destructive Testing Supervisor, Level III

J. Geyer, Manager, Radiation Protection

M. Hoehn II, Engineering Supervisor, Engineering Programs

C. Hendricks, Coordinator, Quality Control

T. Herrmann, Site Vice President

R. Hughey, Manager, Shift Operations

L. Kanuckel, Director, Nuclear Oversight

S. Kovaleski, Director, Engineering

Design

S. McLaughlin, Manager, Performance Improvement

J. Nurrenbern, Program Owner, Boric Acid

S. Petzel, Engineer, Regulatory Affairs

D. Purvis, Supervisor, Quality Control

F. Stuckey, Senior Health Physicis

t

S. Thomure, Training Supervisor, Welding Engineering

T. Trent, Senior Health Physicist, Radiation Protection

M. Vonderhaar, Supervisor, Radiation Protection

R. Wink, Manager, Regulatory Affairs

T. Witt, Engineer, Regulatory Affairs

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000483/2016002

-01 NCV Failure to Account for Water Hammer Stresses in Essential Service Water System Calculations

(Section 1R04)

05000483/2016002

-02 NCV Failure to Meet Applicable ASME Code Requirements for Repairs to Components in the Essential Service Water System (Section 1R07)

05000483/2016002

-0 3 NCV Failure to Adequately Evaluate Operability for a Degraded Condition (Section 1R15)

05000483/2016002

-05 FIN Failure to Follow Plant Foreign Material Exclusion Procedure

(Section 4OA3) Open

05000483/2016002

-04 VIO Failure to Promptly Correct Conditions Adverse to Quality

(Section 4OA2.3)

Closed

05000483/2014

-006-00 LER Main Generator Excitation Transformer Faulted to Ground, Causing Reactor Trip

(Section 4OA3)

Section 1R01: Adverse Weather Protection

Procedures

Number Title Revision

AUE-ADM-2222 Communication and Coordination

AUE-ADM-2223 Disturbance Reporting

AUE-ADM-2227 Reliability Coordination

- Responsibility and Authorities

OSP-NE-00001 Class 1E Electrical Source Verification

OSP-NE-00003 Technical Specification Actions

- A.C. Sources

OTO-MA-00008 Rapid Load Reduction

OTO-ZZ-00012 Severe Weather 33

PDP-ZZ-00027 Seasonal Readiness Program Callaway Action Requests

201508013

201604020

Job s

13000681

Miscellaneou

s Number Title Revision

2016 Summer Reliability Plan

2010009 Health Issue:

Given an EDG HVAC equipment failure, operability cannot be restored within the 72

hour allowed outage time

2015005 Health Issue:

Degradation of ESW Piping in Containment

Section 1R04

Equipment Alignment

Procedures

Number Title Revision

OTN-AL-00001 Auxiliary Feedwater System

OTN-AL-00001 , Checklist 1

Auxiliary Feedwater Valve Alignment

OTN-AL-00001 Checklist 2

MD-AFP A and B Switch Alignment Drawing s Number Title Revision E-012.2-00002 Large Induction Motors Outline

E-21010(Q) DC Main Single Line Diagram

LP-06 NB/NG/NK/NN

-1, Safeguards Power Training Diagram

M-22AL01(Q) Auxiliary Feedwater System Piping and Instrumentation Diagram 46 M-143A-00003 Concentric Restricting Orifice Plates Outline Drawing Miscellaneou

s Number Title Revision

GEK-72150 General Electric Instructions for Class

1E Auxiliary Feedwater Pump Motors Section 1R05

Fire Protection

Procedures

Number Title Revision

APA-ZZ-00703 Fire Protection Operability Criteria and Surveillance Requirements

APA-ZZ-00750 Hazard Barrier Program

EDP-ZZ-04107 HVAC Pressure Boundary Control

OTO-KC-00001 Add A-03 Auxiliary Building 1974'

- Boric Acid Tank Rooms

OTO-KC-00001 Add A-18 Auxiliary Building 2026'

- North Electrical Pen Room

OTO-KC-00001 Add C-15 Control Building 2016' Switchboard and Battery Rooms 2

and 4 0

Procedures

Number Title Revision

OTO-KC-00001 Add C-16 Control Building 2016' Switchboard and Battery Rooms 1

and 3 0

OSP-KC-00015 Fire Door Inspections Drawing s Number Title Revision A-2804 Architectural Fire Delineation Floor Plan, El 2047'

-6" 27

Callaway Action Requests

201605406

Job s

16003139

Miscellaneou

s Number Title Revision

Fire Preplan Manual

KC-64 C-15 Detailed Fire Modeling Report

KC-65 C-16 Detailed Fire Modeling Report

KC-83 Fire Safety Analysis Calculation for Fire Area A

-3 1

KC-98 Fire Safety Analysis Calculation for Fire Area A

-18 1

KC-126 Fire Safety Analysis for Fire Area C

-15 1

KC-102 Fire Safety Analysis Calculation for Fire Area A

-22 1

KC-127 Fire Safety Analysis Calculation for Fire Area C

-16 1

ME-014 Detailed Fire Modeling

Section 1R08: Inservice Inspection Activities

Callaway Action Requests

199800739

199800740

199800741

200207750

200404532

200703197

200703247

200703257

200703491

200810348

200810384

200811050

201003386

201109846

201303346

201303370

201303451

201303502

201303702

201303736

Callaway Action Requests

201406864

201407222

201407245

201407246

201407248

201408130

201500430

201501125

201502944

201503385

201504450

201504861

201504926

201505694

201505757

201506100

201506290

201506544

201507559

201508349

201508887

201600224

201600727

201601320

201601742

201602378

201602824

201603031

201603166

201603256

201603472

201603484

201604058

201604063

201603640

201603661

Drawings

Number Title Revision BG23-H004/231 (Q)

Pip Supports

- CVCS Charging and Excess Letdown Sys. Reactor Building

EF01-C012/311 (Q)

Pipe Supports

- Essential Service Water Sys.

