ML17283A392

NRC Design Bases Assurance Inspection Report 05000483/2017007 and Notice of Violation

ML17283A392
Person / Time
Site:	Callaway
Issue date:	10/06/2017
From:	Thomas Farnholtz Region 4 Engineering Branch 1
To:	Diya F Ameren Missouri
References
IR 2017007
Download: ML17283A392 (43)

See also: IR 05000483/2017007

Text

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION IV

1600 E. LAMAR BLVD

ARLINGTON, TX 76011-4511

October 6, 2017

Mr. Fadi Diya, Senior Vice President

and Chief Nuclear Officer

Ameren Missouri

Callaway Plant

P.O. Box 620

Fulton, MO 65251

SUBJECT: CALLAWAY PLANT - NRC DESIGN BASES ASSURANCE INSPECTION

REPORT 05000483/2017007 AND NOTICE OF VIOLATION

Dear Mr. Diya:

On August 28, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection

at your Callaway Plant. On August 4, 2017, the NRC team discussed the preliminary results of

this inspection with Mr. T. Herrmann, Site Vice President, and other members of your staff. On

August 28, 2017, the NRC team discussed the final results of this inspection with

Ms. S. Kovaleski, Director, Design Engineering, and other members of your staff. The team

documented the results of this inspection in the enclosed inspection report.

Based on the results of this inspection, the NRC has identified three issues that were evaluated

under the significance determination process as having very low safety significance (Green).

The NRC has also determined that three violations are associated with these issues. The NRC

is treating two of these violations as non-cited violations (NCVs) in accordance Section 2.3.2.a

of the NRC Enforcement Policy. One violation is cited in the enclosed Notice of Violation

(Notice) and the circumstances surrounding it are described in detail in the subject inspection

report. The violation is being cited because the licensee failed to restore compliance, in a

reasonable time, for not implementing procedures for performing maintenance that can affect

the performance of safety-related equipment. The NRC previously identified this violation as

NCV 05000483/2014007-01.

Further, the team documented a licensee-identified violation which was determined to be of very

low safety significance in this report. The NRC is treating this violation as an NCV consistent

with Section 2.3.2.a of the NRC Enforcement Policy.

You are required to respond to this letter and should follow the instructions specified in the

enclosed 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 compliance with regulatory requirements.

F. Diya 2

If you contest the violations or significance of these NCVs, 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; the Director, Office of Enforcement; and the

NRC resident inspector at the Callaway Plant.

If you disagree with a cross-cutting aspect assignment 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 U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk,

Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; and the

NRC resident inspector at the Callaway Plant.

This letter, its enclosure, and your response (if any) will be made available for public inspection

and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document

Room in accordance with 10 CFR 2.390, Public Inspections, Exemptions, Requests for

Withholding.

Sincerely,

/RA/

Thomas R. Farnholtz, Chief

Engineering Branch 1

Division of Reactor Safety

Docket No. 50-483

License No. NPF-30

Enclosure:

1. Notice of Violation

2. Inspection Report 05000483/2017007

w/Attachment: Supplemental Information

cc w/ enclosure: Electronic Distribution

NOTICE OF VIOLATION

Union Electric Company Docket No. 50-483

Callaway Plant License No. NPF-30

During an NRC inspection conducted July 17 through August 28, 2017, a violation of an NRC

requirement was identified. In accordance with the NRC Enforcement Policy, the violation is

listed below:

Technical Specification 5.4.1.a requires, in part, that written procedures shall be

established, implemented, and maintained covering the applicable procedures

recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.

Regulatory Guide 1.33, Appendix A, Section 9, Procedures for Performing

Maintenance, requires, in part, that maintenance that can affect the performance of

safety-related equipment should be properly pre-planned and performed in accordance

with documented instructions appropriate to the circumstances.

Contrary to the above, from May 2014 through August 4, 2017, the licensee failed to

ensure that maintenance that can affect the performance of safety-related equipment be

properly pre-planned and performed in accordance with documented instructions

appropriate to the circumstances. Specifically, as a result of ineffective corrective action

of Callaway Action Requests CAR-201402827 and CAR-201405312, the licensee failed

to perform preventative maintenance procedures to verify the operation and timing of the

engineered safety feature transformer XNB01 load tap changer.

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

Pursuant to the provisions of 10 CFR 2.201, Union Electric Company is required to submit a

written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document

Control Desk, Washington, DC 20555, with a copy to the Regional Administrator, Region IV,

and a copy to the NRC Resident Inspector at the facility that is the subject of this Notice of

Violation (Notice), within 30 days of the date of the letter transmitting this Notice. This reply

should be clearly marked as Reply to Notice of Violation 05000483/2017007-01, and should

include for the violation (1) the reason for the violation, or if contested, the basis for disputing

the violation or the severity level, (2) the corrective steps that have been taken and the results

achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date

when full compliance will be achieved. Your response may reference or include previous

docketed correspondence, if the correspondence adequately addresses the required response.

If an adequate reply is not received within the time specified in this Notice, an Order or a

Demand for Information may be issued as to why the license should not be modified,

suspended, or revoked, or why such other action as may be proper should not be taken. Where

good cause is shown, consideration will be given to extending the response time.

If you contest this enforcement action, you should also provide a copy of your response, with

the basis for your denial, to the Director, Office of Enforcement, United States Nuclear

Regulatory Commission, Washington, DC 20555-0001.

Because your response will be made available electronically for public inspection in the NRC

Public Document Room, or from the NRC's document system (ADAMS), accessible from the

NRC Web site at http://www.nrc.gov/reading-rm/adams.html, to the extent possible, it should not

Enclosure 1

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 deletes such information. If you request withholding of such material, you must to

in detail the basis of 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 withholding confidential commercial or financial

information.

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

days of receipt.

Dated this 6th day of October 2017.

2

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket: 05000483

License: NPF-30

Report Nos.: 05000483/2017007

Licensee: Union Electric Company

Facility: Callaway Plant

Location: Junction Highway CC and Highway O, Fulton, Missouri

Dates: July 17 through August 28, 2017

Team Leader: R. Kopriva, Senior Reactor Inspector, Engineering Branch 1

Inspectors: I. Anchondo, Reactor Inspector, Engineering Branch 2

J. Watkins, Reactor Inspector, Engineering Branch 2

A. Palmer, Senior Reactor Technology Instructor, Technical Training

Center

Accompanying C. Baron, Contractor, Beckman and Associates

Personnel: J. Nicely, Contractor, Beckman and Associates

Approved By: Thomas R. Farnholtz

Branch Chief, Engineering Branch 1

Division of Reactor Safety

Enclosure 2

SUMMARY

IR 05000483/2017007; 07/17/2017 - 08/28/2017; Callaway Plant; Baseline Inspection, NRC

Inspection Procedure 71111.21M, Design Basis Assurance Inspection.

The report covers an announced inspection by a team of three regional inspectors,

two contractors, and one operations inspector from the NRC training facility. Four findings were

identified. One of these findings was a licensee-identified violation. One of the three

NRC-identified violations is being treated as a cited violation because the licensee had failed to

restore compliance and it was a repeat of a previous NRC identified violation. The findings

were of very low safety significance. The final significance of most findings is indicated by their

color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance

Determination Process. Cross-cutting aspects were determined using Inspection Manual

Chapter 0310, Aspects Within the Cross-Cutting Areas. Findings for which the Significance

Determination Process does not apply may be Green or be assigned a severity level after NRC

management review. All violations of NRC requirements are dispositioned in accordance with

the NRCs Enforcement Policy, dated July 9, 2013. The NRCs program for overseeing the safe

operation of commercial nuclear power reactors is described in NUREG-1649, Reactor

Oversight Process, Revision 6, dated July 2016.

Cornerstone: Mitigating Systems

• Green. The team identified a Green, cited violation of Technical Specification 5.4.1.a

which requires, in part, that written procedures shall be established, implemented, and

maintained covering the applicable procedures recommended in Regulatory Guide 1.33,

Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Appendix A, Section 9,

Procedures for Performing Maintenance, requires, in part, that maintenance that can

affect the performance of safety-related equipment should be properly pre-planned

and performed in accordance with documented instructions appropriate to the

circumstances. Specifically, from May 2014 through August 4, 2017, as a result of

ineffective corrective action of Callaway Action Requests CAR-201402827 and

CAR-201405312, the licensee failed to performed preventative maintenance procedures

to verify the operation and timing of the engineered safety feature transformer XNB01

load tap changer. This violation was previously identified by the NRC and documented

as NCV 05000483/2014007-01. In accordance with Section 2.3.2.a of the NRC

Enforcement Policy, this finding is being cited because the licensee failed to restore

compliance within a reasonable amount of time after the violation was initially identified.

This finding was entered into the licensees corrective action program as Condition

Report CR-201703992, VIO 05000458/2017007-01, Not Verifying the Operation and

Timing of the Engineered Safety Feature Transformer XNB01 Load Tap Changer.

