ML16225A577

NRC Integrated Inspection Report 05000483/2016002 and Notice of Violation

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

See also: IR 05000483/2016002

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

violations in accordance Section 2.3.2.a of the NRC Enforcement Policy, which appears on the

NRCs 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 NRCs 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.

F. Diya -2-

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 NRCs Public Document Room or from the Publicly Available Records (PARS)

component of the NRC's Agencywide Documents Access and Management System (ADAMS).

ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the

Public Electronic Reading Room).

Sincerely,

/RA David Proulx Acting for/

Nicholas H. Taylor, Branch Chief

Project Branch B

Division of Reactor Projects

Docket No. 50-483

License No. NPF-30

Enclosures:

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

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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

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 NRCs 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

-1- Enclosure 1

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 12th day of August 2016

-2-

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket: 05000483

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. ODonnell, Health Physicist

Approved By: Nicholas H. Taylor

Chief, Project Branch B

Division of Reactor Projects

-1- Enclosure 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 1 and June 30,

2016, by the resident inspectors at the Callaway Plant and inspectors from the NRCs Region IV

office. Five findings of very low safety significance (Green) are documented in this report. Four

of these findings involved violations 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 NRCs 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 licensees 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 APA-ZZ-00801, Foreign Material Exclusion, Revision 32.

The licensees 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 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 licensees organization were unaware

the excitation transformer cabinet was classified as a foreign material exclusion Level 2 area

nor the requirements if foreign material is found within the foreign material exclusion area

[H.9]. (Section 4OA3)

-2-

Cornerstone: Mitigating Systems

• Green. The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B,

Criterion III, Design Control, for the licensees 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 licensees 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 Requests 201603472 and 201603819.

The inspectors determined that the licensees 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 licensees 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 licensees failure to repair various ASME Code Class 3 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. 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.

-3-

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 licensees 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 licensees 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 licensees 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 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

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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 licensees 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 licensees 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 licensees 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 licensees 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 issue 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 licensees 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

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available and adequately maintained to respond during a loss of off-site power event [H.1].

(Section 4OA2.3)

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PLANT STATUS

Callaway began the inspection period at 86 percent power while coasting down at the end of the

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

1. REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 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 stations 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 licensees 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 stations readiness for

impending adverse weather conditions. The inspectors reviewed plant design features,

the licensees procedures to respond to severe weather including thunderstorms,

tornadoes and high winds, and the licensees implementation of these procedures. The

inspectors evaluated operator staffing and accessibility of controls and indications for

those systems required to control the plant.

-7-

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.

1R04 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 licensees 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 licensees 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 licensees 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 Callaways response to NRC Generic Letter 96-06,

Assurance of Equipment Operability and Containment Integrity during Design-Basis

Accident Conditions, January 28, 1997. Additionally, there was external operating

experience concerning water hammer phenomena and the impact on system piping.

-8-

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 piping

systems perform their safety-related functions during plant normal, upset, and faulted

conditions. Pipes are subject to various loading conditions like pressures, dead 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 FTHs 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.

-9-

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 systems

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 failed 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 inspectors determined that the licensees 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

- 10 -

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 licensees 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 licensees 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.

- 11 -

1R05 Fire Protection (71111.05)

Quarterly Inspection

a. Inspection Scope

The inspectors evaluated the licensees 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)

• June 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 licensees 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.

1R07 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 exchangers 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 constituted completion of two heat sink performance annual review

samples, 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 licensees failure to repair various ASME Code Class 3

- 12 -

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 licensees 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 required 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

- 13 -

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 licensees 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

- 14 -

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 licensees 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.

1R08 Inservice Inspection Activities (71111.08)

The activities described below constitute completion of two inservice inspection samples,

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 Report Number 5010-16-0057 Magnetic Particle

Feedwater Condensate Storage Tank to Auxiliary

System Feedwater Header Isolation Valve,

Field Weld-25 (Component ALV0202)

Auxiliary Report Number 5010-16-0058 Magnetic Particle

Feedwater Condensate Storage Tank to Auxiliary

System Feedwater Header Isolation Valve,

Field Weld-26 (Component ALV0202)

Auxiliary Report Number 5010-16-0059 Magnetic Particle

Feedwater Condensate Storage Tank to Auxiliary

System Feedwater Header Isolation Valve,

Field Weld-27 (Component ALV0202)

Auxiliary Report Number 5010-16-0060 Magnetic Particle

Feedwater Condensate Storage Tank to Auxiliary

System Feedwater Header Isolation Valve,

Field Weld-28 (Component ALV0202)

Auxiliary Report Number 5010-16-0061 Magnetic Particle

Feedwater Condensate Storage Tank to Auxiliary

System Feedwater Header Isolation Valve,

Field Weld-29 (Component ALV0202)

- 15 -

SYSTEM WELD IDENTIFICATION EXAMINATION TYPE

Safety Injection Report Number 5000-16-0010 Penetrant

System Safety Injection Accumulator D Outlet,

Upstream Check Valve Test Line

Isolation Valve, Field Weld-01

(Component EPHV8877D)

Safety Injection Report Number 5000-16-0011 Penetrant

System Safety Injection Accumulator D Outlet,

Upstream Check Valve Test Line

Isolation Valve, Field Weld-02

(Component EPHV8877D)

Safety Injection Report Number 5000-16-0012 Penetrant

System Safety Injection Accumulator D Outlet,

Upstream Check Valve Test Line

Isolation Valve, Field Weld-03

(Component EPHV8877D)

Reactor Coolant Record Number 5030-16-012 Radiograph

System 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

Joints 16001742-405-FW-05 and 06)

Reactor Coolant Record Number 5030-16-014 Radiograph

System Reactor Coolant System Pressurizer

Chemical and Volume Control System

Auxiliary Spray Supply Drain

(Component BBV0400)

Reactor Coolant Record Number UT-16-024 Ultrasonic

System Reactor Pressure Vessel Stud Number 1

(Component 2-CH-STUD-01)

Reactor Coolant Record Number UT-16-025 Ultrasonic

System Reactor Pressure Vessel Stud Number 2

(Component 2-CH-STUD-02-R1)

Reactor Coolant Record Number UT-16-026 Ultrasonic

System Reactor Pressure Vessel Stud Number 3

(Component 2-CH-STUD-03)

- 16 -

SYSTEM WELD IDENTIFICATION EXAMINATION TYPE

Reactor Coolant Record Number UT-16-050 Ultrasonic

System Reactor Pressurizer Safety Nozzle A

Inner Radius Area Examination

(Component 2-BB03-10A-A-IR,

Exam Angle 55° + 38°)

Reactor Coolant Record Number UT-16-050 Ultrasonic

System Reactor Pressurizer Safety Nozzle A

Inner Radius Area Examination

(Component 2-BB03-10A-A-IR,

Exam Angle 55° - 38°)