Control Bldg. - Trains A & B

EF02-C003/142 (Q)

Pipe Supports

- Essential Service Water Sys. Aux. Bldg. A Train Supply

EF03-C034/134 (Q)

Pipe Supports

- Essential Service Water Sys. Aux. Bldg. A Train Return

M-22EM01 (Q)

Piping and Instrumentation Diagram High Pressure Coolant Injection System

M-23EF01 Piping Isometric Essential Service Water System Control Building

M-23E F02 Piping Isometric Essential Service Water System Auxiliary Building A Train Supply

M-23E F03 Piping Isometric Essential Service Water System Auxiliary Building A Train Return

M-23E F04 Piping Isometric Essential Service Water System Auxiliary Building B Train Supply

2 M-23E F05 Piping Isometric Essential Service Water System Auxiliary Building B Train Return

M-23E F06 Piping Isometric Essential Service Water System Auxiliary Building A and " Train Supply and Return

M-25BG23 (Q)

Hanger Location Drawing - CVCS Charging & Excess Letdown Reactor Building

Drawings

Number Title Revision M-25EF01 (Q)

Hanger Location Drawing

- Essential Service Water Control Bldg. (A &B Train)

M-25EF02 (Q)

Hanger Location Drawing

- Essential Service Water Sys. Aux. Bldg. A Train Supply

M-25EF03 (Q)

Hanger Location Drawing

- Essential Service Water Sys. Aux. Bldg. A Train Return

Procedures

Number Title Revision

APA-ZZ-00350 Measuring and Test Equipment Program

APA-ZZ-00500 Corrective Action Program

APA-ZZ-00500 , Appendix 1 Operability and Functionality Determinations

APA-ZZ-00500 , Appendix 2 Non-Conforming Materials Report

APA-ZZ-00500 , Appendix 3 Past Operability a nd Reportability Evaluations

APA-ZZ-00500 , Appendix 4 Transient Evaluation

APA-ZZ-00500 , Appendix 5 Maintenance Rule

APA-ZZ-00500 , Appendix 6 Collection and Preservation of Evidence

APA-ZZ-00500 , Appendix 7 Effectiveness Reviews

APA-ZZ-00500 , Appendix 8 Corrective Action Program Training Requirements

APA-ZZ-00500 , Appendix 9 Mitigating Systems Performance Index (MSPI) 7

APA-ZZ-00500 , Appendix 10 Trending Program

APA-ZZ-00500 , Appendix 11 Degraded And Nonconforming Condition Resolution

APA-ZZ-00500 , Appendix 12 Significant Adverse Condition

- Significance Level 1

Procedures

Number Title Revision

APA-ZZ-00500 , Appendix 13 Adverse Condition

- Significance Level 2

APA-ZZ-00500 , Appendix 14 Adverse Condition

- Significance Level 3

APA-ZZ-00500 , Appendix 15 Adverse Condition

- Significance Level 4

APA-ZZ-00500 , Appendix 16 Adverse Condition

- Significance Level 5

APA-ZZ-00500 , Appendix 17 Screening Process Guidelines

APA-ZZ-00500 , Appendix 18 Equipment Performance Evaluation

APA-ZZ-00500 , Appendix 19 Common Cause Evaluation (CCE)

5

APA-ZZ-00500 , Appendix 20 Prompt Human Performance Evaluation (PHPE)

3

APA-ZZ-00500 , Appendix 21 Other Issues

APA-ZZ-00500 , Appendix 22 Corrective Action Program Definitions

APA-ZZ-00661 Administration of Welding 16

APA-ZZ-00661 , Appendix 3

Personnel Approved to Perform Weld Inspections/Examinations

APA-ZZ-00662 ASME Section X

I Repair/Replacement Program

APA-ZZ-00662 , Appendix A

ASME Section X

I Repair/Replacement Program Mandatory Requirements Class 1, 2 And 3 Items a nd Their NF Supports (Fourth Inspection Interval)

APA-ZZ-00662 Appendix B

ASME Section XI Code Cases Applied to the Fourth Inspection Interval 6

APA-ZZ-00662 Appendix E

ASME Section X

I Repair/Replacement Matrix Minor

4

APA-ZZ-00662 Appendix G

ASME Section XI Repair/Replacement Program Mandatory Requirements Class M C a nd CC Items a nd their NF Supports

(Second Inspection Interval)

APA-ZZ-00750 Hazard Barrier Program

EDP-ZZ-0 00 18

Heat Exchanger Eddy Current Testing Methodology

Procedures

Number Title Revision

EDP-ZZ-01004 Boric Acid Corrosion Control Program

EDP-ZZ-01121 Raw Water Systems Predictive Performance Program 21

ESP-ZZ-01016 ASME Section XI IWE Containment Pressure Boundary Inspection

MDP-ZZ-LM001 Fluid Leak Management Program

MSM-ZZ-QW005 Mechanical Snubber Functional Test

MTW-ZZ-WP001 ASME/ANSI General Welding Requirements

MTW-ZZ-WP002 Welder Performance Qualification

MTW-ZZ-WP003 Control Of Welding Filler Materials

MTW-ZZ-WP004 Post Weld Heat Treatment

MTW-ZZ-WP006 Qualification of Welding Procedures

MTW-ZZ-WP007 Callaway Plant Maintenance Welding Procedure AWS D1.1 General Welding Requirements