The team determined that the failure to implement maintenance procedures to

periodically verify transformer XNB01 load tap changer operation and time testing was a

performance deficiency. The performance deficiency was more than minor, and

therefore a finding, because it was associated with the equipment performance attribute

of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective

to ensure the availability, reliability, and capability of systems that respond to initiating

events to prevent undesirable consequences. Specifically, the failures to perform,

periodic verification of the operation and time testing of the load tap changer could result

in adverse operation of the load tap changer during a design basis event. If the load tap

changer did not operate correctly, the safety-related buses may not have adequate

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voltage to reset the degraded voltage relay, thus spuriously disconnecting from the

offsite power source. In accordance with Inspection Manual Chapter 0609, Appendix A,

Significance Determination Process (SDP) for Findings At-Power, dated June 19,

2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as

having very low safety significance (Green) because it was a design or qualification

deficiency that did not represent a loss of operability or functionality; did not represent an

actual loss of safety function of the system or train; did not result in the loss of one or

more trains of non-technical specification equipment; and did not screen as potentially

risk-significant due to seismic, flooding, or severe weather. The finding had a

cross-cutting aspect in the area of human performance, work management, because the

licensee failed to plan, control, and execute work activities such that nuclear safety is the

overriding priority. Specifically, the licensee did not plan and execute the testing of the

transformer XNB01 load tap changer in a timely manner (H.5). (Section 1R21.2.4)

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

Criterion XI, Test Control, which requires, in part, that a test program shall be

established to assure that all testing required to demonstrate that structures, systems,

and components will perform satisfactorily in service is identified and performed in

accordance with written procedures. Specifically, prior to August 3, 2017, the licensee

failed to have a program to completely test the interlock circuit for safety injection pump

and recirculation suction isolation valves, EJ-HV-8804A and EJ-HV-8804B. When the

licensee personnel performed a review the interlock circuits for the valves, they identified

that there had been gaps in the testing. In response to this issue, the licensee

investigated all of the testing activities associated with the valve interlock circuits and

identified that in 2010, a comprehensive test of the circuits had been performed as the

result of a modification. The licensee has entered this issue into their corrective action

program as Condition Report CR-201703962.

The team determined that the failure to develop and implement testing programs for

verifying that the circuits for the multiple interlocks associated with safety injection

valve EJ-HV-8804A would perform as designed was a performance deficiency. The

performance deficiency was more than minor, and therefore a finding, because it was

associated with the equipment performance attribute of the Mitigating Systems

Cornerstone and adversely affected the cornerstone objective to ensure the availability,

reliability, and capability of systems that respond to initiating events to prevent

undesirable consequences. Specifically, the licensee failed to establish a testing

program to verify that the valve interlock circuits for valve EJ-HV-8804A were being

tested. A failure of the interlocks and an operator error could result in an inadvertent

release path to the environment. In accordance with Inspection Manual Chapter 0609,

Appendix A, The Significance Determination Process (SDP) for Findings At-Power,

dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue

screened as having very low safety significance (Green) because it was a design or

qualification deficiency that did not represent a loss of operability or functionality; did not

represent an actual loss of safety function of the system or train; did not result in the loss

of one or more trains of non-technical specification equipment; and did not screen as

potentially risk-significant due to seismic, flooding, or severe weather. The team

determined that this finding did not have a cross-cutting aspect because the most

significant contributor did not reflect current licensee performance.

(Section 1R21.2.8.b.1)

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

Criterion III, Design Control, which requires, in part, that measures shall be established

to assure that the design basis is correctly translated into procedures and instructions.

Specifically, prior to on August 4, 2017, the licensee had design calculations that

assumed operator actions to mitigate internal flooding of certain areas within specified

time durations. These time requirements for the design basis flooding calculations had

not been translated into any procedures or instructions. In response to this issue, the

licensee performed a preliminary evaluation and determined that operator actions to

support the design calculation could be performed within the time required. The

licensee has entered this issue into their corrective action program as Condition

Report CR-201703981.

The team determined that the failure to translate operator time requirements for

mitigating design basis flooding of critical areas into procedures or instructions was a

performance deficiency. The performance deficiency was more than minor, and

therefore a finding, because it was associated with the equipment performance attribute

of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective

to ensure the availability, reliability, and capability of systems that respond to initiating

events to prevent undesirable consequences. Specifically, the licensee failed to confirm

that design basis inputs had been translated into procedures or instructions. In

accordance with NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2,

"Mitigating Systems Screening Questions," the issue screened as having very low safety

significance (Green) because it was a design or qualification deficiency that did not

represent a loss of operability or functionality; did not represent an actual loss of safety

function of the system or train; and did not result in the loss of one or more trains of

nontechnical specification equipment. The team determined that this finding did not

have a cross-cutting aspect because the most significant contributor did not reflect

current licensee performance. (Section 1R21.2.8.b.2)

Licensee-Identified Violations

A violation of very low safety significance was identified by the licensee and has been reviewed

by the team. Corrective actions taken or planned by the licensee have been entered into the

licensees corrective action program. This violation and associated corrective action tracking

numbers are listed in Section 4OA7 of this report.

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REPORT DETAILS

1. REACTOR SAFETY

Inspection of component design bases and modifications made to structures, systems,

and components verifies that plant components are maintained within their design basis.

Additionally, this inspection provides monitoring of the capability of the selected

components and operator actions to perform their design bases functions. The

inspection also monitors the implementation of modifications to structures, systems, and

components. Modifications to one system may also affect the design bases and

functioning of interfacing systems as well as introduce the potential for common cause

failures. As plants age, modifications may alter or disable important design features

making the design bases difficult to determine or obsolete. The plant risk assessment

model assumes the capability of safety systems and components to perform their

intended safety function successfully. This inspectable area verifies aspects of the

Initiating Events, Mitigating Systems, and Barrier Integrity Cornerstones for which there

are no indicators to measure performance.

1R21 Component Design Bases Inspection (71111.21M)

The inspection team selected risk-significant components, industry operating experience

issues, modifications, and operator actions for review using information contained in the

licensees probabilistic risk assessment. In general, this included components, industry

operating experience issues, modifications, and operator actions that had a risk

achievement worth factor greater than 2 or a Birnbaum value greater than 1E-6.

.1 Inspection Scope for Components Selected

To verify that the selected components and modification would function as required,

the team reviewed design basis assumptions, calculations, and procedures. In some

instances, the team performed calculations to independently verify the licensee's

conclusions. The team also verified that the condition of the components was consistent

with the design bases and that the tested capabilities met the required criteria.

The team reviewed maintenance work records, corrective action documents, and

industry operating experience records to verify that licensee personnel considered

degraded conditions and their impact on the components. For the review of operator

actions, the team observed operators during simulator scenarios, as well as during

simulated actions in the plant.

The team performed a margin assessment and detailed review of the selected

risk-significant components to verify that the design bases have been correctly

implemented and maintained. This design margin assessment considered original

design issues, margin reductions because of modifications, and margin reductions

identified as a result of material condition issues. Equipment reliability issues were also

considered in the selection of components for detailed review. These included items

such as failed performance test results; significant corrective actions; repeated

maintenance; 10 CFR 50.65(a)1 status; operable, but degraded conditions; NRC

resident inspector input of problem equipment; system health reports; industry operating

experience; and licensee problem equipment lists. Consideration was also given to the

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uniqueness and complexity of the design, operating experience, and the available

defense in-depth margins.

The team selected permanent plant modifications, including permanent plant changes,

design changes, set point changes, procedure changes, equivalency evaluations,

suitability analyses, calculations, and commercial grade dedications to verify that design

bases, licensing bases, and performance capability of components have not been

degraded through modifications. The team determined whether post-modification testing

established operability. The team verified that supporting design basis documentation,

such as calculations, design specifications, vendor manuals, the updated final safety

analysis report, technical specification and bases, and plant specific safety evaluation

reports, were updated consistent with the design change. The team verified that other

design basis features, such as structural, fire protection, flooding, environmental

qualification, and potential emergency core cooling system strainer blockage mitigation,

which could be affected by the modification, were not adversely impacted. The team

verified that procedures and training plans, such as abnormal operating procedures,

alarm response procedures, and licensed operator training manuals, affected by the

modifications were updated.

The inspection procedure requires a review of four to six components based on risk

significance and four to six modifications to mitigation structures, systems, and

components. One of the inspection samples selected shall be associated with

containment-related structures, systems, and components which are considered for

large early release frequency (LERF) implications. The samples selected for this

inspection were eight components (one containment-related component),

five modifications, and three operating experience items.

The selected inspection items supported risk-significant functions as follows:

• Electrical power to mitigation systems: The team selected several components in

the offsite and on-site electrical power distribution systems to verify operability to

supply alternating current (ac) and direct current (dc) power to risk-significant and

safety-related loads in support of safety system operation in response to initiating

events, such as loss-of-offsite-power accident, station blackout, and a loss-of-coolant

accident with offsite power available. As such, the team selected:

• Pressurizer pressure transmitters (BBPT-0455, -0456, -0457, and -0458).

• Pressurizer power-operated relief valve PCV455A, including its associated block

valve.

• Modifications: pressurizer pressure transmitter replacement (MP-08-0054) and

replace the Ametek Iso-Limiter transformers XPN07 and XPN08 (MP-09-0051).

• Operating Experience: IN 2012-003 Design Vulnerability in Electric Power

Systems and associated Modification MP15-0008; RIS 2011-12, Revision 1,

Adequacy of Station Electric Distribution System Voltages.

• Motor-operated valves BGLCV0112 B/C charging pump suction isolation valves

from the volume control tank.

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• Safety-related 4KV and 480V switchgear NB01 and NG01 and associated

breaker replacement Modifications MP07-0070 and MP07-0069.

• Components that affect LERF: The team reviewed components required to perform

functions that mitigate or prevent an unmonitored release of radiation. The team

selected the following components:

• Safety Injection piggyback valve EJ-HV-8804A.

• Mitigating systems needed to attain safe shutdown: The team reviewed components

required to perform the safe shutdown of the plant. As such the team selected:

• Residual heat removal pump A.

• Essential service water returns to ultimate heat sink valve EFHV0037.

• Modifications: TM 14-0003 and MP 05-3025.

• Safety injection check valve EM8926A.

• Modification 10-0009 (reactor coolant pump seal replacement).