Reactor Coolant Record Number UT-16-052 Ultrasonic

System Reactor Pressurizer Safety Nozzle B

Inner Radius Area Examination

(Component 2-BB03-10B-B-IR,

Exam Angle 55° + 38°)

Reactor Coolant Record Number UT-16-052 Ultrasonic

System Reactor Pressurizer Safety Nozzle B

Inner Radius Area Examination

(Component 2-BB03-10B-B-IR,

Exam Angle 55° - 38°)

Reactor Coolant Record Number UT-16-053 Ultrasonic

System Reactor Pressurizer Safety Nozzle B

to Top Head Weld

(Component 2-TBB03-10B-B-W,

Exam Angle 55° - 38°)

Reactor Coolant Acquisition Log No. DM/Pipe 22-1 Ultrasonic

System 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)

Reactor Coolant Acquisition Log No. DM/Pipe 158-1 Ultrasonic

System 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)

- 17 -

SYSTEM WELD IDENTIFICATION EXAMINATION TYPE

Reactor Coolant Acquisition Log No. DM/Pipe 202-1 Ultrasonic

System 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)

Reactor Coolant Acquisition Log No. DM/Pipe 338-1 Ultrasonic

System 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)

Safety Injection Report Number 5041-16-0020 Visual

System Safety Injection Pumps - Crosstie to Cold

Leg Loops Numbers 1, 2, 3, and 4

(Component Location P049)

Reactor Coolant Report Number 5041-16-0021 Visual

System Reactor Pressure Vessel Head

(Component RBB01)

Essential Record Number 5042-16-0035 Visual

Service Water Essential Service Water System Support

System (Component EF02C003142)

Essential Record Number 5042-16-0036 Visual

Service Water Essential Service Water System Support

System Hanger (Component EF03C034134)

Essential Record Number 5042-16-0037 Visual

Service Water Essential Service Water System Support

System (Component EF01C012311)

Emergency Record Number 5042-16-0038 Visual

Diesel Diesel Generator A Jacket Water Heat

Generator Exchanger Supports (Component EKJ06A)

Emergency Record Number 5042-16-0039 Visual

Diesel Diesel Generator A Jacket Water Heat

Generator Exchanger Supports (Component EJH06A)

- 18 -

SYSTEM WELD IDENTIFICATION EXAMINATION TYPE

Chemical and Report Number 5042-16-0056 Visual

Volume Control Chemical and Volume Control System

System Pipe Support (Component BG23H004231)

The inspectors reviewed records for the following nondestructive examinations:

SYSTEM IDENTIFICATION EXAMINATION TYPE

Condensate Report Number 5010-16-0040 Magnetic Particle

System High Pressure Condensate Main Steam

Dump Valve Low Point Drain Steam Trap

Bypass Valve (Component ABV0184)

Auxiliary Report Number 5010-16-0042 Magnetic Particle

Feedwater Condensate Storage Tank to Auxiliary

System Feedwater Pump Suction Check Valve

(Component ALV0217)

Auxiliary Report Number 5010-16-0048 Magnetic Particle

Feedwater Auxiliary Feedwater System 3-inch

System Tee to 3-inch Spool Piece

(Job Number 15001243, Field

Weld FW-16)

Auxiliary Report Number 5010-1-0049 Magnetic Particle

Feedwater Hardened Condensate Storage Tank

System to Auxiliary Feedwater Pump Header

Isolation Valve (Component ALV0202,

Job Number 15000069, Field

Weld FW-30)

Safety Injection Report Number 5000-16-0008 Penetrant

System Safety Injection Pump B Loop 4 Hot Leg

Test Line Isolation HV

(Component EMHV8889D)

Safety Injection Report Number 5000-16-0010 Penetrant

System Safety Injection Accumulator D Outlet

Upstream Check Valve Test Line Isolation

(Component EPHV8877D, Downstream

Side of Valve)

- 19 -

SYSTEM IDENTIFICATION EXAMINATION TYPE

Safety Injection Report Number 5000-16-0011 Penetrant

System Safety Injection Accumulator Outlet

Upstream Check Valve Test Line Isolation

(Component EPHV8877D, Upstream

Side of Valve)

Chemical and Report Number 5000-16-0018 Chemical Penetrant

Volume Control and Volume Control System Letdown

System Throttle Valve B (Component BGV0002)

Reactor Coolant Record Number 5030-16-010 Radiograph

System 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)

Reactor Coolant Record Number 5030-16-011 Radiograph

System 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-02)

Reactor Coolant Report Number 5042-16-028 Visual

System Reactor Pressure Vessel Head

(Component RBB01, Second Inspection)

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.

- 20 -

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

SYSTEM WELD IDENTIFICATION WELD TYPE

Reactor Coolant Valve BBV-0400, Reactor Coolant Manual Gas Tungsten Arc

System System Pressurizer Chemical and Welding

Volume Control System Auxiliary

Spray Supply Drain

(Job 15001126-500, ASME Code

Class 2, Field Weld FW-03)

Chemical and Valve BGV-0003, CVCS Letdown Manual Gas Tungsten Arc

Volume Control Orifice A Outlet Throttle Valve Piping Welding

System (Job 13005673-510, ASME Code

Class 2, Field Weld FW-03, -04

and -05)

Chemical and Valve BGV-0002, CVCS Letdown Manual Gas Tungsten Arc

Volume Control Orifice A Outlet Throttle Valve Piping Welding

System (Job 13005672-510, ASME Code

Class 2, Field Weld FW-01, -02,

and -03)

Auxiliary Hardened Condensate Storage Manual Gas Tungsten Arc

Feedwater Tank Re-Circulation Line And Welding

System Tie-In to Existing Auxiliary

Feedwater System Piping

(Job 15001243-500, Field Welds

FW-11, -12, -13, -14, -15, and -16)

The inspectors reviewed records of the following welding activities:

SYSTEM WELD IDENTIFICATION WELD TYPE

Chemical and Valve BGV-0001, CVCS Letdown Manual Gas Tungsten Arc

Volume Control Orifice A Outlet Throttle Valve Piping Welding

System (Job 13005670-510, ASME Code

Class 2, Field Weld FW-03, -04,

and -05)

Chemical and Valve BGV-0001, CVCS Letdown Manual Gas Tungsten Arc

Volume Control Orifice A Outlet Throttle Valve Piping Welding

System (Job 13005670-010, ASME Code

Class 2, Field Weld FW-01, and -02)

- 21 -

Chemical and Valve BGV-0002, CVCS Letdown Manual Gas Tungsten Arc

Volume Control Orifice A Outlet Throttle Valve Piping Welding

System (Job 13005672-010, ASME Code

Class 2, Field Weld FW-04, and -05)

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 licensees 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, RF-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 licensees conclusion.

No findings were identified.