MTW-ZZ-WP501 Callaway Plant Maintenance Welding Procedure Welding of P

-1 Materials

MTW-ZZ-WP502 Callaway Plan Maintenance Welding Procedure Welding of P

-1 to P-3 Materials

MTW-ZZ-WP50 3 Callaway Plan Maintenance Welding Procedure Welding of P

-1 to P-4 Materials

MTW-ZZ-WP50 4 Callaway Plan Maintenance Welding Procedure Welding of P

-1 to P-5 Materials

MTW-ZZ-WP50 5 Callaway Plan Maintenance Welding Procedure Welding of P

-1 to P-8 Materials

MTW-ZZ-WP50 6 Callaway Plan Maintenance Welding Procedure Welding of P

-4X (Including Welding of P

-1 and P-8 to P-4X) Materials

MTW-ZZ-WP50 9 Callaway Plan Maintenance Welding Procedure Welding of P

-3 Materials

MTW-ZZ-WP5 10 Callaway Plan Maintenance Welding Procedure Welding of P

-4 Materials

MTW-ZZ-WP5 11 Callaway Plan Maintenance Welding Procedure Welding of P

-5 Materials

MTW-ZZ-WP5 1 2 Callaway Plan Maintenance Welding Procedure Welding of P

-5 to P-8 Materials

Procedures

Number Title Revision

MTW-ZZ-WP5 13 Callaway Plan Maintenance Welding Procedure Welding of P

-6 to P-8 Materials

MTW-ZZ-WP5 14 Callaway Plan Maintenance Welding Procedure Welding of P

-8 Materials

MTW-ZZ-WP5 24 Callaway Plan Mechanical Technical Procedure Torch Brazing of Copper Alloys

MTW-ZZ-WP5 25 Callaway Plan Maintenance Welding Procedure Welding of P

-4 to P-8 Materials

MTW-ZZ-WP5 26 Callaway Plan Maintenance Welding Procedure Welding of P

-8 to P-34 Materials

MTW-ZZ-WP5 27 Callaway Plan Maintenance Welding Procedure Welding of P

-34 Materials

MTW-ZZ-WP5 60 Callaway Plan Maintenance Welding Procedure Fusing of High Density Polyethylene (HDPE) Materials for Nuclear Service

MTW-ZZ-WP5 61 Callaway Plan Maintenance Welding Procedure Fusing of High Density Polyethylene (HDPE) Materials for Non

-Nuclear Service

MTW-ZZ-WP 701 AWS Welding of P

-1 Materials

MTW-ZZ-WP702 Callaway Plant Maintenance Technical Procedure

AWS Welding of Studs

PDI-ISI-254-SE Remote Inservice Examination of Reactor Vessel Nozzle to Safe End, Nozzle to Pipe and Safe End to Pipe Welds

PDI-ISI-254-SE-NB Remote Inservice Examination of Reactor Vessel Nozzle to Safe End, Nozzle to Pipe and Safe End to Pipe Welds Using the Nozzle Scanner 0

QCP-ZZ-05000 Liquid Penetrant Examination

QCP-ZZ-05010 Magnetic Particle Examination

QCP-ZZ-05019 Ultrasonic Thickness Measurement

QCP-ZZ-05030 Radiographic Procedure for Examination

o f Weldments and Castings

QCP-ZZ-05041 Visual Examination to ASME VT

-2 26

QCP-ZZ-05048 Boric Acid Walkdown for Reactor Coolant System Pressure Boundary

QCP-ZZ-05049 Reactor Pressure Vessel Head Bare Metal Examination

Procedures

Number Title Revision

UT-2 Ultrasonic Examination of Vessel Welds and Adjacent Base Metal

UT-94 Ultrasonic Examination of Ferritic Piping Welds

UT-95 Ultrasonic Examination of Austenitic Piping Welds

UT-96 Ultrasonic Through Wall Sizing in Piping Welds

UT-103 Ultrasonic Examination of Dissimilar Metal Piping Welds 5

WDI-SSP-1101 Manual Ultrasonic Examination of Reactor Vessel Threads in Flange for Callaway Unit 1

WDI-STD-088 Underwater Remote Visual Examination of Reactor Vessel Internals

WDI-STD-146 ET Examination of Reactor Vessel Pipe Welds Inside Surface 11

Relief Requests Number Title Date Letter:

Michael T. Markley to Fadi Diya Callaway Plant, Unit 1

- Request for Relief 14R

-01, Alternative to ASME Code Inservice Inspection Requirements for Class 3 Buried Piping (TAC NO. MF4271) May 12, 2015

ULNRC-06115 NRC Letter, "Relief Request 13

R-10 for Third 10

-Year Inservice Inspection Interval

- Use of Polyethylene Pipe in Lieu of Carbon Steel Pipe in Buried Essential Service Water Piping System (TAC No. MD6792)," dated November 7, 2008 (Accession No. ML0831 00288) June 10, 20

ULNRC-06146 Ameren Missouri Letter ULNRC

-06115, "10

CFR 50.55a Request:

Proposed Alternative to ASME Section XI

Requirements for Class 3 Buried Piping," dated Ju ne 10, 2014 (ADAMS Accession No.

ML14161A399)

September 30, 2014

UNNRC-06214 Docket Number 50-483 Callaway Plant Unit 1

Union Electric Co.

Facility Operating License N

PF-30 Revision of 10 C

FR 50.55a Request: Proposed Alternative t

o ASME Section X I Requirements for Class 3 Buried Piping (TAC NO. MF4271)

April 24, 2015

Work Packages

15000069-520

15507345 16001742-405

16503498 15000069-505

15507967 16001742-405

16503745

Work Packages

15001243-500 16001742-550 16001743-400

Jobs

10002667

16001870

Miscellaneous

Number Title Revision/Date Various Non Destructive Examination Reports for ESW components

206

EZ-FLO Garlock Sealing Technologies Expansion Joint Test November 15, 2006

0516-19-F01 Secondary Side Visual Inspection Plan for Ameren UE, Callaway RF 21

February 10, 2016

51-9252420-000 AREVA Engineering Information Record: Callaway 1RF021 SG ECT Inspection Plan March 21, 2016