.2 Results of Detailed Reviews of Components

.2.1 Pressurizer Pressure Transmitters: BBPT0455, BBPT0456, BBPT0457, and BBPT0458

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design

basis documents, the current system health report, selected drawings and calculations,

maintenance and test procedures, and condition reports associated with the pressurizer

pressure transmitters. The team also reviewed photos detailing the installed

configuration and conducted interviews with system and design engineering personnel to

ensure the capability of these components to perform their desired design basis function.

Specifically, the team reviewed:

• Component maintenance history and corrective action program reports to verify the

monitoring of potential degradations.

• As-built installation drawings and equipment to verify that the equipment and

associated raceways are installed correctly to meet the environmental requirements.

• The environmental qualifications testing evaluations for the replacement transmitters,

conduit seal assemblies, connector assemblies, and splicing procedures.

• Equivalency evaluations for the replacement transmitters for applicability.

b. Findings

No findings were identified.

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.2.2 Pressurizer Power Operated Relief Valve PCV455A including its associated block valve

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, the current

system health report, selected drawings, maintenance procedures, test procedures, and

condition reports associated with the pressurizer power-operated relief valve PCV455A

including its associated block valve. The team also reviewed photos detailing the

installed configuration and conducted interviews with system engineering personnel to

ensure capability of this component to perform its desired design basis function.

Specifically, the team reviewed:

• Piping and instrumentation drawing, schematic control and power drawings for the

power-operated relief valves and the associated block valves.

• Power-operated relief valve inservice test closing and opening speeds for the

two previous in service tests.

• Motor-operated valve block valve sizing and torque calculations.

• Motor-operated valve block valve breaker and overload overcurrent protection

specification calculations.

• Motor-operated valve block valve voltage drop calculations.

• Motor-operated valve block valve stroke timing tests.

b. Findings

No findings were identified.

.2.3 Motor-Operated Valves BGLCV0012 B/C Charging Pump Suction Isolation Valves from

the Volume Control Tank

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, the current

system health report, selected drawings, maintenance procedures, test procedures, and

condition reports associated with motor-operated valves BGLCV0012 B/C, charging

pump suction isolation valves from the volume control tank. The team also conducted

interviews with system engineering personnel to ensure capability of the component to

perform its desired design basis function. Specifically, the team reviewed:

• Motor calculations that establish the motor voltage drop, protection and coordination

and short circuit for the motor power supply and feeder cables.

• Calculations for the degraded voltage at the motor-operated valve terminals to

ensure the proper voltage was utilized in the teams review of motor-operated valve

torque calculations.

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• Calculations that establish motor-operated valve control circuit voltage drop, short

circuit, and protection/coordination including thermal overload sizing and application.

b. Findings

No findings were identified.

.2.4 Safety-Related 4KV and 480V SWGR NB01 and NG01 and Associated Breaker

Replacements

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design

basis documents, the current system health report, selected drawings and calculations,

maintenance and test procedures, and corrective action program reports associated with

SR 4KV and 480V SWGR NB01 and NG01, and their associated breaker replacements.

The team also performed walkdowns and conducted interviews with system engineering

personnel to ensure the capability of this component to perform its desired design basis

function. Specifically, the team reviewed:

• 4.16KV and 480V breaker replacement modifications MP-07-0069 and MP-07-0070

were reviewed to ensure the adequacy and consistency of design for the new

replacement breakers.

• Corrective action and maintenance history documents and system health reports to

determine whether there were any adverse operating trends and to assess the

stations ability to evaluate and correct problems.

• Calculations for electrical distribution, system load flow/voltage drop, short-circuit,

and electrical protection to verify that bus capacity and voltages remained within

minimum acceptable limits.

• Protective device settings and circuit breaker ratings to ensure adequate selective

protection coordination of connected equipment during worst-case short circuit

conditions.

• Degraded and loss of voltage relays and associated time delays were set in

accordance with calculations, and that associated calibration procedures were

consistent with calculation assumptions, associated time delays and set point

accuracy calculations.

• Coordination and interface with the transmission system operator for plant voltage

requirements and notification set points were reviewed.

• Procedures for preventive maintenance, inspection, and testing to compare

maintenance practices against industry and vendor guidance.

• Visual non-intrusive inspection to assess material condition, the presence of

hazards, and consistency of installed equipment with design documentation and

analyses.

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b. Findings

Not Verifying the Operation and Timing of the Engineered Safety Feature

Transformer XNB01 Load Tap Changer

Introduction. The team identified a Green, cited violation of Technical

Specification 5.4.1.a, Procedures, involving the failure to implement adequate

maintenance procedures to periodically verify transformer XNB01 load tap changer

operation and time testing. Specifically, due to the ineffective corrective action of

Callaway Action Requests CAR-200202970, CAR-201402827, CAR-201405312, and

CAR-201508240, the licensee did not implement preventative maintenance activities to

verify the operation and timing of the engineered safety feature transformer XNB01 load

tap changer. As a result, the timing of the load tap changer may not be consistent with

plant electrical analysis, ZZ-62, which credits the load tap changer operation in order to

reset the degraded voltage relays between sequenced load steps.

Description. In 2001, under modification MP 99-1083, the licensee installed engineered

safety feature transformers XNB01 and XNB02 with load tap changers. During the

installation, the licensee performed a review of industry operating experience and found

information identifying that time testing of the load tap changer operation was required to

confirm that the load tap changers would work properly. This would ensure operability of

the off-site power sources. Operating experience had shown that the load tap changer

mechanical operation could slow down over time due to aging mechanisms such as

friction and hardened grease. This could result in the unmonitored degraded

performance of the load tap changer to not provide acceptable voltages from the offsite

power sources to the safety-related power distribution system. As a result, the expected

speed of the load tap changer, to correct for low voltage, may not meet design

requirements.

Callaway Action Request CAR-200202970 was written to ensure that a preventive

maintenance activity was generated to periodically check for proper load tap changer

operation and timing. Callaway Action Request CAR-200202970 was closed to the

Maintenance Optimization Project.

In 2006, the preventative maintenance basis and transformer preventative maintenance

was initially created, but the preventative maintenance activity did not include the timing

requirements for the load tap changers. In CAR-200909389, as a result of Nuclear

Electric Insurance Limited insurance requirements, the licensee changed the frequency

of testing the on-site transformers, including the transformers XNB01 and XNB02, from

every 8 refueling outages to every four refueling outages (every 6 years).

In May 2014, the NRC issued Violation 05000483/2014007-001 (CAR-201402827 and

CAR-201405312), due to the ineffective corrective action of CAR-200202970, where the

licensee had not established preventative maintenance procedures to verify the

operation and timing of the engineered safety feature transformers XNB01 and XNB02

load tap changers. On June 9, 2014, Preventive Maintenance Procedures, PM1001510

and PM1001506 for transformers XNB01 and XNB02, respectively, were revised to

include timing tests.

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Under Job 08510933.510, the operation and timing test for transformer XNB02

was performed in the fall of 2014 (R19), after the 2014 violation was identified. The

recorded time between steps of the load tap changer alternated between 0.9 and

2.9 seconds. This met the acceptance criteria of less than 3 seconds (identified in

licensee calculation ZZ-62), but was not in accordance with the typical times from the

Reinhausen (load tap changer manufacturer) Vendor Manual of 2 seconds per step.

During the 2017 NRC design basis assurance inspection, the team questioned the

testing results. This had been the first time this test had been performed at the Callaway

Plant.

During the licensees performance of Formal Self-Assessment 201500920-18, Problem

and identification and Resolution Pre-Inspection Assessment, they identified three

examples where response to non-cited violations had not been timely, potentially

representing a violation of their procedures. One of these examples was the non-cited

violation for failure to test the timing of engineered safety feature transformer load tap

changers. The load tap changer for the B Train XNB02 was tested in RF20, by

job 085109333, and had no issues identified. Therefore, the licensee felt as though

there was reasonable assurance that the load tap changer for XNB01 was acceptable

without testing and was scheduled to be tested on April 19, 2019, even though it had not

been tested since 2001 and that there was operating experience available pertaining to

a decline in load tap changer performance over time due to aging and hardening of

lubrication. After further review, the licensee moved the testing of the load tap changer

for XNB01 to the fall of 2017 (Refuel 21).

On August 2, 2017, the team identified that the licensee had not implemented

PM1001510 to perform a timing test of the transformer XNB01 load tap changer and

had credited the successful testing of transformer XNB02 in 2014 as partial justification

to extend the testing on transformer XNB01 until 2019 (R22). As a result of initiating

Callaway Action Report CAR-201508240, the licensee changed the scheduling of the

testing of transformer XNB01, including the load tap changer, to the fall of 2017 (R21).

During the 2017 NRC design basis assurance inspection, the team questioned the 2014

testing results of XNB02 load tap changer. This had been the first time this test had

been performed at the Callaway Plant, and none of the licensees personnel had

questioned the differences in the licensees results of recorded data of 9.9 and

2.9 seconds between step changes versus the typical times from the Reinhausen (load

tap changer manufacturer) Vendor Manual of 2 seconds per step. After two weeks of

internal and external discussions with the load tap changer manufacturer, the licensee

concluded that the test results were acceptable, but noted that the data obtained from

the testing would be different depending on whether the measurements are taken of

voltage changes versus load tap changer position changes. Since obtaining the timing

data for the load tap changer in 2014 for transformer XNB02, the licensee had not

questioned the differences in the testing data obtained compared to what the vendor

identified as the expected time for position changes of the load tap changer. The team

determined that acceptance of the load tap changer testing results in 2014 without a

questioning attitude of why the results were not in accordance with the vendor manual

was unacceptable.

