ML15173A450
| ML15173A450 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 06/22/2015 |
| From: | Anton Vegel Division of Reactor Safety IV |
| To: | Limpias O Nebraska Public Power District (NPPD) |
| References | |
| EA-15-089 IR 2015007 | |
| Download: ML15173A450 (46) | |
See also: IR 05000298/2015007
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION IV
1600 E. LAMAR BLVD
ARLINGTON, TX 76011-4511
June 22, 2015
Mr. Oscar A. Limpias
Vice President-Nuclear and CNO
Nebraska Public Power District
Cooper Nuclear Station
72676 648A Avenue
P.O. Box 98
Brownville, NE 68321
SUBJECT: COOPER NUCLEAR STATION - NRC COMPONENT DESIGN BASIS
INSPECTION REPORT NO. 05000298/2015007 AND NOTICE OF VIOLATION
Dear Mr. Limpias:
On May 8, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at
the Cooper Nuclear Station. The NRC inspectors discussed the results of this inspection with
you and other members of your staff. Inspectors documented the results of this inspection in
the enclosed inspection report.
During this inspection, the NRC staff examined activities conducted under your license as they
relate to public health and safety to confirm compliance with the Commission's rules and
regulations and with the conditions of your license. Within these areas, the inspection consisted
of selected examination of procedures and representative records, observations of activities,
and interviews with personnel.
Based on the results of this inspection, the NRC has determined that a cited violation is
associated with this inspection. The violation is being cited because Cooper Nuclear Station
failed to restore compliance with NRC requirements within a reasonable time after a previous
violation was identified in NRC Inspection Report 05000298/2010007 (issued December 3,
2010). This is consistent with the NRC Enforcement Policy; Section 2.3.2.a, which states, in
part, that a cited violation will be considered if the licensee fails to restore compliance within a
reasonable time after a violation is identified.
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.
O. Limpias -2-
In addition, the NRC has determined that a Severity Level IV violation of NRC requirements
occurred. The NRC also identified two additional issues that were evaluated under the risk
significance determination process as having very low safety significance (Green). The NRC
has also determined that violations are associated with these issues. These violations are
being treated as non-cited violations (NCVs), consistent with Section 2.3.2.a of the NRC
Enforcement Policy. These NCVs are described in the subject inspection report.
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: (1) the Regional Administrator, Region IV; (2) the Director, Office of Enforcement,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and (3) the NRC Senior
Resident Inspector at the Cooper Nuclear Station.
If you disagree with a cross-cutting aspect assignment or a finding not associated with a
regulatory requirement in this report, you should provide a response within 30 days of the date
of this inspection report, with the basis for your disagreement, to the Regional Administrator,
Region IV; and the NRC Senior Resident Inspector at the Cooper Nuclear Station.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its
enclosures, and your response, will be made available electronically for public inspection in the
NRC Public Document Room or from the NRC's Agencywide Documents Access and
Management System (ADAMS), accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html. To the extent possible, your response should not
include any personal privacy or proprietary information so that it can be made available to the
public without redaction.
Sincerely,
/RA/
Anton Vegel, Director
Division of Reactor Safety
Docket No. 50-298
License No. DPR-46
Enclosures:
1. Notice of Violation
2. Inspection Report No. 05000298/2015007
w/Attachment: Supplemental Information
cc: Electronic Distribution
x SUNSI Review ADAMS x Publicly Available x Non-Sensitive Keyword:
By: WCS x Yes No Non-Publicly Available Sensitive NRC-002
OFFICE SRI:EB1 RI:EB2 TTC SRI:EB1 BC:EB1 BC:DRP/C ACES D:DRS
NAME WSifre NOkonkwo EEmrich RLatta TFarnholtz GWarnick MHay AVegel
SIGNATURE /RA/ /RA/ /RA/ /RA/ /RA/ RVA for JMR for /RA/
DATE 6/17/15 6/17/15 6/16/15 6/17/15 6/19/15 6/17/15 6/17/15 6/22/15
Letter to Oscar A. Limpias from Anton Vegel, dated June 22, 2015
SUBJECT: COOPER NUCLEAR STATION - NRC COMPONENT DESIGN BASIS
INSPECTION REPORT NO. 05000298/2015007 AND NOTICE OF VIOLATION
Electronic distribution by RIV:
Regional Administrator (Marc.Dapas@nrc.gov)
Deputy Regional Administrator (Kriss.Kennedy@nrc.gov)
DRP Director (Troy.Pruett@nrc.gov)
DRP Deputy Director (Ryan.Lantz@nrc.gov)
DRS Director (Anton.Vegel@nrc.gov)
DRS Deputy Director (Jeff.Clark@nrc.gov)
Acting Senior Resident Inspector (James.Nance@nrc.gov)
Branch Chief, DRP/C (Greg.Warnick@nrc.gov)
Senior Project Engineer (Ray.Azua@nrc.gov)
Project Engineer (Paul.Nizov@nrc.gov)
Project Engineer (Michael.Stafford@nrc.gov)
Administrative Assistant (Amy.Elam@nrc.gov)
Public Affairs Officer (Victor.Dricks@nrc.gov)
Affairs Officer (Lara.Uselding@nrc.gov)
Project Manager (Siva.Lingam@nrc.gov)
Team Leader, DRS/TSST (Don.Allen@nrc.gov
ACES (R4Enforcement.Resource@nrc.gov)
RITS Coordinator (Marisa.Herrera@nrc.gov)
Regional Counsel (Karla.Fuller@nrc.gov)
Technical Support Assistant (Loretta.Williams@nrc.gov)
Congressional Affairs Officer (Jenny.Weil@nrc.gov)
Congressional Affairs Officer (Angel.Moreno@nrc.gov)
RIV/ETA: OEDO (Michael.Waters@nrc.gov)
ROPreports
NOTICE OF VIOLATION
Nebraska Public Power District Docket No. 50-298
Cooper Nuclear Station License No. DPR-46
During an NRC inspection conducted April 6 through May 8, 2015, a violation of NRC
requirements was identified. In accordance with the NRC Enforcement Policy, the violation is
listed below:
Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that
measures shall be established to assure that applicable regulatory requirements and the
design basis, as defined in 10 CFR 50.2 and as specified in the license application, for
those components to which this appendix applies, are correctly translated into
specifications, drawings, procedures, and instructions.
Contrary to the above, since July 2010 the licensee failed to assure that applicable
regulatory requirements and the design basis were correctly translated into
specifications, drawings, procedures, and instructions. Specifically, the licensee failed to
correctly translate regulatory and design basis requirements associated with tornado and
high wind-generated missiles into design information necessary to protect the
emergency diesel generator fuel oil storage tank vent line components.
This violation is associated with a Green Significance Determination Process finding.