.3 Boric Acid Corrosion Control Inspection Activities

a. Inspection Scope

The inspectors reviewed the licensees 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 licensees 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

- 22 -

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 Code 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 DESCRIPTION CALLAWAY ACTION

NUMBER REQUEST

BBHV8002A and Reactor Head Vent Valve Tailpieces on Top 201406993

BHV8002B of the Reactor Head

EEJ01A Residual Heat Removal (RHR) System 201406827

Heat Exchanger A - Flange

EEJ01B Residual Heat Removal (RHR) System 201406528

Heat Exchanger B - Flange

BB10-C503 Hangar BB10-C503 (Adjacent Valve 201407170

BBHV8141C, RCP C SEAL # 1 SEAL WTR

OUT ISO HV Experienced Packing

Leakage)

EMHV8923A Refueling Water Storage Tank to Safety 201407454

Injection Pump A Suction Isolation Valve

EPV0124 Downstream Isolation Valve for Test Header 201407589

Valve EPHV8879D

EMV0179 Safety Injection Pump A from Residual Heat 201408130

Removal Heat Exchanger A Suction Vent

Valve

ENV0123 B Containment Spray Pump Casing and

Seal Housing Vent Valve

EJ8842 Residual Heat Removal Trains A&B Safety 201409218

Injection System Hot Leg Recirculation

Supply Header Pressure Relief Valve

BBHV8351A Reactor Coolant Pump A Seal Water Supply 201500874

Isolation Valve

BGFCV0110A Blending Tee Flow Control Valve and 201503867

BGPIS0141 Seal Water Injection Filter B

- 23 -

BGV0551 Chemical and Volume Control System Seal 201504450

Water Injection Filter B Outlet Drain Valve

(Bolted Blind Flange Assembly Downstream

of Valve)

EPHV8877B Safety Injection System Upstream Check 201505362

Test Line Isolation Valve

EMHV8923A Refueling Water Storage Tank to Safety 201600224

Injection Pump A Suction Isolation Valve

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 licensees 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 generators) was

performed.

- 24 -

The inspectors reviewed the licensees 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 generators:

o Prior to water lancing, a pre-look visual inspection was performed to

examine the sludge piles in two steam generators.

• 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)

- 25 -

• 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 action 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.

- 26 -

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 plants 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 constituted completion of one quarterly licensed operator performance

sample, as defined in Inspection Procedure 71111.11.

b. Findings

No findings were identified.

1R12 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 licensees corrective

actions. The inspectors reviewed the licensees work practices to evaluate whether

these may have played a role in the degradation of the structures, systems, and

components. The inspectors assessed the licensees 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.

- 27 -

1R13 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 licensees

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 constituted completion of five maintenance risk assessments and

emergent work control inspection samples, as defined in Inspection Procedure 71111.13.

b. Findings

No findings were identified.

1R15 Operability Determinations and Functionality Assessments (71111.15)

a. Inspection Scope

The inspectors reviewed six operability determinations and functionality assessments

that the licensee performed for degraded or nonconforming structures, systems, or

components:

- 28 -

• April 11, 2016, operability determination of safety related instrument bus inverters

• April 14, 2016, operability determination of leaks identified during train 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 licensees

evaluations. Where the licensee determined the degraded structures, systems, or

components to be operable or functional, the inspectors verified that the licensees

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 component.

These activities constituted 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

licensees 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

stations corrective action program. The licensee subsequently documented a prompt

operability determination for the essential service water system.

Inspectors subsequently reviewed the licensees 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,

- 29 -

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 licensees 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 licensees 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 licensees 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 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 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

- 30 -

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 licensees

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 licensees 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 licensees corrective action program as Callaway Action

Requests 201605488: NCV 05000483/2016002-03, Failure to Adequately Evaluate

Operability for a Degraded Condition.

1R18 Plant Modifications (71111.18)

Permanent Modifications

a. Inspection Scope

The inspectors reviewed three permanent plant modifications 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)

- 31 -

The inspectors reviewed the design and implementation of the modifications. The

inspectors verified that work activities involved in implementing the modifications 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 constituted completion of three samples of permanent modifications, as

defined in Inspection Procedure 71111.18.

b. Findings

No findings were identified.

1R19 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 the

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 components.

These activities constituted completion of five post-maintenance testing inspection

samples, as defined in Inspection Procedure 71111.19.

b. Findings

No findings were identified.

- 32 -

1R20 Refueling and Other Outage Activities (71111.20)

a. Inspection Scope

During the stations refueling outage that concluded on May 10, 2016, the inspectors

evaluated the licensees 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 licensees outage plan prior to the outage

• Review and verification of the licensees 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 constituted completion of one refueling outage sample, as defined in

Inspection Procedure 71111.20.

b. Findings

No findings were identified.

1R22 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 components 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 components

following testing.

These activities constituted completion of three surveillance testing inspection samples,

as defined in Inspection Procedure 71111.22.

- 33 -

b. Findings

No findings were identified.

2. RADIATION SAFETY

Cornerstones: Public Radiation Safety and Occupational Radiation Safety

2RS1 Radiological Hazard Assessment and Exposure Controls (71124.01)

a. Inspection Scope

The inspectors evaluated the licensees performance in assessing the radiological

hazards in the workplace associated with licensed activities. The inspectors assessed

the licensees 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 plants isotopic mix

and isotopic percent abundance, hard-to-detect radionuclides and potential alpha

hazards. The inspectors also reviewed the licensees 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 licensees controls for highly activated or contaminated materials

(non-fuel) stored within spent fuel and other storage pools and evaluated

airborne radioactive controls and monitoring.

- 34 -

• 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.

2RS3 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 licensees airborne monitoring

protocols, including verification that alarms and set points were appropriate.

• Use of respiratory protection devices and evaluation of the licensees 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

- 35 -

apparatus (SCBA) bottles, qualification and training of personnel, and user

performance.

• Self-contained breathing apparatus for emergency use including the licensees

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

4. OTHER ACTIVITIES

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency

Preparedness, Public Radiation Safety, Occupational Radiation Safety, and

Security

4OA1 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 index performance

indicator, as defined in Inspection Procedure 71151.

b. Findings

No findings were identified.

- 36 -

.2 Reactor Coolant System Identified Leakage (BI02)

a. Inspection Scope

The inspectors reviewed the licensees 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, Revision 7, to determine the

accuracy of the reported data.

- 37 -

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.

4OA2 Problem Identification and Resolution (71152)

.1 Routine Review

a. Inspection Scope

Throughout the inspection period, the inspectors performed daily reviews of items

entered into the licensees corrective action program and periodically attended the

licensees 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 licensees

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 201602658)

• Negative trend involving leaks on plant equipment as a result of train B

engineering safety feature actuation system testing (Callaway Action

Request 201603472)

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:

- 38 -

• 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

licensees 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 licensees 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 Callaways response to a non-cited violation that was issued in

Inspection Report 05000483/2010006 (ML103540576).

The inspectors assessed the licensees 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.

