ML16225A577
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
SUNSI Review ADAMS Non- Publicly Available Keyword:
By: DLP Yes No Sensitive Non-Publicly Available NRC-002
Sensitive
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
-5-
available and adequately maintained to respond during a loss of off-site power event [H.1].
(Section 4OA2.3)
-6-
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
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
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
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