Pursuant to the provisions of 10 CFR 2.201, Nebraska Public Power District is hereby required
to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission,
ATTN: Document Control Desk, Washington, DC 20555-0001 with a copy to the Regional
Administrator, Region IV, and a copy to the NRC Senior Resident Inspector at the facility that is
the subject of this Notice, within 30 days of the date of the letter transmitting this Notice of
Violation (Notice). This reply should be clearly marked as a "Reply to a Notice of Violation;
EA-15-089" and should include for each violation: (1) the reason for the violation, or, if
contested, the basis for disputing the violation or severity level, (2) the corrective steps that
have been taken and the results achieved, (3) the corrective steps that will be taken, and (4) the
date when full compliance will be achieved. Your response may reference or include previous
docketed correspondence if the correspondence adequately addresses the required response.
If an adequate reply is not received within the time specified in this Notice, an order or a
Demand for Information may be issued as to why the license should not be modified,
suspended, or revoked, or why such other action as may be proper should not be taken. Where
good cause is shown, consideration will be given to extending the response time.
If you contest this enforcement action, you should also provide a copy of your response, with
the basis for your denial, to the Director, Office of Enforcement, United States Nuclear
Regulatory Commission, Washington, DC 20555-0001.
Because your response will be made available electronically for public inspection in the NRC
Public Document Room or from the NRCs ADAMS, accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html, to the extent possible, it should not include any
Enclosure 1
personal privacy, proprietary, or safeguards information so that it can be made available to the
public without redaction. If personal privacy or proprietary information is necessary to provide
an acceptable response, then please provide a bracketed copy of your response that identifies
the information that should be protected and a redacted copy of your response that deletes such
information. If you request withholding of such material, you must specifically identify the
portions of your response that you seek to have withheld and provide in detail the bases for your
claim of withholding (e.g., explain why the disclosure of information will create an unwarranted
invasion of personal privacy or provide the information required by 10 CFR 2.390(b) to support a
request for withholding confidential commercial or financial information). If safeguards
information is necessary to provide an acceptable response, please provide the level of
protection described in 10 CFR 73.21.
Dated this 22nd day of June 2015
-2- Enclosure 1
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket: 05000298
License: DPR-46
Report No.: 05000298/2015007
Licensee: Nebraska Public Power District
Facility: Cooper Nuclear Station
Location: P.O. Box 98
Brownville, NE 68321-0098
Dates: April 6 - May 8, 2015
Team Leader: W. Sifre, Senior Reactor Inspector, Engineering Branch 1
Inspectors: R. Latta, Senior Reactor Inspector, Engineering Branch 1
N. Okonkwo, Reactor Inspector, Engineering Branch 2
M. Emrich, Senior Reactor Technology Instructor, Technical Training
Center
Accompanying C. Barron, Contractor, Beckman and Associates
Personnel: S. Kobylarz, Contractor, Beckman and Associates
Approved By: Thomas R. Farnholtz
Branch Chief, Engineering Branch 1
Division of Reactor Safety
Enclosure 2
SUMMARY
IR 05000298/2015007; 04/06/2015 - 05/08/2015; Cooper Nuclear Station; Component Design
Basis Inspection.
The inspection activities described in this report were performed between April 6, 2015, and
May 8, 2015, by three inspectors from the NRCs Region IV office, one instructor from the
NRCs Technical Training Center, and two contractors. Four findings of very low safety
significance (Green) are documented in this report. Three of these findings involved violations
of NRC requirements and one of these violations was determined to be Severity Level IV under
the traditional enforcement process. The significance of inspection findings is indicated by their
color (Green, White, Yellow, or Red), which is determined using Inspection Manual
Chapter 0609, Significance Determination Process. Their cross-cutting aspects are
determined using Inspection Manual Chapter 0310, Aspects Within the Cross-Cutting Areas.
Violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement
Policy. The NRCs program for overseeing the safe operation of commercial nuclear power
reactors is described in NUREG-1649, Reactor Oversight Process.
Cornerstone: Initiating Events
- Green. The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B,
Criterion V, Instructions, Procedures, and Drawings, which states, in part, Activities
affecting quality shall be prescribed by documented procedures of a type appropriate to the
circumstances and shall be accomplished in accordance with these procedures.
Specifically, prior to April 6, 2015, the licensee failed to follow Procedure .05.OPS,
Operations Review of Condition Reports/Operability Determination, to ensure that an
operability review was performed for Condition Report CR-CNS-2015-01268, which was
initiated during the self-audit for the Component Design Bases Inspection to document that
Cooper Nuclear Station has under-voltage relays that could be affected by harmonics. In
response to this issue, the licensee performed an operability review and an operability
evaluation for the under-voltage relays. This finding was entered into the licensees
corrective action program as Condition Report CR-CNS-2015-02337.
The team determined that failure to perform an operability review associated with Condition
Report CR-CNS-2015-01268 was a performance deficiency. This finding was more than
minor because it was associated with the human performance attribute of the Initiating
Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of
events that upset plant stability and challenge critical safety functions during shutdown, as
well as power operations. Specifically, the licensee failed to perform the required operability
review for the identified condition. In accordance with Inspection Manual Chapter 0609,
Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated
June 19, 2012, Exhibit 1, Initiating Event Screening Questions, the issue screened as
having very low safety significance (Green) because the finding did not cause a reactor trip
and it did not involve the loss of mitigation equipment. This finding had a cross-cutting
aspect in the area of human performance associated with teamwork because individuals
and work groups failed to communicate and coordinate their activities across organizational
boundaries to ensure nuclear safety is maintained [H.4]. (Section 1R21.2.7)
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Cornerstone: Mitigating Systems
- Green. The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B,
Criterion III, Design Control, which states, in part, that design control measures shall
provide for verifying or checking the adequacy of design, such as by the performance of
design reviews, by the use of alternate or simplified calculational methods, or by the
performance of a suitable testing program. Specifically, prior to April 6, 2015, the licensee
failed to maintain procedure changes to periodically monitor and add nitrogen to fire
protection system headers in the reactor building to mitigate the effects of water hammer. In
response to this issue, the licensee determined that the fire protection system remained
functional without nitrogen based on empirical evidence suggesting that the system was
capable of absorbing the shockwave from a water hammer event. This finding was entered
into the licensees corrective action program as Condition Report CR-CNS-2015-02085.