51-9253319-000 AREVA Engineering Information Record: Callaway 1R21 Degradation Assessment April 2016

225-TR-002 Containment F Cooler Response to a Simultaneous

LOCA & LOOP Event

0096-020-CALC-01 Callaway Water Hammer Load Calculation

A190.0002 Procedure Review Form UT

-2 Ultrasonic Examination of Vessel Welds and Adjacent Base Metal, Revision 30

October 8, 2014

A190.0002 Procedure Review Form UT

-94 Ultrasonic Examination of Ferritic Piping Welds , Revision 9

October 8, 2014

A190.0002 Procedure Review Form UT

-95 Ultrasonic Examination of Austenitic Piping Welds , Revision 8 October 8, 2014

A190.0002 Procedure Review Form UT

-96 Ultrasonic Through Wall Sizing in Piping Welds, Revision 7

October 8, 2014 A190.0002 Procedure Review Form UT

-103 Ultrasonic Examination of Dissimilar Metal Piping Welds , Revision 5

October 8, 2014

AP14-008 Self-Assessment: Nuclear Oversight ISI

- IST Audit

October 8, 2014

EDP-ZZ-00016 Self-Assessment:

Checklist for Program Review of Alloy 600 Program October 8, 2014

EDP-ZZ-00016 Self-Assessment: ISI Program June 20, 2014

Miscellaneous

Number Title Revision/Date

RIS 2016-02 OMB Control No. 3150-0011 NRC Regulatory Issue Summary 2016

-02, Design Basis Issues Related to Tube

-To-Tube sheet Joints in Pressurized

-Water Reactor Steam Generators. (ML15169A543)

March 23, 2016

T65.0212 6

Callaway Fall Protection February 14, 2014

Section 1R11

Licensed Operator Requalification Program

Procedures

Number Title Revisio n

ODP-ZZ-00001 Operations Department

- Code of Conduct

OSP-AC-00005 Turbine Actual Overspeed Trip

OTG-ZZ-00005 Plant Shutdown 20% Power to Hot Standby Callaway Action Requests

200601332

201600670

Miscellaneou

s Title Date Dynamic Simulator Exam Scenario, Cycle

16-2 As Found February 1, 2016

Section 1R12

Maintenance Effectiveness

Procedures

Number Title Revision

EDP-ZZ-01128 Maintenance Rule Program

EDP-ZZ-01128 , Appendix 1

SSCs in Scope of the Maintenance Rule at Callaway

EDP-ZZ-01128 , Appendix 4

Maintenance Rule System Functions

Callaway Action Requests

201602435

201602658

201602738

201602824

201603229

201603471

201603472

201603473

201603484

Job s

11504345

16001349

Miscellaneou

s Number Title Revision/Date

Procon1, LLC Evaluation of Room Cooler SGL

-10A Tube Leak Repair April 13, 2016 1784 Union Electric Company Laboratory Services

- Metallurgical Report

- Examination of Failed Room Cooler Tubing September 22, 19 94

04060221 AmerenUE Technical Support Services - Metallurgical Report - Examination of Callaway Room Cooler Tubes September 30, 2004

13050249 Ameren Missouri Technical Support

- Metallurgical Report - Examination of Callaway Room Cooler Tubing May 23, 2013

GL-137 SGL10A/B Room Cooler Heat Removal Capabilities Section 1R13

Maintenance Risk Assessment and Emergent Work Controls

Procedures

Number Title Revision

APA-ZZ-00315 Configuration Risk Management Program

ODP-ZZ-00002, Appendix 1

Protected Equipment Program

ODP-ZZ-00002, Appendix 1, Checklist 5

Placing Train A Protected Equipment Barriers, Mode 5 & 6

ODP-ZZ-00002, Appendix 1, Checklist 7

Placing Train B Protected Equipment Barriers, Mode 5 & 6

Procedures

Number Title Revision

ODP-ZZ-00002, Appendix 1, Checklist 9

Placing Train A Protected Equipment Barriers, Defueled

ODP-ZZ-00002, Appendix 1, Checklist 17

Placing Protected Equipment Barriers for SFP Cooling Outage 1

ODP-ZZ-00002, Appendix 2

Risk Management Actions for Planned Risk Significant Activities

ODP-ZZ-00002, Appendix 2, Checklist 9

Postings for Lowered Inventory Operations Callaway Action Requests

201601830

201602875

201603382

201605725

201605766

Job s

06112970

06116947

10505244

13507816

13507818

14512791

14512792

14512793

14512629

14512630

14512631

14512632

14512774

14512780

14512784

14512873

14513123

14513124

14513125

14512846

14512893

14512923

14513455

14514354

15506373

16003488

16003529

16003530

16003531

Miscellaneou

s Number Title Revision

Shutdown Safety Management Plan

PRAER 16-405 PRA Evaluation Request - Mode Change from Mode 4 to Mode 3 with Equipment OOS

Section 1R15

Operability Evaluations

Procedures

Number Title Revision

KDP-ZZ-00013 Emergency Response Facility and Equipment Evaluation

MTE-ZZ-QA013 MOVATS UDS Testing of Torque Controlled Limitorque Motor Operated Rising Stem Valves

Procedures

Number Title Revision

ODP-ZZ-00002 Equipment Status Control

OSP-EJ-V002A RHR Pump Containment Sump Suction and RWST Suction Inservice Test Drawing s Number Title Revision 8600-X-89645 High Pressure & Low Pressure Nitrogen Gas Storage & Transfer System Site Gas Systems (KH2) Piping and Instrumentation Diagram

E-23BB12A(Q)