Callaway Technical Specification 5.4.1.a requires, in part, that written procedures shall

be established, implemented, and maintained covering the applicable procedures

recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.

11

Regulatory Guide 1.33, Appendix A, Section 9, Procedures for Performing

Maintenance, requires, in part, that maintenance that can affect the performance of

safety-related equipment should be properly pre-planned and performed in accordance

with documented instructions appropriate to the circumstances.

The team determined that the licensee had not: 1) adequately performed a timing test of

the transformer XNB01 load tap changer to ensure proper operation; and 2) periodically

performed a timing test of the transformer XNB01 load tap changer to ensure proper

operation to maintain the operability of the offsite power sources. Since the time that the

NRC issued the violation in 2014, the licensee had opportunities to perform a timing test

of transformer XNB01 load tap changer (refueling outages in the fall of 2014, and the

refueling outage in the spring of 2016).

Analysis. The team determined that the failure to implement maintenance procedures to

periodically verify transformer XNB01 load tap changer operation and time testing was a

performance deficiency. The performance deficiency was more than minor, and

therefore a finding, because it was associated with the equipment performance attribute

of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective

to ensure the availability, reliability, and capability of systems that respond to initiating

events to prevent undesirable consequences. Specifically, the failures to perform,

periodic verification of the operation and time testing of the load tap changer could result

in adverse operation of the load tap changer during a design basis event. If the load tap

changer did not operate correctly, the safety-related buses may not have adequate

voltage to reset the degraded voltage relay, thus spuriously disconnecting from the

offsite power source.

In accordance with Inspection Manual Chapter 0609, Appendix A, Significance

Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2,

Mitigating Systems Screening Questions, the issue screened as having very low safety

significance (Green) because it was a design or qualification deficiency that did not

represent a loss of operability or functionality; did not represent an actual loss of safety

function of the system or train; did not result in the loss of one or more trains of non-

technical specification equipment; and did not screen as potentially risk-significant due to

seismic, flooding, or severe weather. The finding had a cross-cutting aspect in the area

of human performance, work management, because the licensee failed to plan, control,

and execute work activities such that nuclear safety is the overriding priority.

Specifically, the licensee did not plan and execute the testing of the transformer XNB01

load tap changer in a timely manner (H.5).

Enforcement. The team identified a Green, cited violation of Technical Specification 5.4.1.a, which requires, in part, that written procedures shall be

established, implemented, and maintained covering the applicable procedures

recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.

Regulatory Guide 1.33, Appendix A, Section 9, Procedures for Performing

Maintenance, requires, in part, that maintenance that can affect the performance of

safety-related equipment should be properly pre-planned and performed in accordance

with documented instructions appropriate to the circumstances. Contrary to the above,

from May 2014 through August 4, 2017, the licensee failed to ensure that maintenance

that can affect the performance of safety-related equipment be properly pre-planned and

perform in accordance with documented instructions appropriate to the circumstances.

Specifically, as a result of ineffective corrective action of Callaway Action

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Requests CAR-201402827 and CAR-201405312, the licensee failed to perform

preventative maintenance procedures to verify the operation and timing of the

engineered safety feature transformer XNB01 load tap changer. This violation was

previously identified by the NRC and documented as NCV 05000483/2014007-01. In

accordance with Section 2.3.2.a of the NRC Enforcement Policy, this finding is being

cited because the licensee failed to restore compliance within a reasonable amount of

time after the violation was initially identified. This finding was entered into the

licensees corrective action program as Condition Report CR-201703992,

VIO 05000458/2017007-01, Not Verifying the Operation and Timing of the Engineered

Safety Feature Transformer XNB01 Load Tap Changer.

.2.5 Residual Heat Removal Pump A

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design basis

documents, the current system health report, selected drawings and calculations,

maintenance and test procedures, and condition reports associated with residual heat

removal pump A. The team also performed walkdowns and conducted interviews with

system engineering personnel to ensure the capability of this component to perform its

desired design basis function. Specifically, the team reviewed:

• Calculations and equipment room survivability analysis associated with a loss of

ventilation in the residual heat removal pump A room. The team verified that a loss

of room ventilation would not result in the room temperature going above the

maximum allowable pump motor design temperature.

• Recalculation of piping heat transfer into the residual heat removal pump A room.

The team verified that all insulated and uninsulated piping was accounted for and

that the room coolers were capable of providing adequate cooling during worst-case

scenarios.

• System health reports, component maintenance history, and corrective action

program reports to verify the monitoring and correction of potential degradation.

b. Findings

No findings were identified.

.2.6 Essential Service Water Returns to Ultimate Heat Sink Valve EFHV0037

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, selected

drawings, maintenance and test procedures, and condition reports associated with the

essential service water returns to ultimate heat sink valve EFHV0037. The team also

performed walkdowns and conducted interviews with system engineering personnel to

ensure the capability of this component to perform its desired design basis function.

Specifically, the team reviewed:

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• Calculations and implementation of the inservice testing program associated with

maintaining an adequate margin for its safety function to open.

• System health reports, component maintenance history, and corrective action

program reports to verify the monitoring and correction of potential degradation.

b. Findings

No findings were identified.

.2.7 Safety Injection check valve EM8926A

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, selected

drawings, maintenance and test procedures, and condition reports associated with the

safety injection check valve EM8926A. The team also performed walkdowns and

conducted interviews with system engineering personnel to ensure the capability of this

component to perform its desired design basis function. Specifically the team reviewed:

• Leakage test procedures to verify the allowable leakage from the emergency core

cooling system to the refueling water storage tank under accident conditions.

• Recent inservice test results associated with this valve.

• The basis for the inservice leakage test acceptance criteria for this valve and

associated valves to verify the total allowable leakage from the emergency core

cooling system to the refueling water storage tank under accident conditions.

b. Findings

No findings were identified.

.2.8 Safety Injection Piggyback Valve EJ-HV-8804A.

a. Inspection Scope

The team reviewed the updated safety analysis report, system description, design basis

documents, the current system health report, selected drawings and calculations,

maintenance and test procedures, and condition reports associated with safety injection

piggyback valve EJ-HV-8804A. The team also performed walkdowns and conducted

interviews with system and design engineering personnel to ensure the capability of

these components to perform their desired design basis function. Specifically, the team

reviewed:

• The design thrust calculations to verify the capability of the valve to perform its

design function under limiting design conditions.

• Results of recent motor-operated valve diagnostic testing to verify the current

capability of the valve.

14

• Calculations associated with the transfer of the emergency core cooling system from

the refueling water storage tank to the containment sump to verify the basis of the

valves required stroke time.

• Operating procedures associated with the use of the valve under post-accident

conditions to verify its operation was consistent with its design.

• Testing of control circuits associated with valve interlocks to verify the capability of

the interlocks to function as designed.

b. Findings

1. Safety Injection Piggyback Valve EJ-HV-8804A Valve Interlocks Not Tested.

Introduction. The team identified a Green, non-cited violation of 10 CFR Part 50,

Appendix B, Criterion XI, Test Control, for the licensees failure to fully test the

control circuits associated with the interlocks for valve EJ-HV-8804A. Specifically,

the team identified that the licensee failed to have a program to completely test the

interlock circuit for safety injection pump and recirculation suction isolation valves,

EJ-HV-8804A and B. Also, when the licensee did review the interlock circuits for the

valves, they identified that there had been gaps in their testing (i.e. that some of the

contacts in the circuit had not been tested).

Description. The team questioned the licensee on whether the interlock circuits

associated with valve EJ-HV-8804A were periodically tested. In accordance with the

guidance of the Institute of Electrical and Electronics Engineers Standard 379-1972,

potential undetectable failures must be assumed to be in their failed mode prior to a

postulated accident. The team identified that the licensee failed to have a program

to completely test the interlock circuit for safety injection pump and recirculation

suction isolation valves, EJ-HV-8804A and EJ-HV-8804B. Also, when the licensee

did review the interlock circuits for the valves, they identified that there had been

gaps in their testing (i.e., that some of the contacts in the circuit had not been

tested). The Final Safety Analysis Report, Section 6.3.2.1, states, The safety

injection pump and emergency core cooling system charging pump recirculation

suction isolation valves, EJ-HV-8804A and EJ-HV-8804B, can be opened provided

that either the safety injection system minimum flow isolation valve, BN-HV-8813,

or both safety injection pump minimum flow isolation valves, EM-HV-8814A and B,

are closed. Additionally, one of the two residual heat removal hot leg suction valves

on Loop 1, BB-PV-8702A and EJ-HV-8701A, and on Loop 4, BB-PV-8702B and

EJ-HV-8701B, must be closed. In response to this issue, the licensee

investigated all of the testing activities associated with the valve interlock circuits

and identified that in 2010, a comprehensive test of the circuits had been

performed, with acceptable results, as the result of a modification. In Condition

Report CR-201703962, the licensees immediate operability determination concluded

that based on review of the condition description, EJ-HV-8804A was operable, but

degraded or nonconforming. The licensee stated that while these interlocks are not

being programmatically tested on a periodic basis, all the valve interlocks in the

OPEN circuit/logic for EJ-HV-8804A had been previously tested to verify they OPEN

when their associated valve is OPEN. Based on previous testing, there was

15

reasonable assurance that the interlocks contacts will open when their associated

valves are open, to prevent inadvertent opening of EJ-HV-8804A. The licensee

has entered this issue into their corrective action program as Condition

Report CR-201703962.