- 39 -

These activities constituted 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 licensees 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 licensees 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 licensees 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 licensees 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

- 40 -

• 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 Packages

10-0003 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 stations 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 licensees 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 issue 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

- 41 -

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 licensees 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 licensees 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.

4OA3 Follow-up of Events and Notices of Enforcement Discretion (71153)

(Closed) Licensee Event Report 2014-006-00, 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

dispositioned 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.

- 42 -

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 insulating 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 licensees

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 licensees 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 licensees 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 APA-ZZ-00801, Foreign Material Exclusion,

Revision 32.

Description. On December 3, 2014, an unexpected turbine and reactor trip occurred.

The licensees 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 licensees 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 preventive maintenance strategy was provided by the transformer

manufacturer nor identified by the licensees 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

- 43 -

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-00801, 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 licensees 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 licensees

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].

- 44 -

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 constituted completion of one event follow-up sample, as defined in Inspection

Procedure 71153.

4OA6 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 Officer, 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 19, 2016, the resident 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.

- 45 -

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 Physicist

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)05000483/2016002-03 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)

A1-1 Attachment 1

Closed

05000483/2014-006-00 LER Main Generator Excitation Transformer Faulted to Ground,

Causing Reactor Trip (Section 4OA3)

LIST OF DOCUMENTS REVIEWED

Section 1R01: Adverse Weather Protection

Procedures

Number Title Revision

AUE-ADM-2222 Communication and Coordination 0

AUE-ADM-2223 Disturbance Reporting 0

AUE-ADM-2227 Reliability Coordination - Responsibility and Authorities 0

OSP-NE-00001 Class 1E Electrical Source Verification 39

OSP-NE-00003 Technical Specification Actions - A.C. Sources 30

OTO-MA-00008 Rapid Load Reduction 34

OTO-ZZ-00012 Severe Weather 33

PDP-ZZ-00027 Seasonal Readiness Program 6

Callaway Action Requests

201508013 201604020

Jobs

13000681

Miscellaneous

Number Title Revision

2016 Summer Reliability Plan 3

2010009 Health Issue: Given an EDG HVAC equipment failure,

operability cannot be restored within the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allowed

outage time

2015005 Health Issue: Degradation of ESW Piping in Containment

A1-2

Section 1R04: Equipment Alignment

Procedures

Number Title Revision

OTN-AL-00001 Auxiliary Feedwater System 34

OTN-AL-00001, Auxiliary Feedwater Valve Alignment 22

Checklist 1

OTN-AL-00001 MD-AFP A and B Switch Alignment 18

Checklist 2

Drawings

Number Title Revision

E-012.2-00002 Large Induction Motors Outline 4

E-21010(Q) DC Main Single Line Diagram 14

LP-06 NB/NG/NK/NN-1, Safeguards Power Training Diagram 1

M-22AL01(Q) Auxiliary Feedwater System Piping and Instrumentation 46

Diagram

M-143A-00003 Concentric Restricting Orifice Plates Outline Drawing 19

Miscellaneous

Number Title Revision

GEK-72150 General Electric Instructions for Class 1E Auxiliary 0

Feedwater Pump Motors

Section 1R05: Fire Protection

Procedures

Number Title Revision

APA-ZZ-00703 Fire Protection Operability Criteria and Surveillance 26

Requirements

APA-ZZ-00750 Hazard Barrier Program 37

EDP-ZZ-04107 HVAC Pressure Boundary Control 29

OTO-KC-00001 Auxiliary Building 1974 - Boric Acid Tank Rooms 0

Add A-03

OTO-KC-00001 Auxiliary Building 2026 - North Electrical Pen Room 0

Add A-18

OTO-KC-00001 Control Building 2016 Switchboard and Battery Rooms 2 0

Add C-15 and 4

A1-3

Procedures

Number Title Revision

OTO-KC-00001 Control Building 2016 Switchboard and Battery Rooms 1 0

Add C-16 and 3

OSP-KC-00015 Fire Door Inspections 17

Drawings

Number Title Revision

A-2804 Architectural Fire Delineation Floor Plan, El 2047-6 27

Callaway Action Requests

201605406

Jobs

16003139

Miscellaneous

Number Title Revision

Fire Preplan Manual 38

KC-64 C-15 Detailed Fire Modeling Report 1

KC-65 C-16 Detailed Fire Modeling Report 1

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 0

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

A1-4

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 7

Sys. Reactor Building

EF01-C012/311 (Q) Pipe Supports - Essential Service Water Sys. Control 4

Bldg. - Trains A & B

EF02-C003/142 (Q) Pipe Supports - Essential Service Water Sys. Aux. 6

Bldg. A Train Supply

EF03-C034/134 (Q) Pipe Supports - Essential Service Water Sys. Aux. 6

Bldg. A Train Return

M-22EM01 (Q) Piping and Instrumentation Diagram High Pressure 36

Coolant Injection System

M-23EF01 Piping Isometric Essential Service Water System 25

Control Building

M-23EF02 Piping Isometric Essential Service Water System 33

Auxiliary Building A Train Supply

M-23EF03 Piping Isometric Essential Service Water System 33

Auxiliary Building A Train Return

M-23EF04 Piping Isometric Essential Service Water System 22

Auxiliary Building B Train Supply

M-23EF05 Piping Isometric Essential Service Water System 22

Auxiliary Building B Train Return

M-23EF06 Piping Isometric Essential Service Water System 26

Auxiliary Building A and Train Supply and Return

M-25BG23 (Q) Hanger Location Drawing - CVCS Charging & Excess 16

Letdown Reactor Building

A1-5

Drawings

Number Title Revision

M-25EF01 (Q) Hanger Location Drawing - Essential Service Water 14

Control Bldg. (A &B Train)