The team determined that the failure to adequately maintain control of the fire protection
system design to prevent water hammer events was a performance deficiency. This finding
was more than minor because it was associated with the design control attribute of the
Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure
the availability, reliability, and capability of systems that respond to events to prevent
undesirable consequences. Specifically, the licensee failed to maintain procedure changes
to periodically monitor and add nitrogen to fire protection system headers in the reactor
building. 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 1R21.3.1)
- SL IV. The team identified three examples of a Severity Level IV, non-cited violation, of
10 CFR 50.71, Maintenance of Records, Making of Reports, Section (e), which states,
in part, each person licensed to operate a nuclear power reactor under the provisions of
10 CFR 50.21 or 10 CFR 50.22 shall update periodically the final safety analysis report
(FSAR) originally submitted as part of the application for the license, to assure that the
information included in the report contains the latest information developed. This submittal
shall contain all the changes necessary to reflect information and analyses submitted to the
Commission by the licensee since the submittal of the original FSAR, or as appropriate, the
last update to the FSAR under this section. Specifically, in January 2012 and February
2015, the licensee failed to update the Updated Safety Analysis Report for changes made to
their Anticipated Transient Without Scram analyses and plant conduct of operations
procedures. This finding was entered into the licensees corrective action program as
Condition Reports CR-CNS-2015-02106, CR-CNS-2015-02090, and CR-CNS-2015-02393.
The team determined that the failure to update the Final Safety Analysis Report to assure
that the information included in the report contains the latest information developed was a
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performance deficiency. This finding was evaluated using traditional enforcement because
it had the potential for impacting the NRCs ability to perform its regulatory function. This
finding was more than minor because each example potentially rendered portions of the
safety analyses for Anticipated Transient Without Scram events described in the Updated
Safety Analysis Report less conservative or contradicted previous information regarding the
licensees flooding analysis contained in the Updated Safety Analysis Report. The
traditional enforcement violation was determined to be a Severity Level IV violation
consistent with the example in paragraph 6.1.d(3) of the NRC Enforcement Policy. Since
this was a traditional enforcement violation, no cross-cutting aspects were assigned per the
guidance contained in Inspection Manual Chapter 0612, Section 07.03(c). (Section 1R21.4)
- Green. The team identified a Green, cited violation of 10 CFR Part 50, Appendix B,
Criterion III, Design Control, which states, in part, Design control measures shall
provide for verifying or checking the adequacy of design, such as by the performance of
design reviews, by the use of alternate or simplified calculational methods, or by the
performance of a suitable testing program. Specifically, since July 2010 the licensee
failed to verify the adequacy of design of the vents for the emergency diesel generator 1
and 2 fuel oil storage tanks to withstand impact from a tornado driven missile hazard, or to
evaluate for exemption from missile protection requirements using an approved
methodology. This finding was entered into the licensees corrective action program as
Condition Report CR-CNS-2015-02366.
The team determined that the failure to evaluate the lack of missile protection on the
emergency diesel generator 1 and 2 fuel storage tank vents was a performance deficiency.
This finding was more than minor because it was associated with the design control attribute
of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to
ensure the availability, reliability, and capability of systems that respond to events to prevent
undesirable consequences. Specifically, the licensee failed to evaluate a design
nonconformance on the emergency diesel generator 1 and 2 fuel storage tanks for lack of
missile protection. 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, this finding 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. This finding had a cross-cutting aspect in the area of
human performance associated with conservative bias because individuals failed to use
decision making practices that emphasize prudent choices over those that are simply
allowable [H.14]. (Section 4OA2)
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REPORT DETAILS
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
This inspection of component design bases 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
basis functions. 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 Basis Inspection (71111.21)
.1 Overall Scope
To assess the ability of the Cooper Nuclear Station, equipment and operators to
perform their required safety functions, the team inspected risk-significant components
and the licensees responses to industry operating experience. The team selected
risk-significant components for review using information contained in the Cooper Nuclear
Station, probabilistic risk assessments and the U.S. Nuclear Regulatory Commissions
(NRC) standardized plant analysis risk model. In general, the selection process focused
on components that had a risk achievement worth factor greater than 1.3 or a risk
reduction worth factor greater than 1.005. The items selected included components in
both safety-related and nonsafety-related systems including pumps, circuit breakers,
heat exchangers, transformers, and valves. The team selected the risk-significant
operating experience to be inspected based on its collective past experience.
To verify that the selected components would function as required, the team reviewed
design basis assumptions, calculations, and procedures. In some instances, the team
performed calculations to independently verify the licensees conclusions. The team
also verified that the condition of the components was consistent with the design basis
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 selected components, 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 basis 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
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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;
Title 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 uniqueness and complexity of the design, operating experience, and the available
defense in-depth margins.
The inspection procedure requires a review of 15 to 25 total samples that include
risk-significant and low design margin components, components that affect the large
early release frequency (LERF), and operating experience issues. The sample selection
for this inspection was 16 components, 2 components that affect LERF, and 4 operating
experience items. The selected components and associated operating experience items
supported risk-significant functions including the following:
a. Electrical power to mitigation systems: The team selected several components in the
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,
station blackout, and a loss of coolant accident with offsite power available. As such, the
team selected:
- 125 Vdc Battery 1A
- 125 Vdc Charger 1A
- 125 Vdc Bus 1A
- Standby Liquid Control Pump Motor Protection
- Core Spray Pump Motor Protection
- 480 Vac Safety-Related Motor Control Center K
- 4160 Vac Safety-Related Switchgear 1F
- Startup Station Service Transformer
b. 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:
- Main Steam Isolation Valve MS-AOV-AO80C
- Suppression Chamber Spray A Inboard Throttle Valve RHR-MOV-MO38A
c. 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:
- Core Spray Pump CS-P-A
- Standby Liquid Control Pump SLC-P-A
- Residual Heat Removal Heat Exchanger B
- Residual Heat Removal Service Water Booster Pump 1C
- Residual Heat Removal Heat Exchanger B Bypass Valve
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- Residual Heat Removal Service Water Motor Operated Valve 89B
.2 Results of Detailed Reviews for Components
.2.1 125 Vdc Battery 1A
a. Inspection Scope
The team reviewed the updated safety analysis report, design basis documents, the
current system health report, selected drawings and calculations, maintenance and test
procedures, and condition reports associated with 125 Vdc Battery 1A. 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:
- Component maintenance history and corrective action program reports to verify
the monitoring of potential degradation.
- Calculations for electrical system load flow/voltage drop to verify that system
voltages remained within minimum acceptable limits.
- Calculations to verify design loading, input assumptions, and environmental
parameters are appropriate and that the battery cell is sized to perform the
battery design basis function in accordance with the technical specifications.
- Procedures for preventive maintenance, inspection, and testing to compare
maintenance practices against industry and vendor guidance.
- Results of completed surveillance testing in accordance with technical
specifications.
b. Findings
No findings were identified.
.2.2 125 Vdc Charger 1A
a. Inspection Scope
The team reviewed the updated safety analysis report, design basis documents, the
current system health report, selected drawings and calculations, maintenance and test
procedures, and condition reports associated with 125 Vdc Charger 1A. 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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- Component maintenance history and corrective action program reports to verify
the monitoring of potential degradation.