RHR Loop 1 Inlet Isolation Valve Schematic Diagram

E-1038-00004 Schematic 7.5kVA Inverter 125VDC, 120VAC, 1PH, 60Hz

- Alarms 1 E-1038-00003 Schematic 7.5kVA Inverter 120VAC, 1ø, 60Hz

E-1038-00006, S002 Outline 7.5kVA Inverter Front Panel Identification

M-22AB02(Q) Main Steam System Piping and Instrumentation Diagram

M-22FA01 Auxiliary Boiler System Piping and Instrumentation Diagram

M-22KH01 Service Gas System Piping and Instrumentation Diagram

M-622.1-00023 Condensing Unit

E-23KJ08A(Q)

Standby Jacket Coolant Heater EKJ01A Schematic Diagram

E-23KJ09B(Q)

Standby Jacket Coolant Circ. Pump PKJ01A Schematic Diagram 2 M-22KJ01(Q) Standby Diesel Generator "A" Cooling Water System Piping and Instrumentation Diagram Callaway Action Requests

201603312

201603353

201603598

201603711

201603739

201603758

201604998

201605016

201605045

201605324

201605917

201105227

Job s

10507721

10507762

13505626

14511766

16001888

16002253

16002356

16003607

Miscellaneou

s Number Title Revision

BO-05 Addendum 19

Revised Temperatures for 3601, 3605, and 3609 for Station Black Out 1

BO-07 Control Room SBO Heat Load Calculation

EF-123 UHS Thermal Performance Analysis using GOTHIC 7.2(b) CAR#201001813

RFR 17478 Perform Evaluation for NRC GL96

-06 Response

C RFR 201603756

Request for Resolution: Modify low pressure nitrogen system piping and penetrations Section 1R18

Plant Modifications

Procedures

Number Title Revision

APA-ZZ-00600 Design Change Control

EDP-ZZ-04015 Evaluating and Processing Requests for Resolution (RFR)

Drawing s Number Title Revision M-22AL01(Q) Auxiliary Feedwater System Piping and Instrumentation Diagram 46 M-22AN01 Demineralized Water Storage and Transfer System Piping and Instrumentation Diagram

M-22AP01 Condensate Storage and Transfer System Piping and Instrumentation Diagram

M-22AP02 Hardened Condensate Storage Tank Composite Piping and Instrumentation Diagram

M-22AQ02 Feedwater Chemical Addition System Piping and Instrumentation Diagram

M-22KA09 Instrument Air System Piping and Instrumentation Diagram Miscellaneou

s Number Title Revision/Date

50.59 Screen for MP 13

-0033 Hardened Condensate Storage Tank Refuel 21 Tie

-Ins 4

Applicability Determination for MP 13

-0033 Hardened Condensate Storage Tank Refuel 21 Tie

-Ins 4

Miscellaneou

s Number Title Revision/Date

Evaluation of Scissor Lift Impact on HCST

May 6, 2016 16-05 50.59 Evaluation for MP 13

-0033 Hardened Condensate Storage Tank Refuel 21 Tie

-Ins 4

MP 13-0033 Hardened Condensate Storage Tank Refuel 21 Tie

-Ins 4

Section 1R19

Post-Maintenance Testing

Procedures

Number Title Revision

APA-ZZ-00100 Written Instructions Use and Adherence

APA-ZZ-00320 Work Execution

APA-ZZ-00322 Appendix C

Job Planning

MTE-ZZ-QA013 MOVATS UDS Testing of Torque Controlled Limitorque Motor Operated Rising Stem Valves

OSP-JE-00001 Emergency Fuel Oil Transfer Pumps Cross

-connection Line Fill Verification Test

OSP-NE-0001A Standby Diesel Generator A Periodic Tests

OTN-NB-0001A Addendum 3

NB01 transfer to XNB02 Single Offsite Source Operation and Restoration

OTN-NE-0001A Standby Diesel Generation System

-Train A 48

Drawing s Number Title Revision E-23BB12A(Q)

RHR Loop 1 Inlet Isolation Valve Schematic Diagram

M22-KH01 Service Gas System Piping and Instrumentation Diagram Callaway Action Requests

201602435

201603496

201603598

201603758

201604092

201605141

201605393

Job s

10507721

10507762

16001888

16001887

16001349

14005657

15505373

13505566

14511620

16002253

Job s

16003027

Section 1R20

Refueling and Other Outage Activities

Procedures

Number Title Revision

APA-ZZ-00908 Fitness for Duty Programs

APA-ZZ-00911 Fatigue Management

ESP-ZZ-00024 Low Power Physics Testing Data Acquisition

OSP-SA-00004 Visual Inspection of Containment for Loose Debris

OTG-ZZ-00001 Plant Heatup Cold Shutdown to Hot Standby 85

OTG-ZZ-00002 Reactor Startup

- IPTE 57

OTG-ZZ-00003 Plant Startup Hot Zero Power to 30 Percent Power

- IPTE 60

OTG-ZZ-00005 Plant Shutdown 20 Percent Power to Hot Standby

OTG-ZZ-00006 Plant Cooldown Hot Standby to Cold Shutdown

OTG-ZZ-00007 Refueling Preparation, Performance and Recovery Callaway Action Requests

201600506

201603464

201603496

201603498

201603531

201603598

201603725

201603729

201603739

201603799

201603889

201603909

201603917

201603931

Section 1R22

Surveillance Testing

Procedures

Number Title Revision

APA-ZZ-00350 Measuring and Test Equipment Program

OSP-BN-V0005 BN Suction Header Valves Inservice Test

OSP-EJ-0006A RHR Mini Flow Valve Time Response Test Train A

OSP-EJ-0006B RHR Mini Flow Valve Time Response Test Train B

OSP-EJ-PV04A Train A RHR and RCS Check Valve Inservice Test

OSP-EJ-PV04B Train B RHR and RCS Check Valve Inservice Test

OSP-EJ-V002B RWST to RHR Suction Check Valve Inservice Test

Procedures

Number Title Revision

OSP-EM-P0002 Train A and Train B Safety Injection Comprehensive Pump Test 9

OSP-EM-V0003 ECCS Check Valve Inservice Test

OSP-EM-V003A CCP A and B Full Flow Test

OSP-EM-V0004 RHR Check Valve and SI Pump Recirc Valve Inservice Test

OSP-EM-V0005 EM8922A and EM8922B Closure Inservice Test

OSP-EP-V0006 SI Accumulator Discharge Check Valve Test

OSP-NE-0001B Standby Diesel Generator B Periodic Tests

OSP-SA-2413B Train B Diesel Generator and Sequencer Testing

OTN-NE-0001B Standby Diesel Generation System

- Train B 51

OTS-SB-0002B SSPS Train B Operation in Modes 5, 6, and No Mode Callaway Action Requests