M-25EF02 (Q) Hanger Location Drawing - Essential Service Water 44

Sys. Aux. Bldg. A Train Supply

M-25EF03 (Q) Hanger Location Drawing - Essential Service Water 31

Sys. Aux. Bldg. A Train Return

Procedures

Number Title Revision

APA-ZZ-00350 Measuring and Test Equipment Program 29

APA-ZZ-00500 Corrective Action Program 63

APA-ZZ-00500, Operability and Functionality Determinations 25

Appendix 1

APA-ZZ-00500, Non-Conforming Materials Report 17

Appendix 2

APA-ZZ-00500, Past Operability and Reportability Evaluations 18

Appendix 3

APA-ZZ-00500, Transient Evaluation 2

Appendix 4

APA-ZZ-00500, Maintenance Rule 19

Appendix 5

APA-ZZ-00500, Collection and Preservation of Evidence 2

Appendix 6

APA-ZZ-00500, Effectiveness Reviews 10

Appendix 7

APA-ZZ-00500, Corrective Action Program Training Requirements 13

Appendix 8

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

Appendix 9

APA-ZZ-00500, Trending Program 11

Appendix 10

APA-ZZ-00500, Degraded And Nonconforming Condition Resolution 8

Appendix 11

APA-ZZ-00500, Significant Adverse Condition - Significance Level 1 24

Appendix 12

A1-6

Procedures

Number Title Revision

APA-ZZ-00500, Adverse Condition - Significance Level 2 25

Appendix 13

APA-ZZ-00500, Adverse Condition - Significance Level 3 23

Appendix 14

APA-ZZ-00500, Adverse Condition - Significance Level 4 20

Appendix 15

APA-ZZ-00500, Adverse Condition - Significance Level 5 13

Appendix 16

APA-ZZ-00500, Screening Process Guidelines 27

Appendix 17

APA-ZZ-00500, Equipment Performance Evaluation 8

Appendix 18

APA-ZZ-00500, Common Cause Evaluation (CCE) 5

Appendix 19

APA-ZZ-00500, Prompt Human Performance Evaluation (PHPE) 3

Appendix 20

APA-ZZ-00500, Other Issues 18

Appendix 21

APA-ZZ-00500, Corrective Action Program Definitions 13

Appendix 22

APA-ZZ-00661 Administration of Welding 16

APA-ZZ-00661, Personnel Approved to Perform Weld 3

Appendix 3 Inspections/Examinations

APA-ZZ-00662 ASME Section XI Repair/Replacement Program 22

APA-ZZ-00662, ASME Section XI Repair/Replacement Program 5

Appendix A Mandatory Requirements Class 1, 2 And 3 Items and

Their NF Supports (Fourth Inspection Interval)

APA-ZZ-00662 ASME Section XI Code Cases Applied to the Fourth 6

Appendix B Inspection Interval

APA-ZZ-00662 ASME Section XI Repair/Replacement Matrix Minor 4

Appendix E

APA-ZZ-00662 ASME Section XI Repair/Replacement Program 0

Appendix G Mandatory Requirements Class MC and CC Items

and their NF Supports (Second Inspection Interval)

APA-ZZ-00750 Hazard Barrier Program 37

EDP-ZZ-00018 Heat Exchanger Eddy Current Testing Methodology 3

A1-7

Procedures

Number Title Revision

EDP-ZZ-01004 Boric Acid Corrosion Control Program 17

EDP-ZZ-01121 Raw Water Systems Predictive Performance 21

Program

ESP-ZZ-01016 ASME Section XI IWE Containment Pressure 6

Boundary Inspection

MDP-ZZ-LM001 Fluid Leak Management Program 15

MSM-ZZ-QW005 Mechanical Snubber Functional Test 17

MTW-ZZ-WP001 ASME/ANSI General Welding Requirements 26

MTW-ZZ-WP002 Welder Performance Qualification 27

MTW-ZZ-WP003 Control Of Welding Filler Materials 24

MTW-ZZ-WP004 Post Weld Heat Treatment 11

MTW-ZZ-WP006 Qualification of Welding Procedures 9

MTW-ZZ-WP007 Callaway Plant Maintenance Welding Procedure 4

AWS D1.1 General Welding Requirements

MTW-ZZ-WP501 Callaway Plant Maintenance Welding Procedure 14

Welding of P-1 Materials

MTW-ZZ-WP502 Callaway Plan Maintenance Welding Procedure 10

Welding of P-1 to P-3 Materials

MTW-ZZ-WP503 Callaway Plan Maintenance Welding Procedure 8

Welding of P-1 to P-4 Materials

MTW-ZZ-WP504 Callaway Plan Maintenance Welding Procedure 10

Welding of P-1 to P-5 Materials

MTW-ZZ-WP505 Callaway Plan Maintenance Welding Procedure 10

Welding of P-1 to P-8 Materials

MTW-ZZ-WP506 Callaway Plan Maintenance Welding Procedure 8

Welding of P-4X (Including Welding of P-1 and P-8 to

P-4X) Materials

MTW-ZZ-WP509 Callaway Plan Maintenance Welding Procedure 8

Welding of P-3 Materials

MTW-ZZ-WP510 Callaway Plan Maintenance Welding Procedure 9

Welding of P-4 Materials

MTW-ZZ-WP511 Callaway Plan Maintenance Welding Procedure 10

Welding of P-5 Materials

MTW-ZZ-WP512 Callaway Plan Maintenance Welding Procedure 5

Welding of P-5 to P-8 Materials

A1-8

Procedures

Number Title Revision

MTW-ZZ-WP513 Callaway Plan Maintenance Welding Procedure 4

Welding of P-6 to P-8 Materials

MTW-ZZ-WP514 Callaway Plan Maintenance Welding Procedure 16

Welding of P-8 Materials

MTW-ZZ-WP524 Callaway Plan Mechanical Technical Procedure 8

Torch Brazing of Copper Alloys

MTW-ZZ-WP525 Callaway Plan Maintenance Welding Procedure 4

Welding of P-4 to P-8 Materials

MTW-ZZ-WP526 Callaway Plan Maintenance Welding Procedure 3

Welding of P-8 to P-34 Materials

MTW-ZZ-WP527 Callaway Plan Maintenance Welding Procedure 3

Welding of P-34 Materials

MTW-ZZ-WP560 Callaway Plan Maintenance Welding Procedure 9

Fusing of High Density Polyethylene (HDPE)

Materials for Nuclear Service

MTW-ZZ-WP561 Callaway Plan Maintenance Welding Procedure 5

Fusing of High Density Polyethylene (HDPE)

Materials for Non-Nuclear Service

MTW-ZZ-WP701 AWS Welding of P-1 Materials 3

MTW-ZZ-WP702 Callaway Plant Maintenance Technical Procedure 2

AWS Welding of Studs

PDI-ISI-254-SE Remote Inservice Examination of Reactor Vessel 2

Nozzle to Safe End, Nozzle to Pipe and Safe End to

Pipe Welds

PDI-ISI-254-SE-NB Remote Inservice Examination of Reactor Vessel 0

Nozzle to Safe End, Nozzle to Pipe and Safe End to

Pipe Welds Using the Nozzle Scanner

QCP-ZZ-05000 Liquid Penetrant Examination 25

QCP-ZZ-05010 Magnetic Particle Examination 19

QCP-ZZ-05019 Ultrasonic Thickness Measurement 14

QCP-ZZ-05030 Radiographic Procedure for Examination of 17

Weldments and Castings

QCP-ZZ-05041 Visual Examination to ASME VT-2 26

QCP-ZZ-05048 Boric Acid Walkdown for Reactor Coolant System 8

Pressure Boundary

QCP-ZZ-05049 Reactor Pressure Vessel Head Bare Metal 3

Examination

A1-9

Procedures

Number Title Revision

UT-2 Ultrasonic Examination of Vessel Welds and 30

Adjacent Base Metal

UT-94 Ultrasonic Examination of Ferritic Piping Welds 9

UT-95 Ultrasonic Examination of Austenitic Piping Welds 8

UT-96 Ultrasonic Through Wall Sizing in Piping Welds 7

UT-103 Ultrasonic Examination of Dissimilar Metal Piping 5

Welds

WDI-SSP-1101 Manual Ultrasonic Examination of Reactor Vessel 1

Threads in Flange for Callaway Unit 1

WDI-STD-088 Underwater Remote Visual Examination of Reactor 9

Vessel Internals

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

Surface

Relief Requests

Number Title Date

Letter: Michael T. Callaway Plant, Unit 1 - Request for Relief 14R-01, May 12, 2015