- Calculations for sizing to verify that charger capacity supported design basis
requirements in accordance with technical specifications.
- Calculations for the electrical protection to verify the charger protective devices
satisfied design basis requirements.
- Procedures for preventive maintenance, inspection, and testing to verify vendor
guidance and design requirements were adequately incorporated.
- Results of completed preventative maintenance and surveillance testing in
accordance with technical specifications.
b. Findings
No findings were identified.
.2.3 125 Vdc Bus 1A
a. Inspection Scope
The team reviewed the updated safety analysis report, design basis documents, the
current system health report, selected drawings and calculations, maintenance and test
procedures, and condition reports associated with 125 Vdc Bus 1A. 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:
- Component maintenance history and corrective action program reports to verify
the monitoring of potential degradation.
- 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.
- The protective device ratings to ensure adequate selective protection
coordination of connected equipment during worst-case short circuit conditions.
- Procedures for preventive maintenance, inspection, and testing to verify vendor
guidance and design requirements were adequately incorporated.
b. Findings
No findings were identified.
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.2.4 Standby Liquid Control Pump Motor Protection
a. Inspection Scope
The team reviewed the updated safety analysis report, design basis documents, the
current system health report, selected drawings and calculations, maintenance and test
procedures, and condition reports associated with the standby liquid control pump motor.
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:
- Component maintenance history and corrective action program reports to verify
the monitoring of potential degradation.
- The protective device ratings to ensure adequate motor circuit protection and
selective protection coordination of connected equipment during worst-case short
circuit conditions.
b. Findings
No findings were identified.
.2.5 Core Spray Pump Motor Protection
a. Inspection Scope
The team reviewed the updated safety analysis report, design basis documents, the
current system health report, selected drawings and calculations, maintenance and test
procedures, and condition reports associated with the core spray pump motor. 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:
- Component maintenance history and corrective action program reports to verify
the monitoring of potential degradation.
- Calculations for electrical protection to verify adequate overcurrent relay settings
for motor circuit design basis requirements.
- Protective relay test and calibration results to verify motor overcurrent relays
performed in accordance with acceptable setting tolerances.
b. Findings
No findings were identified.
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.2.6 480 Vac Safety-Related Motor Control Center K
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 480 Vac Safety-
Related Motor Control Center K. 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:
- Component maintenance history and corrective action program reports to verify
the monitoring of potential degradation.
- 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.
- The protective device settings and circuit breaker ratings to ensure adequate
selective protection coordination of connected equipment during worst-case short
circuit conditions.
- Procedures for preventive maintenance, inspection, and testing to compare
maintenance practices against industry and vendor guidance; including the cable
aging management program.
- Results of completed preventative maintenance on switchgear and breakers,
including breaker tracking.
b. Findings
No findings were identified.
.2.7 4160 Vac Safety-Related Switchgear 1F
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 4160 Vac
safety-related switchgear 1F. 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:
- Component maintenance history and corrective action program reports to verify
the monitoring of potential degradation.
- 10 -
- Calculations for electrical distribution, system load flow/voltage drop, short-circuit,
cables routing, and electrical protection to verify that bus capacity and voltages
remained within acceptable limits.
- The protective device settings and circuit breaker ratings to ensure adequate
selective protection coordination of connected equipment during worst-case short
circuit conditions.
- Procedures for preventive maintenance, inspection, and testing to compare
maintenance practices against industry and vendor guidance; including the cable
aging management program.
- Results of completed preventative maintenance on switchgear and breakers,
including breaker tracking.
b. Findings
Failure to Perform an Operability Review of a Condition Report
Introduction. The team identified a Green, non-cited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees
failure to accomplish an operability review in accordance with Procedure .05.OPS,
Operations Review of Condition Reports/Operability Determination. Specifically,
the licensee failed to ensure that an operability review was performed for Condition
Report CR-CNS-2015-01268, which was initiated to investigate how plant equipment
is effected due to harmonics on under-voltage relays.
Description. The team performed a review of corrective actions associated with
4160 Vac safety-related switchgear 1F. Under-voltage relays (27-XX) are used
to monitor voltage levels (degraded or loss of voltage) in 4160 Vac switchgear.
Relay 27-1F2 is the under-voltage relay for 4160 Vac Switchgear 1F. In January 2015
Cooper Nuclear Station personnel performed a Component Design Basis Inspection
focused self-assessment in advance of the April 2015 NRC Component Design Basis
Inspection. During this assessment, it was found that the impact of harmonics on
under-voltage relays had not been considered in Calculation NEDC 88-086B,
Setpoint Determination of Second Level Under-Voltage Relays. Condition
Report CR-CNS-2015-01268 was then initiated to evaluate the impact of harmonics on
under-voltage relays. Harmonics are available on medium voltage switchgear through
breaker functions and rotating machinery. It is stated in the condition report that
Harmonics may affect when the undervoltage relays trip, causing the essential buses to
be shed earlier or not at all when it is desired.
In reviewing the corrective action specified in Condition Report CR-CNS-2015-01268,
the team determined that, though an operability review was stated to be required, none
had been performed. The licensee also concluded that the harmonics condition does
not affect installed plant equipment. The team questioned the validity of this conclusion
when no operability review was performed and it is known that the type of under-voltage
- 11 -
relays installed at Cooper Nuclear Station are susceptible to harmonic effect. In
response, the licensee initiated Condition Report CR-CNS-2015-02337 to document
the deficiency and perform the operability review in accordance with Procedure .05.OPS,
Operations Review of Condition Reports/Operability Determination. The result of
the operability review was to perform an operability evaluation in accordance with
Procedure 0.5.OPS to support a prompt operability determination.
Analysis. The team determined that failure to perform an operability review associated
with Condition Report CR-CNS-2015-01268 was a performance deficiency. This finding
was more than minor because it was associated with the human performance attribute of
the Initiating Events cornerstone and adversely affected the cornerstone objective to limit
the likelihood of events that upset plant stability and challenge critical safety functions
during shutdown, as well as power operations. Specifically, the licensee failed to
perform the required operability review for the identified condition. In accordance with
Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process
(SDP) for Findings At-Power, dated June 19, 2012, Exhibit 1, Initiating Event Screening
Questions, the issue screened as having very low safety significance (Green) because
the finding did not cause a reactor trip and it did not involve the loss of mitigation
equipment. This finding had a cross-cutting aspect in the area of human performance
associated with teamwork because individuals and work groups failed to communicate
and coordinate their activities across organizational boundaries to ensure nuclear safety
is maintained [H.4].