201604838

201508227

201503020

Job s

10506673

13504474

13504816

14511319

14511384

14511393

14511394

14511398

14511402

14511437

14511604

14511834

14512880

16507235 15004983

Section 2RS1: Radiological Hazard Assessment and Exposure Controls

Procedures

Number Title Revision

APA-ZZ-00014 Conduct of Operations

- Radiation Protection

APA-ZZ-01000 Callaway Energy Center Radiation Protection Program

APA-ZZ-01004 Radiological Work Standards

HDP-ZZ-01200 Radiation Work Permi ts 29

HDP-ZZ-01500 Radiological Postings

HDP-ZZ-03000 Radiological Survey Program

HDP-ZZ-03000 APPA Frequency and Location of Routine Radiological Surveys

HTP-ZZ-02004 Control of Radioactive Sources

Procedures

Number Title Revision

HTP-ZZ-06001 High Radiation / Locked High Radiation / Very High Radiation Area Access Callaway Action Requests

201507836

201507921

201508154

201508367

201508546

201508801

201600369

201601938

201602105

201602672

Specific Radiation Work Permits Number Title Revision

13005670 Replace Valves BGV001, BGV002, and BGV003

14006281 BB8948D Maintenance , Disassemble, Inspect, Repair leak-by and Reassemble Check Valve BB8948D

14006280 BB8949D Disassembly and Repair , Remove/Reinstall Insulati on, Disassemble, Repair Leak, Clean Studs, Reassemble, Perform VT

-1 and

VT-3 Inspection and Engineering Oversight

210803625 Motor Change on B Reactor Coolant Pump and Associated Tasks 1 15001126500

Replace BBV0400

Radiation Survey Records Survey Number Title Date

01181621 Fuel Building 2047' December 27, 2012

CA-M-20140715-4 RW7225 Low Level Drum Storage Area July 15, 2014

CA-M-20150821-4 1106 Moderating Heat Exchanger Room

- Deposit from

HRA August 21, 2015

CA-M-20151119-11 1124 Valve Area BACC Walkdown, Job

15505065 November 19, 2015

CA-M-20160104-5 1322 South Piping Pen Monthly Routine January 4, 2016

CA-M-20160203-1 7225 Low Level Drum Storage Area February 3, 2016

CA-M-20160402-8 RB2000 Initial Entry General Area for RFO21

April 2, 2016

CA-M-20160404-1 1322 South Piping Penetration Rm

- Down Posting April 4, 2016

Radiation Survey Records Survey Number Title Date

CA-M-20160404-25 1323 North Piping Penetration Room April 4, 2016

CA-M-20160408-33 RB2026VC Pre

-job

BGV-001, 002, 003

April 8, 2016

CA-M-20160409-9 1124 Valve Compartment Hold Off, Job

10505104 April 9, 2016

CA-M-20160410-29 RB2026VC 14512081/500 Pre

-shielding survey April 10, 2016

CA-M-20160411-33 RB2000 Routine Daily April 11, 2016

CA-M-20160412-5 RB2026VC Letdown Valve Cubicle fit

-up and welding of new

BGV-001 valve and piping April 12, 2016

Air Sampling Sample Number Location Date 1604101612

Cavity April 10, 2016 1604111442

RB 2026 Letdown Cubicle April 11, 2016

1604120400

RB 2026 April 12, 2016

1604121345

BB8948D RB 2000

April 12, 2016

1604121800

D SG Manway April 13, 2016

1604122215

BB8949D April 13, 2016

Miscellaneous

Number Title Date

Accountable Source Inventory List Custodial Source Inventory List

15507830

HSP-ZZ-00001: Sealed Beta

-Gamma Source Leak Test January 19, 2016

Section 2RS3:

In

-plant Airborne Radioactivity Control and Mitigation

Procedures

Number Title Revision

HDP-ZZ-08000 Respiratory Protection Program

HDP-ZZ-08002 Respiratory Protection Issue and Use

HTP-ZZ-08203-DTI-REGULATORS

Testing Scott Regulators And Respirators Using The Biosystems Posichek3 Tester

Procedures

Number Title Revision

HTP-ZZ-08208-DTI-FITPRO-TESTING Quantitative Respirator Fit Testing Using The Tsi Portacount Pro System

HTP-ZZ-08208-DTI-FIT-TESTING Quantitative Respirator Fit Testing Using The Tsi Portacount Plus System

HTP-ZZ-08300-DTI-AIRPAK75 Scott Air-Pak 75 SCBA Respirator Inspection and Storage 9

HTP-ZZ-08300-DTI-POST HYDRO

Post Hydrostatic Testing of Breathing Air Cylinders

HTP-ZZ-08300-DTI-SKAPAK

SKA-PAK at SCBA Respirator Storage and Inspection

HTP-ZZ-08301-DTI-RESPRO CLEAN

Manual Cleaning of Respiratory Protection Equipment

HTP-ZZ-0830 1-DTI-SCOTT-RES-CLEAN Manual Cleaning of Scott Mask Mounted Regulator