Markley to Fadi Alternative to ASME Code Inservice Inspection

Diya Requirements for Class 3 Buried Piping

(TAC NO. MF4271)

ULNRC-06115 NRC Letter, "Relief Request 13R-10 for Third 10-Year June 10, 2014

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. ML083100288)

ULNRC-06146 Ameren Missouri Letter ULNRC-06115, "10 CFR 50.55a September 30,

Request: Proposed Alternative to ASME Section XI 2014

Requirements for Class 3 Buried Piping," dated

June 10, 2014 (ADAMS Accession No. ML14161A399)

UNNRC-06214 Docket Number 50-483 Callaway Plant Unit 1 Union April 24, 2015

Electric Co. Facility Operating License NPF-30 Revision

of 10 CFR 50.55a Request: Proposed Alternative to

ASME Section XI Requirements for Class 3 Buried

Piping (TAC NO. MF4271)

Work Packages

15000069-520 15507345 16001742-405 16503498

15000069-505 15507967 16001742-405 16503745

A1-10

Work Packages

15001243-500 16001742-550 16001743-400

Jobs

10002667 16001870

Miscellaneous

Number Title Revision/Date

Various Non Destructive Examination Reports for

ESW components

206EZ-FLO Garlock Sealing Technologies Expansion Joint November 15, 2006

Test

0516-19-F01 Secondary Side Visual Inspection Plan for February 10, 2016

Ameren UE, Callaway RF 21

51-9252420-000 AREVA Engineering Information Record: March 21, 2016

Callaway 1RF021 SG ECT Inspection Plan

51-9253319-000 AREVA Engineering Information Record: April 2016

Callaway 1R21 Degradation Assessment

96225-TR-002 Containment F Cooler Response to a 1

Simultaneous LOCA & LOOP Event

0096-020-CALC-01 Callaway Water Hammer Load Calculation 0

A190.0002 Procedure Review Form UT-2 Ultrasonic October 8, 2014

Examination of Vessel Welds and Adjacent Base

Metal, Revision 30

A190.0002 Procedure Review Form UT-94 Ultrasonic October 8, 2014

Examination of Ferritic Piping Welds, Revision 9

A190.0002 Procedure Review Form UT-95 Ultrasonic October 8, 2014

Examination of Austenitic Piping Welds,

Revision 8

A190.0002 Procedure Review Form UT-96 Ultrasonic October 8, 2014

Through Wall Sizing in Piping Welds, Revision 7

A190.0002 Procedure Review Form UT-103 Ultrasonic October 8, 2014

Examination of Dissimilar Metal Piping Welds,

Revision 5

AP14-008 Self-Assessment: Nuclear Oversight ISI - IST October 8, 2014

Audit

EDP-ZZ-00016 Self-Assessment: Checklist for Program Review October 8, 2014

of Alloy 600 Program

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

A1-11

Miscellaneous

Number Title Revision/Date

RIS 2016-02 NRC Regulatory Issue Summary 2016-02, March 23, 2016

OMB Control Design Basis Issues Related to Tube-To-

No. 3150-0011 Tubesheet Joints in Pressurized-Water Reactor

Steam Generators. (ML15169A543)