Enforcement. The team identified a Green, non-cited violation of 10 CFR Part 50,
Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part,
Activities affecting quality shall be prescribed by documented procedures of a type
appropriate to the circumstances and shall be accomplished in accordance with these
procedures. Contrary to the above, prior to April 6, 2015, the licensee failed to ensure
that activities affecting quality as prescribed by documented procedures of a type
appropriate to the circumstances were accomplished in accordance with those
procedures. Specifically, the licensee failed to follow Procedure .05.OPS, Operations
Review of Condition Reports/Operability Determination, to ensure that an operability
review was performed for Condition Report CR-CNS-2015-01268, which was initiated
during the self-audit for the Component Design Bases Inspection to document that
Cooper Nuclear Station has under-voltage relays that could be affected by harmonics.
In response to this issue, the licensee performed an operability review and an operability
evaluation for the under-voltage relays. This finding was entered into the licensees
corrective action program as Condition Report CR-CNS-2015-02337. 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 05000298/2015007-01, Failure to Perform an Operability Review of a Condition
Report.
- 12 -
.2.8 Startup Station Service Transformer
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 startup station
service transformer. 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:
- The design bases document and updated safety analysis report to verify design
bases requirement for the transformers.
- Component maintenance history and corrective action program reports to verify
the monitoring of potential degradation.
- Calculations for electrical distribution, system load flow/voltage drop, short-circuit,
and electrical and mechanical protection to verify transformer loading and voltage
capacity limits.
- The transformer protective device settings and tap changer settings to ensure
adequate selective protection coordination of connected equipment during worst-
case short circuit conditions.
- Procedures for preventive maintenance, inspection, and testing to compare
maintenance practices against industry and vendor guidance; including the cable
and segregated bus aging management program.
- Transformer dissolved gas analysis test reports to evaluate the result and trend
to ensure the health of the transformer oil and insulations.
b. Findings
No findings were identified.
.2.9 Main Steam Isolation Valve MS-AOV-AO80C
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 main steam
isolation valve MS-AOV-AO80C. The team also conducted interviews with system
engineering personnel to ensure the capability of this component to perform its desired
design basis function. Specifically, the team reviewed:
- 13 -
- Component maintenance history and corrective action program reports to verify
the monitoring of potential degradation.
- Valve stroke time and leakage test procedures, acceptance criteria, and recent
test results.
- Evaluation of the impact of instrument air pressure, steam flow, and building
pressure on the valve closing time.
- Calculation inputs associated with main steam isolation valve minimum and
maximum allowable closing times.
b. Findings
No findings were identified.
.2.10 Suppression Chamber Spray A Inboard Throttle Valve RHR-MOV-MO38A
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 suppression
chamber spray A inboard throttle valve RHR-MOV-MO38A. 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:
- Component maintenance history and corrective action program reports to verify
the monitoring of potential degradation.
- Calculations for the required thrust to operate the motor-operated valve under the
most limiting conditions.
- Motor-operated valve test procedures, acceptance criteria, and recent test
results.
- Component maintenance history and corrective action program reports to verify
the monitoring of potential degradation.
- Calculations for motor-operated valve design basis operating conditions to verify
acceptable methodology for the selection of motor thermal overload protection.
- Periodic testing for motor thermal overload relays to verify there was no
unacceptable deterioration for the relays not bypassed during design basis
conditions.
- 14 -
b. Findings
No findings were identified.
.2.11 Core Spray Pump CS-P-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 core spray
pump CS-P-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:
- Component maintenance history and corrective action program reports to verify
the monitoring of potential degradation.
- Calculations for the required pump flow, head, minimum flow, and net positive
suction head under the most limiting conditions, including under and over-
frequency conditions.
- Pump test procedures, acceptance criteria, and recent test results.
b. Findings
No findings were identified.
.2.12 Standby Liquid Control Pump SLC-P-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 standby liquid
control pump SLC-P-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:
- Component maintenance history and corrective action program reports to verify
the monitoring of potential degradation.
- Calculations for the required pump flow, head, minimum flow, and net positive
suction head under the most limiting conditions, including postulated Anticipated
Transient Without Scram events.
- 15 -
- Pump test procedures, acceptance criteria, and recent test results.
b. Findings
No findings were identified.
.2.13 Residual Heat Removal Heat Exchanger B
a. Inspection Scope
The team reviewed the updated safety analysis report, system description, the current
system health report, selected drawings, maintenance and test procedures, and
condition reports associated with residual heat removal heat exchanger B. 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:
- Work orders and corrective action program documents for the last three years.
- System design criteria and system health reports.
- Corrective action program reports to verify the monitoring and correction of
potential degradation, operability evaluations, and apparent cause evaluations.
- Piping and instrumentation diagrams.
- Residual heat removal heat exchanger plugged tube map.
b. Findings
No findings were identified.
.2.14 Residual Heat Removal Service Water Booster Pump 1C
a. Inspection Scope
The team reviewed the updated safety analysis report, system description, the current
system health report, selected drawings, maintenance and test procedures, and
condition reports associated with residual heat removal service water booster pump 1C.
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:
- Past maintenance records for the last three years.
- Surveillance test results, procedures and preventive maintenance work orders.
- 16 -
- Associated condition reports for the past three years.
- System design basis documents and system modifications.
- Preventive maintenance procedures and completed maintenance work orders.
b. Findings
No findings were identified.
.2.15 Residual Heat Removal Heat Exchanger B Bypass Valve
a. Inspection Scope
The team reviewed the updated safety analysis report, system description, the current
system health report, selected drawings, maintenance and test procedures, and
condition reports associated with residual heat removal heat exchanger B bypass valve.
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:
- Vendor installation instructions.
- Past maintenance records for the last three years.
- Surveillance procedures and surveillance results.
- Leak rate testing for last three years.
- Associated condition reports for the past three years.
- Piping and instrumentation diagram for the residual heat removal heat
exchanger B bypass valve.
b. Findings
No findings were identified.
.2.16 Residual Heat Removal Service Water Motor Operated Valve 89B
a. Inspection Scope
The team reviewed the updated safety analysis report, design bases documents,
selected drawings, maintenance and test procedures, and condition reports associated
with motor-operated valve 89B. The team also performed system walkdowns and
conducted interviews with the system engineering personnel to ensure the capability of
- 17 -
this component to perform its desired design basis function. Specifically, the team
reviewed:
- Technical specifications and basis documents.
- Motor sizing data.
- System design criteria and operating instructions.
- Corrective action program documents and system health reports for the last three
years.
- Piping and instrumentation diagrams.
- Component surveillance test results and trend reports.
- Maintenance records and operational history.
b. Findings
No findings were identified.