HTP-ZZ-08501-DTI-AIR TEST Testing of Breathing Air

HTP-ZZ-08502-DTI-MAC-CAL Scott Mobile Air Cart Calibration

HTP-ZZ-08503-DTI-UNIIICOMPRESSOR

Operation of Bauer UNICUS III, 25 CFM Breathing Air Compressor and Breathing Air Cascade System

RP-DTI-RESPRO-STORAGE Storage of Respirators Callaway Action Requests

2014 07682

201407882

201408905

201500688

201501023

201502128

201502189

201502356

201503288

201503299

201503490

201600547

201600548

Title Date SCBA and Ska

-Pak CBT Records March 9, 2016

Ska-Pak Proficiency Certification Record March 9, 2016

Breathing Air Sample Data Sheet March 26, 2014

Breathing Air Sample Data Sheet June 26, 2014

Breathing Air Sample Data Sheet September 12, 2014

Breathing Air Sample Data Sheet December 29, 2014

Title Date Breathing Air Sample Data Sheet March 17, 2015

Breathing Air Sample Data Sheet June 19, 2015

Breathing Air Sample Data Sheet September 22, 2015

Breathing Air Sample Data Sheet December 15, 2015

Breathing Air Sample Data Sheet March 7, 2016

Training Certificates Number Title Date Technician A

Air-Pak 2.2/3.0/4.5/Fifty/75 SCBA Maintenance and Overhaul September 20, 2016

Technician B

Air-Pak 2.2/3.0/4.5/Fifty/75 SCBA Maintenance and Overhaul July 13, 2017

Miscellaneous

Title Date Respiratory Protection Maintenance Records

2014-2015 Respiratory Protection Equipment Inspection Record April 2015

- March 2016

Section 4OA1: Performance Indicator Verification

Procedures

Number Title Revision

RRA-ZZ-00001 NRC Performance Indicator Program

OSP-BB-00009 RCS Inventory Balance Callaway Action Requests

201502229

201505332

201505796

Job s

16503927

Miscellaneou

s Number Title Revision Date

Mitigating Systems Performance Index (MSPI) Basis Document 16

Miscellaneou

s Number Title Revision Date

NRC Performance Indicator Transmittal Report, Second Quarter 2015, Mitigating Systems Cornerstone July 9, 2015

NRC Performance Indicator Transmittal Report, Third Quarter 2015, Mitigating Systems Cornerstone October 12, 2015

NRC Performance Indicator Transmittal Report, Fourth Quarter 2015, Mitigating Systems Cornerstone January 11, 2016

NRC Performance Indicator Transmittal Report, First Quarter 2016, Mitigating Systems Cornerstone April 13, 2016

MSPI Derivation Report, MSPI Heat Removal System, Unavailability Index (UAI)

June 2015

MSPI Derivation Report, MSPI Heat Removal System, Unreliabil ity Index (URI)

June 2015

MSPI Derivation Report, MSPI Heat Removal System, Unavailability Index (UAI) September

2015

MSPI Derivation Report, MSPI Heat Removal System, Unreliabil ity Index (URI)

September

2015

MSPI Derivation Report, MSPI Heat Removal System, Unavailability Index (UAI)

December 2015

MSPI Derivation Report, MSPI Heat Removal System, Unreliabil ity Index (URI)

December 2015

MSPI Derivation Report, MSPI Heat Removal System, Unavailability Index (UAI)

March 2015

MSPI Derivation Report, MSPI Heat Removal System, Unreliabil ity Index (URI)

March 2015

Reactor Coolant System Identified Leakage Data April 1, 2015 through March 30, 2016

NRC Performance Indicator Transmittal Report, Second Quarter 2015, Barrier Integrity Cornerstone July 6, 2015

NRC Performance Indicator Transmittal Report, Third Quarter 2015, Barrier Integrity Cornerstone October 12, 2015

NRC Performance Indicator Transmittal Report, Fourth Quarter 2015, Barrier Integrity Cornerstone January 11, 2016

NRC Performance Indicator Transmittal Report, First Quarter 2016, Barrier Integrity Cornerstone April 8, 2016

LER 2015-001-00 Licensee Event Report

- Completion of a Shutdown Required by the Technical Specifications

Miscellaneou

s Number Title Revision Date

LER 2015-002-00 Licensee Event Report

- Manual Auxiliary Feedwater Actuation 0

LER 2015-003-00 Licensee Event Report

- Reactor Trip Caused b

y Transmission Line Fault

LER 2015-003-01 Licensee Event Report

- Reactor Trip Caused by Transmission Line Fault

LER 2015-004-00 Licensee Event Report

- Auxiliary Feedwater Flow Control Valve Inoperable due to Faulty Electronic Positioner Card

Section 4OA2: Identification and Resolution of Problems

Procedures

Number Title Revision

APA-ZZ-00500 , Appendix 8 Corrective Action Program Training Requirements

APA-ZZ-00500 , Appendix 9 Mitigating Systems Performance Index (MSPI)

APA-ZZ-00500 , Appendix 10 Trending Program

APA-ZZ-00500 , Appendix 11 Degraded And Nonconforming Condition Resolution

APA-ZZ-00500 , Appendix 12 Significant Adverse Condition

- Significance Level 1

APA-ZZ-00500 , Appendix 13 Adverse Condition

- Significance Level 2

APA-ZZ-00500 , Appendix 14 Adverse Condition

- Significance Level 3

APA-ZZ-00500 , Appendix 15 Adverse Condition

- Significance Level 4

APA-ZZ-00500 , Appendix 16 Adverse Condition

- Significance Level 5

APA-ZZ-00500 , Appendix 17 Screening Process Guidelines

APA-ZZ-00500 , Appendix 18 Equipment Performance Evaluation

Procedures

Number Title Revision

APA-ZZ-00500 , Appendix 19 Common Cause Evaluation (CCE)

APA-ZZ-00500 , Appendix 22 Corrective Action Program Definitions

APA-ZZ-00600 Design Change Control Drawing s Number Title Revision M-22AE01 Piping and Instrumentation Diagram Service Water System Callaway Action Requests