T65.0212 6 Callaway Fall Protection February 14, 2014

Section 1R11: Licensed Operator Requalification Program

Procedures

Number Title Revision

ODP-ZZ-00001 Operations Department - Code of Conduct 97

OSP-AC-00005 Turbine Actual Overspeed Trip 11

OTG-ZZ-00005 Plant Shutdown 20% Power to Hot Standby 47

Callaway Action Requests

200601332 201600670

Miscellaneous

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 24

EDP-ZZ-01128, SSCs in Scope of the Maintenance Rule at Callaway 10

Appendix 1

EDP-ZZ-01128, Maintenance Rule System Functions 16

Appendix 4

A1-12

Callaway Action Requests

201602435 201602658 201602738 201602824 201603229

201603471 201603472 201603473 201603484

Jobs

11504345 16001349

Miscellaneous

Number Title Revision/Date

Procon1, LLC Evaluation of Room Cooler SGL-10A April 13, 2016

Tube Leak Repair

1784 Union Electric Company Laboratory Services - September 22, 1994

Metallurgical Report - Examination of Failed Room

Cooler Tubing

04060221 AmerenUE Technical Support Services - Metallurgical September 30, 2004

Report - Examination of Callaway Room Cooler Tubes

13050249 Ameren Missouri Technical Support - Metallurgical May 23, 2013

Report - Examination of Callaway Room Cooler Tubing

GL-137 SGL10A/B Room Cooler Heat Removal Capabilities 0

Section 1R13: Maintenance Risk Assessment and Emergent Work Controls

Procedures

Number Title Revision

APA-ZZ-00315 Configuration Risk Management Program 14

ODP-ZZ-00002, Protected Equipment Program 23

Appendix 1

ODP-ZZ-00002, Placing Train A Protected Equipment Barriers, Mode 5 & 6 2

Appendix 1,

Checklist 5

ODP-ZZ-00002, Placing Train B Protected Equipment Barriers, Mode 5 & 6 2

Appendix 1,

Checklist 7

A1-13

Procedures

Number Title Revision

ODP-ZZ-00002, Placing Train A Protected Equipment Barriers, Defueled 2

Appendix 1,

Checklist 9

ODP-ZZ-00002, Placing Protected Equipment Barriers for SFP Cooling 1

Appendix 1, Outage

Checklist 17

ODP-ZZ-00002, Risk Management Actions for Planned Risk Significant 11

Appendix 2 Activities

ODP-ZZ-00002, Postings for Lowered Inventory Operations 2

Appendix 2,

Checklist 9

Callaway Action Requests

201601830 201602875 201603382 201605725 201605766

Jobs

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

Miscellaneous

Number Title Revision

Shutdown Safety Management Plan 3

PRAER 16-405 PRA Evaluation Request - Mode Change from Mode 4 to 0

Mode 3 with Equipment OOS

Section 1R15: Operability Evaluations

Procedures

Number Title Revision

KDP-ZZ-00013 Emergency Response Facility and Equipment Evaluation 13

MTE-ZZ-QA013 MOVATS UDS Testing of Torque Controlled Limitorque 19

Motor Operated Rising Stem Valves

A1-14

Procedures

Number Title Revision

ODP-ZZ-00002 Equipment Status Control 83

OSP-EJ-V002A RHR Pump Containment Sump Suction and RWST Suction 31

Inservice Test

Drawings

Number Title Revision

8600-X-89645 High Pressure & Low Pressure Nitrogen Gas Storage & 15

Transfer System Site Gas Systems (KH2) Piping and

Instrumentation Diagram

E-23BB12A(Q) RHR Loop 1 Inlet Isolation Valve Schematic Diagram 12

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

Alarms

E-1038-00003 Schematic 7.5kVA Inverter 120VAC, 1ø, 60Hz 2

E-1038-00006, Outline 7.5kVA Inverter Front Panel Identification 2

S002

M-22AB02(Q) Main Steam System Piping and Instrumentation Diagram 17

M-22FA01 Auxiliary Boiler System Piping and Instrumentation Diagram 18

M-22KH01 Service Gas System Piping and Instrumentation Diagram 29

M-622.1-00023 Condensing Unit 19

E-23KJ08A(Q) Standby Jacket Coolant Heater EKJ01A Schematic Diagram 2

E-23KJ09B(Q) Standby Jacket Coolant Circ. Pump PKJ01A Schematic 2

Diagram

M-22KJ01(Q) Standby Diesel Generator A Cooling Water System Piping 24

and Instrumentation Diagram

Callaway Action Requests

201603312 201603353 201603598 201603711 201603739

201603758 201604998 201605016 201605045 201605324

201605917 201105227

Jobs

10507721 10507762 13505626 14511766 16001888

16002253 16002356 16003607

A1-15

Miscellaneous

Number Title Revision

BO-05 Revised Temperatures for 3601, 3605, and 3609 for Station 1

Addendum 19 Black Out

BO-07 Control Room SBO Heat Load Calculation 11

EF-123 UHS Thermal Performance Analysis using GOTHIC 7.2(b) 1

CAR#201001813

RFR 17478 Perform Evaluation for NRC GL96-06 Response C

RFR 201603756 Request for Resolution: Modify low pressure nitrogen 0

system piping and penetrations

Section 1R18: Plant Modifications

Procedures

Number Title Revision

APA-ZZ-00600 Design Change Control 57

EDP-ZZ-04015 Evaluating and Processing Requests for Resolution (RFR) 70

Drawings

Number Title Revision

M-22AL01(Q) Auxiliary Feedwater System Piping and Instrumentation 46

Diagram

M-22AN01 Demineralized Water Storage and Transfer System Piping 42

and Instrumentation Diagram

M-22AP01 Condensate Storage and Transfer System Piping and 31

Instrumentation Diagram

M-22AP02 Hardened Condensate Storage Tank Composite Piping and 0

Instrumentation Diagram

M-22AQ02 Feedwater Chemical Addition System Piping and 17

Instrumentation Diagram

M-22KA09 Instrument Air System Piping and Instrumentation Diagram 25

Miscellaneous

Number Title Revision/Date

50.59 Screen for MP 13-0033 Hardened Condensate 4

Storage Tank Refuel 21 Tie-Ins

Applicability Determination for MP 13-0033 Hardened 4

Condensate Storage Tank Refuel 21 Tie-Ins

A1-16

Miscellaneous

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 4

Storage Tank Refuel 21 Tie-Ins

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 33

APA-ZZ-00320 Work Execution 56

APA-ZZ-00322 Job Planning 43

Appendix C

MTE-ZZ-QA013 MOVATS UDS Testing of Torque Controlled Limitorque 19

Motor Operated Rising Stem Valves

OSP-JE-00001 Emergency Fuel Oil Transfer Pumps Cross-connection Line 13

Fill Verification Test

OSP-NE-0001A Standby Diesel Generator A Periodic Tests 62

OTN-NB-0001A NB01 transfer to XNB02 Single Offsite Source Operation 8

Addendum 3 and Restoration

OTN-NE-0001A Standby Diesel Generation System -Train A 48

Drawings

Number Title Revision

E-23BB12A(Q) RHR Loop 1 Inlet Isolation Valve Schematic Diagram 12

M22-KH01 Service Gas System Piping and Instrumentation Diagram 29

Callaway Action Requests

201602435 201603496 201603598 201603758 201604092

201605141 201605393

Jobs

10507721 10507762 16001888 16001887 16001349

14005657 15505373 13505566 14511620 16002253

A1-17

Jobs

16003027

Section 1R20: Refueling and Other Outage Activities

Procedures

Number Title Revision

APA-ZZ-00908 Fitness for Duty Programs 34

APA-ZZ-00911 Fatigue Management 5

ESP-ZZ-00024 Low Power Physics Testing Data Acquisition 9

OSP-SA-00004 Visual Inspection of Containment for Loose Debris 25

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 47

OTG-ZZ-00006 Plant Cooldown Hot Standby to Cold Shutdown 74

OTG-ZZ-00007 Refueling Preparation, Performance and Recovery 38

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 29

OSP-BN-V0005 BN Suction Header Valves Inservice Test 5

OSP-EJ-0006A RHR Mini Flow Valve Time Response Test Train A 2

OSP-EJ-0006B RHR Mini Flow Valve Time Response Test Train B 2

OSP-EJ-PV04A Train A RHR and RCS Check Valve Inservice Test 10

OSP-EJ-PV04B Train B RHR and RCS Check Valve Inservice Test 12

OSP-EJ-V002B RWST to RHR Suction Check Valve Inservice Test 10

A1-18

Procedures

Number Title Revision

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

Test

OSP-EM-V0003 ECCS Check Valve Inservice Test 33

OSP-EM-V003A CCP A and B Full Flow Test 24

OSP-EM-V0004 RHR Check Valve and SI Pump Recirc Valve Inservice Test 22