.3 Results of Reviews for Operating Experience
.3.1 Inspection of NRC Information Notice 98-31, Fire Protection System Design
Deficiencies and Common-Mode Flooding of Emergency Core Cooling System Rooms
at Washington Nuclear Project Unit 2
a. Inspection Scope
The team reviewed the licensees evaluation of Information Notice 98-31, Fire
Protection System Design Deficiencies and Common-Mode Flooding of Emergency
Core Cooling System Rooms at Washington Nuclear Project Unit 2, to verify the
licensee performed an applicability review and took corrective actions, if appropriate, to
address the concerns described in the information notice. This information notice
addressed the rupture of a fire protection system valve at WNP-2 station in 1998,
resulting in the flooding of emergency core cooling system rooms in the reactor building.
b. Findings
Failure to Adequately Maintain Design Modifications to Prevent Fire Protection System
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 adequately
maintain design modifications that were implemented to prevent fire protection system
water hammer events. Specifically, the licensee failed to maintain procedure changes to
- 18 -
periodically monitor and add nitrogen to fire protection system headers in the reactor
building.
Description. The inspectors reviewed Cooper Nuclear Stations evaluation of
NRC Information Notice 98-31, Fire Protection System Design Deficiencies and
Common-Mode Flooding of Emergency Core Cooling System Rooms at Washington
Nuclear Project Unit 2. This information notice addressed the rupture of a fire
protection system valve at WNP-2 station in 1998. At WNP-2, a water hammer event
resulted in the failure of a fire protection system valve in the reactor building and in the
flooding of two emergency core cooling system equipment rooms. In response to the
information notice, Cooper Nuclear Station implemented design change CED 1998-0060
to introduce nitrogen bubbles into three of the five fire protection risers in the reactor
building. The design change included revisions to station procedures to periodically
monitor and add nitrogen to the fire protection system headers. This design change was
based on calculation NEDC 00-097, which determined the forces associated with a
water hammer event would be significantly reduced by the addition of nitrogen. The
modification included procedure changes to periodically monitor and add nitrogen to the
In response to the inspectors questions, Cooper Nuclear Station personnel determined
that the requirement to monitor and add nitrogen to these fire protection risers was no
longer included in the plant procedures. They also determined that the technical basis
for removing this requirement had not been documented and that no design change was
initiated to implement this procedure change. Cooper Nuclear Station personnel initiated
Condition Report CR-CNS-2015-02085 during the inspection to address the issue. The
condition report recommended that the station either introduce nitrogen into the fire
protection risers in accordance with the previous revision of the plant procedure or
develop a technical basis for the removal of the nitrogen. The licensee determined that
the fire protection system remained functional without the nitrogen based on technical
input from Cooper Nuclear Station engineering personnel. This technical input was
partially based on calculation NEDC 00-097, which determined the maximum peak riser
pressure in the northwest corner of the reactor building and determined the peak force in
the fire protection piping system. The technical input also included a discussion of a
similar transient that occurred at Cooper Nuclear Station in 1995; this transient did not
result in a catastrophic water hammer or piping system damage. The technical input
determined that empirical evidence suggested that the system was capable of absorbing
the shockwave from a water hammer event.
Analysis. The team determined that the failure to adequately maintain control of the fire
protection system design to prevent water hammer events was a performance
deficiency. This finding was more than minor because it was associated with the design
control attribute of the Mitigating Systems cornerstone and adversely affected the
cornerstone objective to ensure the availability, reliability, and capability of systems that
respond to events to prevent undesirable consequences. Specifically, the licensee failed
to maintain procedure changes to periodically monitor and add nitrogen to fire protection
system headers in the reactor building. 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,
- 19 -
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.
Enforcement. The team identified a Green, non-cited violation of 10 CFR Part 50,
Appendix B, Criterion III, Design Control, which states, in part, that design control
measures shall provide for verifying or checking the adequacy of design, such as by the
performance of design reviews, by the use of alternate or simplified calculational
methods, or by the performance of a suitable testing program. Contrary to the above,
prior to April 6, 2015, the licensee failed to verify or check the adequacy of the fire
protection system to remain functional in the event of a water hammer event through
calculational methods or through a suitable testing program. Specifically, the licensee
failed to maintain procedure changes to periodically monitor and add nitrogen to fire
protection system headers in the reactor building to mitigate the effects of water
hammer. In response to this issue, the licensee determined that the fire protection
system remained functional without nitrogen based on empirical evidence suggesting
that the system was capable of absorbing the shockwave from a water hammer event.
This finding was entered into the licensees corrective action program as Condition
Report CR-CNS-2015-02085. 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 05000298/2015007-02, Failure to Adequately Maintain
Design Modifications to Prevent Fire Protection System Water Hammer.
.3.2 Inspection of NRC Information Notice 2013-017, Significant Plant Transient Induced by
Safety-Related Direct Current Bus Maintenance at Plant
a. Inspection Scope
The team reviewed the licensees evaluation of NRC Information Notice 2013-017,
Significant Plant Transient Induced by Safety-Related Direct Current Bus Maintenance
at Plant, to verify the licensee performed an applicability review and took corrective
actions, if appropriate, to address the concerns described in the information notice. This
information notice discusses significant plant transients induced by safety-related direct
current bus maintenance at the plant. The licensee documented the evaluation under
LO 2013-0088-032. The team verified that the licensees review adequately addressed
the issues in the information notice.
b. Findings
No findings were identified.
- 20 -
.3.3 Part 21 No. 2015-0014: Defect Identified in ABB K-Line Breaker Secondary Close
Latches (Part Number 716610K01)
a. Inspection Scope
The team reviewed the licensees evaluation of Part 21 No. 2015-0014: Defect
Identified in ABB K-Line Breaker Secondary Close Latches (Part Number 716610K01)
to verify the licensee performed applicability and vulnerability review of the defect
identified in ABB K-Line Breaker Secondary Close Latches (Part Number 716610K01).
The Operating Experience reviews were evaluated and documented by the licensee in
LO 2013-0087-075, LO 2014-0082-009, and LO 2014-0082-048. It was documented
that Cooper Nuclear Station does not use the ABB K-line circuit breakers in plant
systems. The team verified that the licensees review adequately addressed the issues
in the Part 21 report.
b. Findings
No findings were identified.
.3.4 Inspection of NRC Information Notice 2012-11, Age-Related Capacitor Degradation
a. Inspection Scope
The team reviewed the licensees evaluation of Information Notice 2012-11, Age-
Related Capacitor Degradation, to verify the licensee performed an applicability review
and took corrective actions, if appropriate, to address the concerns described in the
information notice. This information notice discusses age-related degradation of
capacitors that results from epoxy insulation hardening and cracking over time that
allows for a high flow of current and excessive heating. The team verified that the
licensees review adequately addressed the issues in the information notice.
b. Findings
No findings were identified.
.4 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
two 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.