Analysis. The team determined that the failure to develop and implement testing

programs for verifying that the circuits for the multiple interlocks associated with

safety injection valve EJ-HV-8804A would perform as designed was a performance

deficiency. The performance deficiency was more than minor, and therefore a

finding, because it was associated with the equipment performance attribute of the

Mitigating Systems Cornerstone and adversely affected the cornerstone objective

to ensure the availability, reliability, and capability of systems that respond to

initiating events to prevent undesirable consequences. Specifically, the licensee

failed to establish a testing program to verify that the valve interlock circuits for

valve EJ-HV-8804A were being tested. A failure of the interlocks and an operator

error could result in an inadvertent release path to the environment. In accordance

with Inspection Manual Chapter 0609, Appendix A, The Significance Determination

Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating

Systems Screening Questions, the issue screened as having very low safety

significance (Green) because it was a design or qualification deficiency that did not

represent a loss of operability or functionality; did not represent an actual loss of

safety function of the system or train; did not result in the loss of one or more

trains of nontechnical specification equipment; and did not screen as potentially

risk-significant due to seismic, flooding, or severe weather. The team determined

that this finding did not have a cross-cutting aspect because the most significant

contributor did not reflect current licensee performance.

Enforcement. The team identified a Green, non-cited violation of 10 CFR Part 50,

Appendix B, Criterion XI, Test Control, which requires, in part, that a test program

shall be established to assure that all testing required to demonstrate that structures,

systems, and components will perform satisfactorily in service is identified and

performed in accordance with written procedures. Contrary to the above, prior to

August 3, 2017, the licensee failed to establish a test program to assure that all

testing required to demonstrate that structures, systems, and components will

perform satisfactorily in service is identified and performed in accordance with

written procedures. Specifically, the licensee failed to have a program to completely

test the interlock circuit for safety injection pump and recirculation suction isolation

valves, EJ-HV-8804A and EJ-HV-8804B. When the licensee personnel performed a

review the interlock circuits for the valves, they identified that there had been gaps in

the testing. In response to this issue, the licensee investigated all of the testing

activities associated with the valve interlock circuits and identified that in 2010, a

comprehensive test of the circuits had been performed as the result of a

modification. The licensee has entered this issue into their corrective action program

as Condition Report CR-201703962. Because this finding was of very low safety

significance and has been entered into the licensees corrective action program, this

violation is being treated as a non-cited violation consistent with Section 2.3.2.a of

the NRC Enforcement Policy: NCV 05000483/2017007-02, Safety Injection

Piggyback Valve EJ-HV-8804A Interlocks Not Tested.

16

2. Inputs to Internal Flooding Calculations Not Translated into Procedures or

Instructions.

Introduction. The team identified a Green, non-cited violation of 10 CFR Part 50,

Appendix B, Criterion III, Design Control, for the licensees failure to ensure that

design basis requirements were correctly translated into procedures and instructions.

Specifically, design calculations assumed operator actions to mitigate an internal

flood of certain areas within specified time durations. These time requirements for

the design basis flooding calculations had not been translated into any procedures or

instructions.

Description. The licensee had performed numerous calculations to address internal

flooding concerns for multiple areas throughout the plant. Several of these

calculations did not clearly differentiate between commercial and design basis

acceptance criteria. These calculations also assumed operator actions to mitigate

the flood within specified time durations. These time requirements for the design

basis flooding calculations had not been translated into any procedures or

instructions. In a previous Callaway Action Request, CAR--200605158, the licensee

confirmed that the credited operator response times had been evaluated. However,

these times were not included in Procedure APA-ZZ-00395, Significant Operator

Response Timing. In response to this issue, the licensee reviewed their flooding

calculations to determine which calculations document commercial margin verses

design basis margin. The licensee will be updating these documents to clearly

describe the flooding program design basis requirements. Also, the licensee

performed a preliminary evaluation and determined that operator actions to support

the design calculations could be performed within the time required. The licensee

has entered this issue into their corrective action program as Condition Report

CR-201703981.

Analysis. The team determined that the failure to translate operator time

requirements for mitigating design basis flooding of critical areas into procedures or

instructions was a performance deficiency. The performance deficiency was more

than minor, and therefore a finding, because it was associated with the equipment

performance attribute of the Mitigating Systems Cornerstone and adversely affected

the cornerstone objective to ensure the availability, reliability, and capability of

systems that respond to initiating events to prevent undesirable consequences.

Specifically, the licensee failed to confirm that design basis inputs had been

translated into procedures or instructions. In accordance with NRC Inspection

Manual Chapter 0609, Appendix A, Exhibit 2, "Mitigating Systems Screening

Questions," the issue screened as having very low safety significance (Green)

because it was a design or qualification deficiency that did not represent a loss of

operability or functionality; did not represent an actual loss of safety function of the

system or train; and did not result in the loss of one or more trains of nontechnical

specification equipment. The team determined that this finding did not have a cross-

cutting aspect because the most significant contributor did not reflect current

licensee performance.

Enforcement. The team identified a Green, non-cited violation of 10 CFR Part 50,

Appendix B, Criterion III, Design Control, which requires, in part, that measures

shall be established to assure that the design basis is correctly translated into

procedures and instructions. Contrary to the above, prior to August 4, 2017, the

17

licensee failed to ensure that design basis was correctly translated into procedures

and instructions. Specifically, the licensee had design calculations that assumed

operator actions to mitigate internal flooding of certain areas within specified time

durations. These time requirements for the design basis flooding calculations had

not been translated into any procedures or instructions. In response to this issue,

the licensee performed a preliminary evaluation and determined that operator actions

to support the design calculations could be performed within the time required. The

licensee has entered this issue into their corrective action program as Condition

Report CR-201703981. Because this finding was of very low safety significance

and has been entered into the licensees corrective action program, this violation

is being treated as a non-cited violation consistent with Section 2.3.2.a of the

NRC Enforcement Policy: NCV 05000483/2017007-03, Inputs to Internal Flooding

Calculations Not Translated into Procedures or Instructions.

.3 Results of Detailed Reviews of Permanent Plant Modifications

a. Inspection Scope

The team reviewed five permanent plant modifications that had been installed in the

plant during the last three years. This review included in-plant walkdowns for portions of

the accessible systems. The modifications were selected based upon risk significance,

safety significance, and complexity. The team reviewed the modifications selected to

determine if:

• Supporting design and licensing basis documentation was updated.

• Changes were in accordance with the specified design requirements.

• Procedures and training plans affected by the modification have been adequately

updated.

• Test documentation as required by the applicable test programs has been updated.

• Post-modification testing adequately verified system operability and/or functionality.

The team also used applicable industry standards to evaluate acceptability of the

modifications.

.3.1 Modification: MP-08-0054, Replace Pressurizer Pressure Transmitters.

The team reviewed Modification MP 08-0054, implemented to replace pressurizer

pressure transmitters BBPT0455, BBPT0456, BBPT0457, and BBPT0458. The purpose

of the modification was to approve and replace obsolete Tobar 32PA1212 transmitters

with Rosemount 1154 Type H pressure transmitters. Due to physical differences with

the replacement transmitters and the environmental requirements of the installation, the

modification package also required the replacement of the transmitter conduit seal and

connector assembly. The pressurizer pressure transmitters are used in the reactor trip

system and also provide input to the engineered safety feature actuation system and are

required to operate before and during a design basis event to provide safety functions.

18

Modification MP-08-0054 evaluated several different replacement transmitters including

Rosemount model 3154, Ametek model PG3200, and Ultra Model N-E11GH. The

Rosemount 1154 was determined to be the best match and was also qualified for the

environment. In addition to the transmitter replacement the conduit and connector

assembly were required be included in the overall modification package in order to meet

the environmental requirements of the installation. The team reviewed the

environmental qualifications evaluations of the transmitters, conduit seal assemblies,

connector assembly and required cable splicing methods to insure the final installation

met all environmental requirements. The team did not identify any issues with the

licensees implementation of this modification.

.3.2 Modifications MP07-0069, Replace 480V Load Center Breakers, and MP07-0070

Replace Safety-Related and Non-Safety Metal-Clad Breakers

The team reviewed the plant modification packages associated with the replacement of

safety-related 4160V switchgear breakers and 480V load center switchgear breakers.

The existing plant breakers were becoming obsolete and spare parts becoming more

expensive as maintenance and overhauls were coming due. The 4160V breakers are

being replaced with Square D Magnum type SVR vacuum breakers and the 480V load

center breakers with Square D Masterpact circuit breakers. The team reviewed the

Appendix B purchase specifications, qualification reports, and resulting calculations

revised to support the change in breakers. The team did not identify any issues with the

licensees implementation of this modification.

.3.3 Modification: MP 05-3025, Maximum Allowed Temperature of CST and AFW System

The team reviewed Modification MP 05-3025, which increased the design rating and

service condition maximum temperature of pipes, valves and equipment in the

condensate transfer and storage system and the auxiliary feedwater system from a

normal operating temperature of 95 ºF to 110 ºF. The modification did not implement

any physical changes to the plant.

The condensate storage tank water temperature has exceeded the 95 ºF normal

operating temperature on multiple occasions throughout the operation of the plant. The

primary concern of raising the water temperature involved the stress rating of the piping

systems and the effects on the available net positive suction head of the auxiliary

feedwater pump. The licensee verified that all piping, valves, and equipment were rated

to operate at higher temperatures than the operating temperature of 95 ºF by referencing

design documents, system calculations, and vendor correspondence. The team did not

identify any issues with the licensees implementation of this modification.

.3.4 Modification¨ M 10-0009, Reactor Coolant Pump Seal Replacement

The team reviewed Modification Package 10-0009, implemented to new Westinghouse

SHIELD passive thermal shutdown seal on each of the reactor coolant pumps.