OSP-EM-V0005 EM8922A and EM8922B Closure Inservice Test 11

OSP-EP-V0006 SI Accumulator Discharge Check Valve Test 9

OSP-NE-0001B Standby Diesel Generator B Periodic Tests 64

OSP-SA-2413B Train B Diesel Generator and Sequencer Testing 26

OTN-NE-0001B Standby Diesel Generation System - Train B 51

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

Callaway Action Requests

201604838 201508227 201503020

Jobs

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 22

APA-ZZ-01000 Callaway Energy Center Radiation Protection Program 41

APA-ZZ-01004 Radiological Work Standards 27

HDP-ZZ-01200 Radiation Work Permits 29

HDP-ZZ-01500 Radiological Postings 44

HDP-ZZ-03000 Radiological Survey Program 43

HDP-ZZ-03000

Frequency and Location of Routine Radiological Surveys 13

APPA

HTP-ZZ-02004 Control of Radioactive Sources 39

A1-19

Procedures

Number Title Revision

High Radiation / Locked High Radiation / Very High

HTP-ZZ-06001 50

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 0

14006281 BB8948D Maintenance, Disassemble, Inspect, Repair 1

leak-by and Reassemble Check Valve BB8948D

14006280 BB8949D Disassembly and Repair, Remove/Reinstall 1

Insulation, 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 1

Tasks

15001126500 Replace BBV0400 0

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 August 21,

HRA 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

A1-20

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 April 12, 2016

new BGV-001 valve and piping

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 23

HDP-ZZ-08002 Respiratory Protection Issue and Use 42

HTP-ZZ-08203-DTI- Testing Scott Regulators And Respirators Using The 8

REGULATORS Biosystems Posichek3 Tester

A1-21

Procedures

Number Title Revision

HTP-ZZ-08208-DTI- Quantitative Respirator Fit Testing Using The Tsi 2

FITPRO-TESTING Portacount Pro System

HTP-ZZ-08208-DTI- Quantitative Respirator Fit Testing Using The Tsi 6

FIT-TESTING Portacount Plus System

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

AIRPAK75 Storage

HTP-ZZ-08300-DTI- Post Hydrostatic Testing of Breathing Air Cylinders 4

POST HYDRO

HTP-ZZ-08300-DTI- SKA-PAK at SCBA Respirator Storage and Inspection 8

SKAPAK

HTP-ZZ-08301-DTI- Manual Cleaning of Respiratory Protection Equipment 1

RESPRO CLEAN

HTP-ZZ-08301-DTI- Manual Cleaning of Scott Mask Mounted Regulator 4

SCOTT-RES-CLEAN

HTP-ZZ-08501-DTI- Testing of Breathing Air 5

AIR TEST

HTP-ZZ-08502-DTI- Scott Mobile Air Cart Calibration 3

MAC-CAL

HTP-ZZ-08503-DTI- Operation of Bauer UNICUS III, 25 CFM Breathing Air 4

UNIIICOMPRESSOR Compressor and Breathing Air Cascade System

RP-DTI-RESPRO- Storage of Respirators 3

STORAGE

Callaway Action Requests

201407682 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

A1-22

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 September 20, 2016

Overhaul

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

Overhaul

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 9

OSP-BB-00009 RCS Inventory Balance 37

Callaway Action Requests

201502229 201505332 201505796

Jobs

16503927

Miscellaneous

Number Title Revision Date

Mitigating Systems Performance Index (MSPI) Basis 16

Document

A1-23

Miscellaneous

Number Title Revision Date

NRC Performance Indicator Transmittal Report, Second July 9, 2015

Quarter 2015, Mitigating Systems Cornerstone

NRC Performance Indicator Transmittal Report, Third October 12,

Quarter 2015, Mitigating Systems Cornerstone 2015

NRC Performance Indicator Transmittal Report, Fourth January 11,

Quarter 2015, Mitigating Systems Cornerstone 2016

NRC Performance Indicator Transmittal Report, First April 13, 2016

Quarter 2016, Mitigating Systems Cornerstone

MSPI Derivation Report, MSPI Heat Removal System, June 2015

Unavailability Index (UAI)

MSPI Derivation Report, MSPI Heat Removal System, June 2015

Unreliability Index (URI)

MSPI Derivation Report, MSPI Heat Removal System, September

Unavailability Index (UAI) 2015

MSPI Derivation Report, MSPI Heat Removal System, September

Unreliability Index (URI) 2015

MSPI Derivation Report, MSPI Heat Removal System, December 2015

Unavailability Index (UAI)

MSPI Derivation Report, MSPI Heat Removal System, December 2015

Unreliability Index (URI)

MSPI Derivation Report, MSPI Heat Removal System, March 2015

Unavailability Index (UAI)

MSPI Derivation Report, MSPI Heat Removal System, March 2015

Unreliability Index (URI)

Reactor Coolant System Identified Leakage Data April 1, 2015

through

March 30, 2016

NRC Performance Indicator Transmittal Report, Second July 6, 2015

Quarter 2015, Barrier Integrity Cornerstone

NRC Performance Indicator Transmittal Report, Third October 12,

Quarter 2015, Barrier Integrity Cornerstone 2015

NRC Performance Indicator Transmittal Report, Fourth January 11,

Quarter 2015, Barrier Integrity Cornerstone 2016

NRC Performance Indicator Transmittal Report, First April 8, 2016

Quarter 2016, Barrier Integrity Cornerstone

LER 2015-001-00 Licensee Event Report - Completion of a Shutdown 0

Required by the Technical Specifications

A1-24

Miscellaneous

Number Title Revision Date

LER 2015-002-00 Licensee Event Report - Manual Auxiliary Feedwater 0

Actuation

LER 2015-003-00 Licensee Event Report - Reactor Trip Caused by 0

Transmission Line Fault

LER 2015-003-01 Licensee Event Report - Reactor Trip Caused by 1

Transmission Line Fault

LER 2015-004-00 Licensee Event Report - Auxiliary Feedwater Flow 0

Control Valve Inoperable due to Faulty Electronic

Positioner Card

Section 4OA2: Identification and Resolution of Problems

Procedures

Number Title Revision

APA-ZZ-00500, Corrective Action Program Training Requirements 13

Appendix 8

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

Appendix 9

APA-ZZ-00500, Trending Program 11

Appendix 10

APA-ZZ-00500, Degraded And Nonconforming Condition Resolution 8

Appendix 11

APA-ZZ-00500, Significant Adverse Condition - Significance Level 1 24

Appendix 12

APA-ZZ-00500, Adverse Condition - Significance Level 2 25

Appendix 13

APA-ZZ-00500, Adverse Condition - Significance Level 3 23

Appendix 14

APA-ZZ-00500, Adverse Condition - Significance Level 4 20

Appendix 15

APA-ZZ-00500, Adverse Condition - Significance Level 5 13

Appendix 16

APA-ZZ-00500, Screening Process Guidelines 27

Appendix 17

APA-ZZ-00500, Equipment Performance Evaluation 8

Appendix 18

A1-25

Procedures

Number Title Revision

APA-ZZ-00500, Common Cause Evaluation (CCE) 5

Appendix 19

APA-ZZ-00500, Corrective Action Program Definitions 13

Appendix 22

APA-ZZ-00600 Design Change Control 57

Drawings

Number Title Revision

M-22AE01 Piping and Instrumentation Diagram Service Water System 22

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

Miscellaneous

Number Title Revision

MP 10-0003 Install Service Water Check Valves to Minimize ESW Water 1

Hammer During LOOP and ESFAS Testing

MP 10-0004 Revise Sequencer Operation of EFHV0037 and EFHV0038 2

Section 4OA3: Event Follow-Up

Procedures

Number Title Revision

APA-ZZ-00500 Corrective Action Program 57

A1-26

Procedures

Number Title Revision

APA-ZZ-00801 Foreign Material Exclusion 32

Callaway Action Requests

200603505 201408897 201606129

Jobs

11509869 13004764

Miscellaneous

Number Title Revision

E-1051-00104 IM for Dry Type Transformer Installation 0

A1-27

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/folder 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.

A2-1 Attachment 2

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

A2-2

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.

A2-3