- 21 -
The selected operator actions were:
- Scenario 1, Part 1: Inadvertent Main Steam Isolation Valve closure caused a
reactor scram. A hydraulic Anticipated Transient Without Scram prevented the
reactor from being shut down. The operating crews were expected to enter
Emergency Operating Procedures 6A and 7A to control reactor pressure vessel
pressure and shut down the reactor using boron injection from the Standby
Liquid Control system. The operating crews were also expected to insert control
rods into the reactor per Procedure 5.8.3, Alternate Rod Insertion Methods.
This portion of the scenario specifically evaluated the licensees ability to
successfully initiate boron injection within 2 minutes as referenced in Updated
Safety Analysis Report XIV, Section 5.9.3.4.4.1 and Procedure 2.0.1.3, Time
Critical Operator Action Control and Maintenance.
conditions described above, the operators were expected to take actions per
Emergency Operating Procedure 3A to mitigate adverse primary containment
parameters. Specifically, this portion of the scenario evaluated the licensees
ability to successfully place both loops of the residual heat removal system in
suppression pool cooling mode of operation within 30 minutes as referenced in
Procedure 2.0.1.3, Time Critical Operator Action Control and Maintenance.
- In-plant job performance measure (JPM) #1: This job performance measure was
designed to evaluate the licensees ability to perform subsequent actions for
ensuring emergency ventilation to essential equipment during a station blackout
event. Specifically, the job performance measure evaluated the ability of an
operator to open control room panel doors and 125V/250V dc switchgear room
doors in accordance with Procedure 5.3SBO, Station Blackout within
30 minutes as specified by 5.3SBO and Procedure 2.0.1.3, Time Critical
Operator Action Control and Maintenance.
- In-plant job performance measure (JPM) #2: This job performance measure was
designed to evaluate the licensees ability to perform actions to inject boron into
the reactor pressure vessel using the reactor core isolation cooling system in
accordance with Procedure 5.8.8, Alternate Boron Injection and Preparation.
b. Findings
Failure to Update the Final Safety Analysis Report
Introduction. The team identified three examples of a Severity Level IV, non-cited
violation of 10 CFR 50.71, Maintenance of Records, Making of Reports, for the
licensees failure to update the Final Safety Analysis Report originally submitted as part
of the application for the license to assure that the information included in the report
contains the latest information developed. Specifically, in April 2008, January 2012, and
February 2015 the licensee made changes to their operating procedures, Anticipated
- 22 -
Transient Without Scram analyses, and plant conduct of operations procedures that
were not subsequently reflected in their Updated Safety Analysis Report.
Description. The first example was identified on April 9, 2015, during the teams
review of the licensees time critical operator actions associated with Anticipated
Transient Without Scram events. In August 2014 the licensee identified that contrary
to Cooper Nuclear Station Procedure 2.0.1.3, Time Critical Operator Action Control
and Maintenance, the control room operators could not successfully place the residual
heat removal system in suppression pool cooling mode of operation during a
100 percent Anticipated Transient Without Scram event with main steam isolation valves
closed within the required 11 minutes (per Cooper Nuclear Station Updated Safety
Analysis Report Chapter XIV, Section 5.9.3.3.2.c). The licensee generated Condition
Report CR-CNS-2014-4453 requesting an engineering review of the time critical
operator action. As a result of the engineering review (ER 15-003, Revision 0, dated
February 4, 2015), the time critical operator action time for placing residual heat removal
in suppression pool cooling mode was changed to 30 minutes. The licensee did not
subsequently change Updated Safety Analysis Report, Chapter XIV, Section 5.9.3.3.2.c,
to reflect the new time critical operator action time to reflect the updated Anticipated
Transient Without Scram analysis that resulted from the engineering review. The
inspection team identified that the most recent Updated Safety Analysis Report was
submitted for approval on April 24, 2015, and covered changes to the Updated Safety
Analysis Report through the 24-month period ending March 10, 2015. The licensee
documented this issue in the corrective action program as Condition Report CR-CNS-
2015-02090.
The second example was identified on April 22, 2015, during the teams review of
the licensees time critical operator actions associated with Anticipated Transient Without
Scram events. In January 2012 the NRC issued Amendment No. 240 to Renewed
Facility Operating License No. DPR-46 for Cooper Nuclear Station, concerning changes
to the Technical Specification 3.4.3, Safety/Relief Valves and Safety Valves. As part of
the analyses submitted with the proposed amendment, Cooper Nuclear Station used
bounding Anticipated Transient Without Scram analysis values for safety/relief valve lift
setpoints of +3 percent of nominal lifting setpoint pressure. Cooper Nuclear Station
Updated Safety Analysis Report, Chapter XIV, Table XIV-5-4, was not updated to reflect
the +3 percent safety/relief valve lift setpoint pressure (the table displays nominal lift
setpoint pressures). The licensee documented this issue in the corrective action
program as Condition Report CR-CNS-2015-02393.
The third example was identified on April 10, 2015, during the teams review of the
internal flooding analysis and related Updated Safety Analysis Report section. The
Updated Safety Analysis Report, Section X-8.2.8.1, Flooding, stated that Two 3" service
water system lines provide an emergency cooling water source for the control room air
conditioner. There is normally no flow in these lines since they have normally closed
manually operated valves in the lines below the 903'6" passageway elevation.
Therefore, these lines pose no problems. However, a corrective action associated
with Condition Report CR-CNS-2007-07623 changed normal valve positions in
Procedure 2.2.76A. As a result, the service water lines were pressurized and were a
potential source of flooding. This change did not invalidate the results of the flooding
- 23 -
analysis, but the statement in the Updated Safety Analysis Report was not correct.
The licensee documented this issue in the corrective action program as Condition
Report CR-CNS-2015-02106.
Analysis. The team determined that the failure to update the Final Safety Analysis
Report to assure that the information included in the report contains the latest
information developed was a performance deficiency. This finding was evaluated using
traditional enforcement because it had the potential for impacting the NRCs ability to
perform its regulatory function. This finding was more than minor because each
example potentially rendered portions of the safety analyses for Anticipated Transient
Without Scram events described in the Updated Safety Analysis Report less
conservative or contradicted previous information regarding the licensees flooding
analysis contained in the Updated Safety Analysis Report. The traditional enforcement
violation was determined to be a Severity Level IV violation consistent with the example
in paragraph 6.1.d(3) of the NRC Enforcement Policy. Since this was a traditional
enforcement violation, no cross-cutting aspects were assigned per the guidance
contained in Inspection Manual Chapter 0612, Section 07.03(c).