Westinghouse developed a reactor coolant pump shutdown seal, the Westinghouse

Reactor Coolant Pump SHIELD Passive Thermal Shutdown Seal, that restricts reactor

coolant system inventory losses to very small values for plant events that result in the

loss of all reactor coolant pump seal cooling. The shutdown seal is a thermally actuated,

passive device that is integral to the No. 1 insert, and sits between the No. 1 seal and

the No. 1 seal leak-off line, to provide a leak-tight seal in the event of a loss of all reactor

19

coolant pump seal cooling. The review included the design change package, post-

modification testing, and the associated 10 CFR 50.59 review. The team did not identify

any issues with the licensees implementation of this modification.

b. Findings

No findings were identified.

.4 Results of Detailed Reviews of Operating Experience

.4.1 Inspection of IN 2012-003, Design Vulnerability in Electric Power System, and

associated Modification MP15-0008

The team reviewed the licensee evaluation of Information Notice 2012-003, Design

Vulnerability in Electric Power System, to verify that the licensee initially performed an

applicability review and took corrective actions, if appropriate, to address the concerns

described in the information notice summary. The team additionally reviewed the

licensees proposed design modification MP15-0008, Open Phase Condition

Protection, to address and resolve the concerns described in the information notice.

The licensee entered this issue into their corrective action program as Callaway Action

Requests CARs 201201245, 201201652, 201205441, 201302829, and 201309622. The

team did not identify any concerns with how the licensee is addressing this operating

experience.

.4.2 Inspection of RIS 2011-12, Revision 1, Adequacy of Station Electric Distribution System

Voltages

The team reviewed the licensees evaluation of Regulatory Issue Summary 2011-12,

Revision 1, Adequacy of Station Electric Distribution System Voltages, to verify that the

licensee performed an applicability review and took corrective actions, if appropriate, to

address the concerns described in the regulatory issue summary. This regulatory issue

summary was issued to clarify the NRC staffs technical position on existing regulatory

requirements. The licensee entered this issue into their corrective action program as

Callaway Action Request CAR-201200050. The team did not identify any concerns with

how the licensee addressed this operating experience.

.4.3 NRC Information Notice 2017-03, Anchor/Darling Double Disc Gate Valve Wedge Pin

and Stem-Disc Separation Failures

The team reviewed the licensees evaluation of Information Notice 2017-03,

Anchor/Darling Double Disc Gate Valve Wedge Pin and Stem-Disc Separation

Failures, and the associated Part 21 notification to verify that potential valve disc

separation issues were appropriately addressed. This information notice addressed the

failures of an Anchor/Darling gate valve due to stem-disc separation events. The team

interviewed engineering personnel and reviewed corrective action documentation to

verify that potentially vulnerable valves had been identified and evaluated. The team did

not identify any concerns with how the licensee is addressing this operating experience.

20

.5 Results of Reviews for Operator Actions

a. Inspection Scope

The team selected risk-significant components and operator actions for review using

information contained in the licensees probabilistic risk assessment. This included

components and operator actions that had a risk achievement worth factor greater

than 2 or Birnbaum value greater than 1E-6.

For the review of operator actions, the team observed operators during simulator

scenarios associated with the selected components as well as observing simulated

actions in the plant. The scenario was a Mode 1 full power Small Break Loss of Coolant

Accident (5250 gpm) Reactor Coolant System Loop C which results in cold leg

recirculation. The selected operator actions were:

• Manually trip reactor coolant pump 5 minutes from the time trip criteria is met

The team observed this task during the simulator scenario post trip and safety

injection actuated. The crew performed the task in accordance with EOP E-0,

Reactor Trip or Safety Injection, Revision 18. The team observed this activity being

performed by a crew of operators. This activity was satisfactorily performed within

the required time.

• Align component cooling water to residual heat removal heat exchangers 11 minutes

from event initiation

The team observed this task during the simulator scenario post trip and safety

injection actuated. The crew performed the task in accordance with EOP E-0,

Reactor Trip or Safety Injection, Revision 18. The team observed this activity being

performed by a crew of operators. This activity was satisfactorily performed within

the required time.

• Complete switchover of emergency core cooling system from injection mode to cold-

leg recirculation mode 8 minutes 20 seconds after the low-low 1 level is reached in

the reactor water storage tank

The team observed this task during the simulator scenario post trip and safety

injection actuated. The crew performed the task in accordance with EOP E-0,

Reactor Trip or Safety Injection, Revision 18, EOP E-1, Loss of Reactor or

Secondary Coolant, Revision 18, and EOP ES-1.3, Transfer to Cold Leg

Recirculation, Revision 12. The team observed this activity being performed by a

crew of operators. This activity was satisfactorily performed within the required time.

• Align containment spray for recirculation 3 minutes after low-low 2 level is reached in

the reactor water storage tank

The team observed this task during a simulator Job Performance

Measure URO-SEN-04-C193J(A)(TC). Two operators performed the task in

accordance with EOP ES-1.3, Transfer to Cold Leg Recirculation, Revision 12.

One operator did not complete the task and the other operator successfully

21

completed the task. This activity was satisfactorily performed by the station within

the required time. The station wrote a condition report and remediated the failed

operator.

The team reviewed the records for the last three years on these tasks to determine if

their program described in APA-ZZ-00395, Significant Operator Response Timing,

Revision 27, was being performed as required and documented in accordance with the

procedure. All of the records reviewed were in compliance with the program procedure

requirements.

b. Findings

No findings were identified.

4OA6 Meetings, Including Exit

On August 4, 2017, the NRC team discussed the preliminary results of this inspection with

Mr. T. Herrmann, Site Vice President, and other members of your staff. On August 28, 2017,

the NRC team discussed the final results of this inspection with Ms. S. Kovaleski, Director,

Design Engineering, and other members of your staff. The licensee acknowledged the issues

presented. The licensee confirmed that any proprietary information reviewed by the inspectors

had been returned or destroyed.

4OA7 Licensee Identified Violation

The following violation of very low safety significance (Green) was identified by the licensee and

is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for

being dispositioned as a licensee-identified, non-cited violation.

Technical Specification 5.4.1.a requires, in part, that written procedures shall be

established, implemented, and maintained covering the applicable procedures

recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.

Section 8 of Regulatory Guide 1.33, Revision 2, Appendix A, Procedures for Control

of Measuring and Test Equipment and for Surveillance Tests, Procedures, and

Calibrations, Part b, requires, in part, that specific procedures for surveillance tests,

inspections, and calibrations, should be written (implementing procedures are required

for each surveillance test, inspection, or calibration, listed in the technical specifications).

Station Procedure EDP-ZZ-01114, Motor Operated Valve Program Guide,

Revision 034, Section 3.6.3.b, requires, in part, that the motor-operated valve engineer

document a signature analysis report within 60 days following a diagnostic test of motor

operated valves. Contrary to the above, on July 17, 2016, the motor-operated valve

engineer failed to generate a signature analysis report within 60 days following a recent

diagnostic test of a motor-operated valve. Specifically, in May 2014, the NRC inspection

team identified NCV 05000483/2014007-06, Failure to Review Motor Operated Valve

(MOV) Data and Complete Analysis of the Data in a Timely Manner. This finding

was entered into the licensee's corrective action program as Callaway Action

Requests CARs 201402987 and 201402992.

During Refueling Outage RF21 (spring of 2016), 33 motor operated valves had been

tested and should have had a signature analysis report completed by the end of June

2016. On July 17, 2016, the licensee personnel recognized that they had not completed

22

the signature analysis report for 31 of the 33 valves tested. The team evaluated the

significance of the issue under the Significance Determination Process, as defined in

Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and 0609

Appendix A, The Significance Determination Process (SDP) for Findings at-Power,

dated June 19, 2012. The team concluded the finding was of very low safety

significance (Green) because all questions in Exhibit 2 could be answered no.

The licensee entered this issue into their corrective action program as Condition

Report CR-201606143.

23

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

S. Abel, Director, Engineering Projects

R. Andreasen, Engineer, Design Engineering

S. Banker, Senior Director, Engineering

B. Bax, Consulting Engineer, Design Engineering

S. Beck, Technician, Operations

E. Berry, Technician, Operations

F. Bianco, Director, Nuclear Operations

L. Bland, Supervisor, Operations

J. Bock, Transformer Engineer, Systems Engineering

J. Bruemmer, Electrical System Engineer, Engineering Systems

J. Copeland, Supervisor, Operations

J. Cortez, Director, Training

M. Covey, Manager, Operations Support

B. Cox, Senior Director, Nuclear Operations

J. Czeschin, Shift Manager, Operations

R. Davis, Career Engineer, Engineering Programs

M. Dunbar, Acting Director, Maintenance

J. Easley, Technician, Operations

L. Eitel, Supervisor, Engineering Design

T. Elwood, Supervisor - Engineer, Regulatory Affairs and Licensing

S. Ewens, Engineer, Engineering Projects

D. Farnsworth, Director, Work Management

C. Farrow, Supervisor, Operations

M. Haag, Senior Electrical Engineer, Design Engineering

T. Herrmann, Site Vice President

S. Kovaleski, Director, Engineering Design

J. Little, Consulting Engineer, Regulatory Affairs and Licensing

B. Long, Shift Manager, Operations

D. Martin, Senior Electrical System Engineer, Engineering Systems

M. Otten, Manager, Operations Training

R. Pohlman, Engineer, Regulatory Affairs

B. Price, Supervisor, Operations

J. Raithel, Engineer, Engineering Projects

J. Sellers, Supervising Engineer, EFIN Support

M. Sellers, Licensed Supervisor, Operations

S. Slayden, Electrical Engineer, Design Engineering

R. Tiefenauer, Senior Training Supervisor, Operations

B. Wentz, Breaker Engineer, System Engineering

L. Wilhelm, Supervisor, Operations

NRC Personnel

D. Bradley, Senior Resident Inspector

Attachment

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000483/2017007-01 NOV Not Verifying the Operation and Timing of the Engineered