201010634

20160440

201602658

201603472

201605488

201109846

201110442

201202852

201303346

201303370

201303451

201303502

201303608

201303702

201303736

201307879

201309041

201309046

201400458

201402778

201406213 2014072222

201407248

201407246

201407245

201503637

201602824

201603119

201603346

201603472

201603471

201603472

201603484

201603526

201604063

201604058

201604092

201604297

201604235

201604378

Jobs

16002133

16002339

Miscellaneou

s Number Title Revision

MP 10-0003 Install Service Water Check Valves to Minimize ESW Water Hammer During LOOP and ESFAS Testing

MP 10-0004 Revise Sequencer Operation of EFHV0037 and EFHV0038

Section 4OA3:

Event Follow

-Up Procedures Number Title Revision

APA-ZZ-00500 Corrective Action Program

Procedures

Number Title Revision

APA-ZZ-00801 Foreign Material Exclusion Callaway Action Requests

200603505

201408897

201606129

Job s

11509869

13004764

Miscellaneou

s Number Title Revision E-1051-00104 IM for Dry Type Transformer Installation

Attachment 2

The following items are requested for the Occupational Radiation Safety Inspection at Callaway Plant

(April 11

- 15, 2016) Integrated Report 2016002

Inspection areas are listed in the attachments below.

Please provide the requested information on or before March 21, 2016

.

Please submit this information using the same lettering system as below.

For example, all contacts and phone numbers for Inspection Procedure 71124.01 should be in a file/folder titled

"1- A," applicable organization charts in file/folde

r "1- B," etc.

If information is placed on ims.certrec.com, please ensure the inspection exit date entered is at least 30 days later than the onsite inspection dates, so the inspectors will have access to the information while writing the report.

In addition to the corrective action document lists provided for each inspection procedure listed below, please provide updated lists of corrective action documents at the entrance meeting.

The dates for these lists should range from the end dates of the original lists to the day of the entrance meeting.

If more than one inspection procedure is to be conducted and the information requests appear to be redundant, there is no need to provide duplicate copies.

Enter a note explaining in which file the information can be found.

If you have any questions or comments, please contact the lead inspector, Pete Hernandez at (817) 200-1168 or Pete.Hernandez@nrc.gov.

PAPERWORK REDUCTION ACT STATEMENT

This letter does not contain new or amended information collection requirements subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.).

Existing information collection requirements were approved by the Office of Management and Budget, control number 3150

-0011.

1. Radiological Hazard Assessment and Exposure Controls (71124.01)

Date of Last Inspection:

October 26, 2015

A. List of contacts (with official title) and telephone numbers for the Radiation Protection Organization Staff and Technicians B. Applicable organization charts

C. Audits, self

-assessments, and LERs written since date of last inspection

, related to this inspection area

D. Procedure indexes for the radiation protection procedures

E. Please provide specific procedures related to the following areas noted below.

Additional Specific Procedures may be requested by number after the inspector reviews the procedure indexes.

1. Radiation Protection Program Description

2. Radiation Protection Conduct of Operations

3. Personnel Dosimetry Program

4. Posting of Radiological Areas

5. High Radiation Area Controls

6. RCA Access Controls and Radworker Instructions

7. Conduct of Radiological Surveys

8. Radioactive Source Inventory and Control

9. Declared Pregnant Worker Program F. List of corrective action documents (including corporate and subtiered systems) since date of last inspection

a. Initiated by the radiation protection organization

b. Assigned to the radiation protection organization

c. Identify any CRs that are potentially related to a performance indicator event

NOTE: The lists should indicate the significance level of each issue and the search criteria used.

Please provide documents which are "searchable" so that the inspector can perform word searches.

If not covered above, a summary of corrective action documents since date of last inspection involving unmonitored releases, unplanned releases, or releases in which any dose limit or administrative dose limit was exceeded (for Public Radiation Safety Performance Indicator verification in accordance with IP 71151)

G. List of radiologically significant work activities scheduled to be conducted during the inspection period (If the inspection is scheduled during an outage, please also include a list of work activities greater than 1 rem, scheduled during the outage with the dose estimate for the work activity.)

H. List of active radiation work permits

I. Radioactive source inventory list

3.

In-Plant Airborne Radioactivity Control and Mitigation (71124.03)

Date of Last Inspection:

October 27, 2014

A. List of contacts and telephone numbers for the following areas:

1. Respiratory Protection Program

2. Self-contained breathing apparatus

B. Applicable organization charts

C. Copies of audits, self

-assessments, vendor or NUPIC audits for contractor support (SCBA), and LERs, written since date of last inspection related to:

1. Installed air filtration systems

2. Self-contained breathing apparatuses

D. Procedure index for:

1. use and operation of continuous air monitors

2. use and operation of temporary air filtration units

3. Respiratory protection

E. Please provide specific procedures related to the following areas noted below.

Additional Specific Procedures may be requested by number after the inspector reviews the procedure indexes.

1. Respiratory protection program

2. Use of self

-contained breathing apparatuses

3. Air quality testing for SCBAs

F. A summary list of corrective action documents (including corporate and subtiered systems) written since date of last inspection

, related to the Airborne Monitoring program including:

1. continuous air monitors

2. Self-contained breathing apparatuses

3. respiratory protection program

NOTE: The lists should indicate the significance level of each issue and the search criteria used.

Please provide documents which are "searchable."

G. List of SCBA qualified personnel

- reactor operators and emergency response personnel

H. Inspection records for SCBAs staged in the plant for use since date of last inspection.

I. SCBA training and qualification records for control room operators, shift supervisors, STAs, and OSC personnel for the last year.

A selection of personnel may be asked to demonstrate proficiency in donning, doffing, and performance of functionality check for respiratory devices.