Safety Feature Transformer XNB01 Load Tap Changer

(Section 1R21.2.4)

Opened and Closed

05000483/2017007-02 NCV Safety Injection Piggyback Valve EJ-HV-8804A Valve

Interlocks Not Tested (Section 1R21.2.8.b.1)05000483/2017007-03 NCV Inputs to Internal Flooding Calculations Not Translated into

Procedures or Instructions (Section 1R21.2.8.b.2)

A-2

LIST OF DOCUMENTS REVIEWED

Calculations

Number Title Revision

AL-16 Determine Available NPSH for the Auxiliary Feedwater 0

Pumps

AL-24 Determine the Effects on Available NPSH for the Aux 0

Feedwater Pumps

AL-24 Determine the Effect of Dissolved Nitrogen on the 0

NPSHA for AL Pumps

Ang-23 MCC Setpoint Calculation 0

BN-16 RWST Drain-down During Transfer to Cold Leg 1

Recirculation

DA-03 Condenser Pit Sizing for CWS Pipe Break Given in 000

FSAR 3B.4.3

DA-03 Condenser Pit Sizing For CWS Pipe Break Given In 0

FSAR 3B.4.3

E-21024 Relay Setting Tabulation & Coordination Curves - NG 9

E-B-09 DC Control Circuit Voltage Drops 1

EJ-22 Calculate Heat Transfer from RHR Pump Casing and 1

Suction Pipe with Insulation Removed

EJ-29, RHR NPSH Margin in Recirculation 2

Appendix A

HV-288 Loss of Ventilation 0

KC-139 Fire Area C-28 Control Room Service Area 001

M-AL-33 Impact of MP 05-3025 Maximum Allowed Temperature of 0

CST and AFW System

M-EF-52 Heat Exchanger Performance Based on Reduced ESW 1

Temperature and Flow

M-FL-02 Determine Flood Levels in Auxiliary Building Rooms 001

1107, 1108, 1109, 1110, 1111, 1112, 1113, and 1114

M-FL-02 Determine Flood Levels in Auxiliary Building Rooms 1

1107, 1108, 1109, 1110, 1111, 1112, 1113, and 1114

A-3

Calculations

Number Title Revision

M-FL-10 Maximum Flood Level for Rooms in the Diesel Generator 002

Building

M-GL-390 Callaway Auxiliary Building HVAC 1

M-HV-288 Loss of Cooling Analysis 0

NB-05 LTC CONTROLLER 4

NG-12 NG MCC Setpoint Calculation 3

NG-22 NG Load Center Overcurrent Setpoint Calculation 1

NG-23 MCC Setpoint Calculation 0

PK-01 PK-11 and PK-12 Battery and Charger Sizing 0

ZZ-12 NG MCC Setpoint Calculation 3

ZZ-145 Short Circuit Calculation 2

ZZ-214 MOV Voltage Drop Calculation 10

ZZ-214 MOV Voltage Drop Calculation 10

ZZ-214 MOV Voltage Drop Calculation 11

ZZ-428 Control Room Dose Calculation 3

ZZ-534 Quarter-Turn MOV Capability and Margin Calculation 1

ZZ-536 Rising Stem MOV Capability and Margin Calculation 1

ZZ-536 Rising-Stem MOV Capability and Margin Calculation 1

ZZ-62 Plant Load Flow Calculation 9

Condition Reports

200202970 201309622 201506417 201609404

200402026 201402827 201507559 201700397

200507878 201402987 201508240 201701520

200600033 201402992 201601675 201701534

A-4

Condition Reports

200605158 201403130 201603215 201701537

200605786 201403366 201604022 201701737

201201245 201403369 201606143 201702402

201201652 201404740 201607559 201702470

201203786 201405312 201607748 201702949

201205441 201405319 201607971 201703566

201302829 201502100 201608145 201703576

Condition Reports Generated During this Inspection

201703576 201703831 201703963 201703979

201703604 201703833 201703971 201703981

201703694 201703931 201703972 201703982

201703700 201703948 201703976 201703992

201703707 201703959 201703978 201720349

201703720 201703962

Design Basis Documents

Number Title Revision

ULDBD-BN-001 Borated Refueling Water Storage System 1

ULDBD-EF-001 Essential Service Water 2

ULDBD-EJ-001 Residual Heat Removal 1

ULDBD-EJ-001 Residual Heat Removal 1

ULDBD-FLOOD- Topical Area - Internal Flooding 1

001

Drawings

Number Title Revision

74940 Valve Assembly - Lift Check D

75010 Valve Assembly - Lift Check F

A-5

Drawings

Number Title Revision

8809D51 Pressurizer Pressure Control Interconnecting Wiring 1

Sheet 28 Diagram Cabinet 1 Card Frame 05

8809D55 Pressurizer Pressure Control Interconnecting Wiring 0

Sheet 38 Diagram Cabinet 5 Card Frame 05

8809D55-S027 Pressurizer Pressure Interconnecting Wiring Diagram 15

Cabinet 1 Card Frame 05

8809D55-S037 Pressurizer Level Control Pressurizer Pressure Control 8

Interconnecting Wiring Diagram Cabinet 5 Card Frame

05

8809D55-S040 Pressurizer Pressure Control Interconnecting Wiring 14

Diagram Cabinet 5 Card Frame 05

8809D56-S032 Pressurizer Pressure Control Interconnecting Wiring 12

Diagram Cabinet 6 Card Frame 04

E-018-00112 Motor Control Center Layout 27

E-018-00113 Motor Control Center Layout 28

E-018-00114 Motor Control Center Layout 16

E-018-00115 Motor Control Center Layout 31

E-050-00006 Layout for 60 Cell NCN-23 Batteries 13

E-21001 Main Single Line Drawing 25

E-21001 Main Single Line Diagram 25

E-21010 DC Main Single Line Diagram 14

E-21010A DC Main Single Line Diagram (PK03, PK04, and PK05) 9

E-21NB01 MV System Class 1E 4.16KV Single Line Diagram 9

E-21NB01 Lower Medium Voltage System Class 1E 4.16KV Single 9

Line Meter and Relay Diagram

E-21NB02 Lower Medium Voltage System Class 1E 4.16KV Single 14

Line Meter and Relay Diagram

E-21NG01 LV System Class 1E Single Line Diagrams 28

A-6

Drawings

Number Title Revision

E-21NG01 Low Medium Voltage System Class 1E 4.16KV Single 28

Line Meter and Relay Diagram

E-21NG02 Low Medium Voltage System Class 1E 4.16KV Single 33

Line Meter and Relay Diagram

E-21NK01 Class 1E 125V DC System Meter and Relay Diagram 11

E-21NK02 Class 1E 125V DC System Meter and Relay Diagram 11

E-23BB39 Schematic Diagram Pressurizer Relief Isolation Valve 15

E-23BB39A Schematic Diagram Pressurizer Relief Isolation Valve 1

E-23BB40 Schematic Diagram Pressurizer Relief Valves 3

E-23NN01 Class 1E Instrument AC Schematic 11

E-27000A Sht. 5 Termination of Selected Class 1E Devices with Pigtails 23

M-22BB02 Piping and Instrumentation Diagram Reactor Coolant 33

System

M-22BG03 Chemical and Volume Control System 56

M-22BG03 Piping and Instrumentation Diagram Chemical and 56

Volume Control System

M-22BG03(Q) P&ID Chemical and Volume Control System 57

M-22BN01 Borated Refueling Water Storage System 26

M-22BN01(Q) P&ID Borated Refueling Water Storage System 26

M-22EF01(Q) P&ID Essential Service Water System 80

M-22EF02 Piping and Instrumentation Diagram Essential Service 75

Water System

M-22EJ01 Residual Heat Removal System 62

M-22EJ01 Residual Heat Removal System 62

M-22EJ01(Q) P&ID Residual Heat Removal System 62

M-22EM01 High Pressure Coolant Injection System 38

A-7

Drawings

Number Title Revision

M-22EM01 Piping and Instrumentation Diagram High Pressure 38

Coolant Injection System

M-22EM01(Q) P&ID High Pressure Coolant Injection System 39

M-22EM02 High Pressure Coolant Injection System 23

M-22EM03 High Pressure Coolant Injection System Test Line 13

M-22EN01(Q) P&ID Containment Spray System 16

M-22GL01(Q) P&ID Auxiliary Building HVAC 34

M-23EJ01 Residual Heat Removal System Auxiliary Building A 21

Train

M-23EJ03 Piping Isometric Residual Heat Removal System 9

Auxiliary Building A&B Train

M-246-00003 Stainless Steel Swing Check Valves 9

M-2G051 Control & DG Buildings and Comm. Corridor el 2000 & 42

2016

MS-01 Piping Class Summary Table of Contest and Revision 96

Description

S-1027-00026 1/2-inch Connector Assembly Bayonet Barton Style P/N 0

880701-2-18-BPEBT3F

Engineering Evaluations

Number Title Revision

201703576 Power-Operated Relief Valve Block Valve Torque and 1

Thrust Evaluations.

J-301- Plant Qualification Evaluation for Rosemount Model 1154 1

000097P01 Series H Pressure Transmitters for Nuclear Service

(Inside Containment, Specification J-301).

S-1027- Plant Qualification Evaluation for EGS 1/2-inch Style 11

00013P01 880701 and 3/4-inch Style 913601 Quick Disconnect

Electrical Connector

A-8

Miscellaneous

Number Title Revision

Date

Engineering Disposition Pressurizer Pressure 1

Transmitter Replacement MP 08-0054

Inservice Testing Program 32

Pumping Temperature Limits, Seal Flush Tap Flowserve June 1, 2006

(I-R) Model 4HMTA-9 Auxiliary Feed Water Pump

13004296.500 Replace Tobar Transmitter BBPT0457 RCS Pressurizer 6

Pressure

13004296.500 Completed Work Order for Replacement of Rosemount July 2, 2013

Model 1154 Series H Pressure Transmitter BBPT0457

536879 Material Request/Order - Cross Reference Listing for 1

Material Request Item 7606153 Electrical Connector

Quick Disconnect

7606153 Material Procurement, Connector, Electrical, Quick 0

Disconnect, 1/2 in. OD

AP06-006 QA Audit of Design Control Component Design Basis July 31, 2006

APA-ZZ-00390 Environmental and Seismic Qualification of Safety- 27

Related Equipment

E-1052 Technical Specification - Replacement Breakers for MV 0

Metal Clad Switchgear and LV Load Centers

EJ-HV-88O4A MOV Predictive Performance Report May 7, 2010

IN 2012-003 Open Phase Voltage Protection

MP 05-3025 Maximum Allowed Temperature of CST and AFW August 4,

System 2012

MP 08-0054 LDCN 13-0011 Applicability Determination Form CA- 0

2510

MPE-ZZ-QS015 4.16KV Square D Magnum Breaker PM 7

MSE-ZZ-QS006 NLI/Square D Masterpact Circuit Breaker PM and 4

Inspection

N/A Inservice Testing Program 32

N/A Flowserve Letter: Stem-Wedge Separation of Anchor- July 11, 2017

Darling Gate Valve

A-9

Miscellaneous

Number Title Revision

Date

QR-06513544-2 Qualification Report for SQ D MV Replacement Circuit 0

Breakers

QRLV- Qualification Report for LGSB13/LGSB2 Circuit 0

06513544-1 Breakers

RFR 07896 RHR Flowrates in Various Modes A

RFR 15112A Approved Storage Locations in Control Building September 30,

1994

RFR 15219 Maximum Allowed Temperature of CST And AFW September 22,

System 2003

RFR 8746 Special Test Report for ETP-EF-ST017 L

RFR 8746 I Verify the Thrust Required for MOVs using Grouping January 24,

1994

UOMNE 91-302 Position on NRC Information Notice 91-56 November 1,

1991

WCAP-11992 ATWS Rule Administration Process December

1988

WCAP-15831-P- WOG Risk-Informed ATWS Assessment and Licensing August, 2017

A Implementation Process

WCAP-8330 Anticipated Transient Without SCRAM Analysis August, 1974

Modifications

Number Title Revision

MP 07-0069 Replace 480V Load Center Breaker Replacement 2

MP 07-0070 Replace Safety Related Metal Clad Breakers 4

MP 08-0054 Pressurizer Pressure Transmitter Replacement 1

MP 10-0009 Installation of New Westinghouse RCP Shutdown Seals 1

MP 15-0008 Open Phase Condition Protection 1

A-10

Procedures

Number Title Revision

APA-ZZ-00395 Significant Operator Response Timing 027

APA-ZZ-00395 Significant Operator Response Timing 27

APA-ZZ-00750 Hazard Barrier Program 039

APA-ZZ-00801 Foreign Material Exclusion 044

BD-ES-1.3 Transfer to Cold Leg Recirculation 8

CTM-EXAM Examination Control 005

CTM-NAP Appendix 2 CEC Orientation Manual 003

E-0 Reactor Trip Or Safety Injection 14

E-1 Loss of Reactor or Secondary Coolant 14

EDP-ZZ-04023 Calculations 043

EOP E-0 Reactor Trip or Safety Injection 018

EOP E-1 Loss of Reactor or Secondary Coolant 018

EOP ES-1.3 Transfer to Cold Leg Recirculation 012

ES-1.3 Transfer to Cold Leg Recirculation 10

ESP-ZZ-00356 Technical Specification 5.5.2.B Verification Integrated 6

Leak Rate Requirements for Primary Coolant Sources

Outside Containment

ISL-BB-0P455 Loop-Pressure: RX Pressurizer Pressure - Protection Set 32

I

ISL-BB-0P456 Loop-Pressure: RX Pressurizer Pressure - Protection Set 32

II

ISL-BB-0P457 Loop-Pressure: RX Pressurizer Pressure - Protection Set 33

III

ISL-BB-0P458 Loop-Pressure: RX Pressurizer Pressure - Protection Set 34

IV

ISL-NN-0P458 Loop-Pressure: RX Pressure - Protection Set II 33

ISL-NN-0P458 Loop-Pressure: RX Pressure - Protection Set IV 34

A-11

Procedures

Number Title Revision

ISP-BB-00001 Pressurizer Pressure Sensor Time Response Test 4

ISP-BB-00001 Pressurizer Pressure Sensor Time Response Report 4

ITG-ZZ-RM003 Generic-Pressure; Rosemount 1150 and 3150 Series 9

Transmitters

MDP-ZZ-0STOR Staging and Storage of Materials, Equipment & Tools at 023

the Callaway Energy Center

MDP-ZZ-LM001 Fluid Leak Management Program 016

MDP-ZZ-S0001 Scaffolding Installation and Evaluation 037

MDP-ZZ-S0001 Scaffolding Installation and Evaluation 37

MTT-ZZ-I004A Raychem Heat Shrink Installation 6

OSP-BB-V002A Power-Operated Relief Valve Inservice -Test 13

OSP-BN-V0003 BNHV8813 Inservice Test 6

OSP-BN-V0004 BN8717 Inservice Test 6

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

OSP-EF-V001A ESW Train A Valve Operability 45

OSP-EJ-P001A RHR Train A Inservice Test - Group A 63

OSP-EJ-PV04A Train A RHR and RCS Check Valve Inservice Test 14

OSP-EJ-V003A RHR Train A Mode 5 Valve Inservice Test 16

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

Test

OSP-NE-0001A Standby Diesel Generator A Periodic Tests 63

OSP-SA-2413A Train A Diesel Generator and Sequencer Testing 024

OTN-AP-0001 Condensate Transfer and Storage System 14

OTN-EF-00001 Essential Service Water System 74

OTN-EP-00001 Accumulator Safety Injection System 27

A-12

Procedures

Number Title Revision

OTN-EP-00001 SI Accumulator Level Control 7

Addendum 1

OTN-KC-00001 Manual Operation of Electric Fire Pump 9

Addendum 13

OTO-ZZ-00005 Flooding 2

Screens

Number Title Revision

MP 08-0054 LDCN 13-0011 50.59 Screen Form CA-2511 0

System Health Reports

Number Title Date

MD 4Q16 - EHV Switchyard Bus March 20,

2017

MR 4Q16 - SU/RES Auxiliary Transformers March 20,

2017

NB 4Q16 - Low MV System 1E March 20,

2017

NG 4Q16 - Low Voltage System 1E March 20,

2017

Residual Heat Removal Performance Monitoring Report Quarter 1

2014

Residual Heat Removal Performance Monitoring Report Quarter 2

2014

Residual Heat Removal Performance Monitoring Report Quarter 3

2014

Residual Heat Removal Performance Monitoring Report Quarter 4

2014

Residual Heat Removal Performance Monitoring Report Quarter 1

2015

Residual Heat Removal Performance Monitoring Report Quarter 2

2015

A-13

System Health Reports

Number Title Date

Residual Heat Removal Performance Monitoring Report Quarter 3

2015

Residual Heat Removal Performance Monitoring Report Quarter 4

2015

Residual Heat Removal Performance Monitoring Report Quarter 1

2016

Residual Heat Removal Performance Monitoring Report Quarter 2

2016

Residual Heat Removal Performance Monitoring Report Quarter 3

2016

Residual Heat Removal Performance Monitoring Report Quarter 4

2016

Vendor Documents

Number Title Revision

00813-0100- Rosemount 1154 Series H Alphaline Nuclear Pressure BA

4631 Transmitter Product Data Sheet

00813-0100- Rosemount 1154 Series H Alphaline Nuclear Pressure BA

4631 Transmitter Reference Manual

10466-M-771-- Instruction Manual for Tobar Model 32PA1 Absolute 1

0337 Pressure Transmitter

E-1044-00001 Reinhausen Instruction Manual for Load Tap Changer 3

E-1052-00017 Instruction Manual for LBSB13/LGSB2 Circuit Breakers 0

E-1052-00031 Instruction Manual for Medium Voltage Circuit Breakers 0

M-1145-00002 Operating & Maintenance Instruction and Parts Catalog 0

For Anchor/Darling 30 Butterfly Valves

M-724-00634 Garrett Instruction Manual 3750014 for Power-Operated 4

Relief Valve

A-14

Work Orders

CALC-00002323 08510933.510 13006121 14511460.500

CALC-00003105 13004296.500 13506258.500 14512497.500

A-15

ML17283A392

SUNSI Review ADAMS: Non-Publicly Available Non-Sensitive Keyword:

By: TFarnholtz Yes No Publicly Available Sensitive NRC-002

OFFICE SRI:EB1 RI:PSB1 RI:EB2 SRTI:TTC SES:ORA

NAME RKopriva IAnchondo JWatkins APalmer JKramer

SIGNATURE /RA/ /RA/ /RA/ /RA/ /RA/

DATE 9/14/17 9/29/17 8/30/17 9/18/17 9/22/17

OFFICE C:PPB C:EB1

NAME NTaylor TFarnholtz

SIGNATURE /RA/ /RA/

DATE 10/5/17 10/6/17