Enforcement. The team identified three examples of a Severity Level IV, Green,
non-cited violation, of 10 CFR 50.71, Maintenance of Records, Making of Reports,
Section (e), which states, in part, each person licensed to operate a nuclear power
reactor under the provisions of 10 CFR 50.21 or 10 CFR 50.22 shall update periodically
the final safety analysis report (FSAR) originally submitted as part of the application for
the license, to assure that the information included in the report contains the latest
information developed. This submittal shall contain all the changes necessary to
reflect information and analyses submitted to the Commission by the licensee since the
submittal of the original FSAR, or as appropriate, the last update to the FSAR under this
section. Contrary to the above, in January 2012 and February 2015, the licensee failed
to update periodically the Final Safety Analysis Report (FSAR) to contain all the changes
necessary to reflect information and analysis since the last update to the Final Safety
Analysis Report. Specifically, the licensee failed to update the Updated Safety Analysis
Report for changes made to their Anticipated Transient Without Scram analyses and
plant conduct of operations procedures. This finding was entered into the licensees
corrective action program as Condition Reports CR-CNS-2015-02106,
CR-CNS-2015-02090, and CR-CNS-2015-02393. Because this finding was not
repetitive or willful and has been entered into the licensees corrective action program,
this Severity Level IV violation is being treated as a non-cited violation, consistent with
Section 2.3.2.a of the NRC Enforcement Policy: NCV 05000298/2015007-03, Failure to
Update the Final Safety Analysis Report (FSAR).
- 24 -
4. OTHER ACTIVITIES
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
4OA2 Problem Identification and Resolution (71152)
Component Design Basis Review
a. Inspection Scope
The team reviewed condition reports associated with the selected components, operator
actions, and operating experience notifications. The team also reviewed corrective
actions associated with items identified in previous inspections. Specifically, the team
reviewed the updated safety analysis report, system description, design basis
documents, selected drawings, maintenance and test procedures, and condition reports
associated with the emergency diesel generator day and storage tank vents. The team
also performed walkdowns and conducted interviews with system engineering personnel
to ensure the capability of this component to perform the desired design function.
Specifically, the team reviewed:
- Component maintenance history and corrective action program reports to
verify the monitoring of potential degradation.
- Design basis adverse weather protection requirements.
- Normal and alternate diesel fuel oil fill procedures.
- Detailed plant drawings and operating, preventive maintenance, and testing
procedures.
b. Findings
Failure to Evaluate the Lack of Missile Protection on the Emergency Diesel Generator 1
and 2, Fuel Oil Storage Tank Vents
Introduction. The team identified a Green, cited violation of 10 CFR Part 50,
Appendix B, Criterion III, Design Control, for the licensees failure to evaluate the lack
of missile protection on the emergency diesel generator 1 and 2 fuel storage tank vents.
Description. The Cooper Nuclear Station Safety Evaluation Report (SER) and Updated
Safety Analysis Report states the following with regard to General Design Criteria and
the emergency diesel generators:
- Updated Safety Analysis Report, Appendix F, states that the licensee complies
with Draft General Design Criteria GDC-2, published July 11, 1967, and the Draft
General Design Criteria GDC-2 requires that the systems and components
- 25 -
needed for accident mitigation remain fully functional before, during, and after a
tornado event.
- Updated Safety Analysis Report, Chapter I-5, Section 5.2, defines Class I
structures and equipment as, Structures and equipment whose failure could
cause significant release of radioactivity or which are vital to a safe shutdown of
the plant and removal of decay and sensible heat.
- Safety Evaluation Report, Section 3.5, states that Class I structures were
designed to withstand the effects of a spectrum of tornado generated missiles of
low level origin, including a 35 foot long utility pole with a 14 inch butt, with an
impact velocity of 200 miles per hour.
- Updated Safety Analysis Report, Chapter XII-2, Section 2.1.2.3, specifically
identifies the Standby Diesel Generator System and Auxiliaries as Class I
equipment.
- Updated Safety Analysis Report, Chapter XII, Section 2.3.3.2.1, states that
Class I structures are designed to provide protection against tornado generated
missiles.
On December 3, 2010, NRC Component Design Basis Inspection (CDBI)
Report 05000298/2010007 (ML103370640), documented Non-cited
Violation 05000298/2010007-04, Inadequate Design Control, for the licensees
failure to establish design control measures, involving the performance of a design
review, or the use of alternate or simplified calculational methods, or the performance
of a suitable testing program to verify that the emergency diesel generator fuel oil
storage and day tank vent lines were adequately protected from tornado generated
missiles. The licensee entered this deficiency into their corrective action program
as Condition Report CR-CNS-2010-05211 and generated Engineering
Evaluation (EE)10-060, Evaluation of the Diesel Generator Fuel Oil Tanks.
Subsequently, on February 13, 2014, NRC Inspection Report 05000298/2013005
(ML14044A105) documented Non-cited Violation 05000298/2013005-01, Failure to
Promptly Identify and Correct a Condition Adverse to Quality, for the licensees failure
to promptly identify and correct Non-cited violation 05000298/2010007-04, Inadequate
Design Control. Specifically, inspectors determined that Engineering
Evaluation EE 10-060 did not evaluate the vent lines with regard to their ability to
withstand tornado generated missiles. Instead, it assumed that if impacted by a missile,
there would be no damage to the fuel oil storage tank and discussed manual actions that
could be implemented if the vent lines were to be damaged by a tornado generated
missile. The licensee entered this deficiency into their corrective action program as
Condition Report CR-CNS-2014-00146.
In response to Non-cited Violation 05000298/201305-01, the licensee provided a reply
contained in a letter from Mr. O. Limpias to the NRC, dated May 20, 2014, which
disputed the use of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, as
- 26 -
the basis for the non-cited violation. As stated in this letter, the licensee denied that a
violation of NRC requirements had occurred, in that, the licensee had previously
evaluated this condition as documented in Condition Report CR-CNS-2010-05211,
which was initiated in response to a nonconforming condition identified during the
2010 Component Design Basis Inspection. The licensee also indicated that they had
re-evaluated these results and concluded the original evaluation remained valid.
Specifically, Engineering Evaluation EE 10-060, Evaluation of the Diesel Generator Fuel
Oil Tank Vents After a Tornado Strike, Revision 0, was performed to establish the basis
for conformance with the pre-General Design Criteria 2, contained in Appendix F, of the
Updated Safety Analysis Report.
In conclusion, the licensees violation denial letter stated that the previous NRC
Component Design Basis question related to the diesel generator fuel oil storage tank
vents ability to withstand a tornado missile strike was adequately resolved under
Condition Report CR-CNS-2010-05211 and appropriately evaluated in a timely manner
commensurate with 10 CFR 50, Appendix B, Criterion XVI.
Violation Denial Review
Consistent with the guidance provided in Policy Guide 0560-3, Region IV Enforcement
Procedures, the NRC staff performed an independent review of the documentation
associated with this finding. Based on the results of this review it was determined that
the requirements of the draft General Design Criteria, Criterion 2, clearly establish the
design function of systems and components of reactor facilities which are essential to
the prevention of accidents which could affect public health and safety or mitigation of
their consequences. These systems and components are required to be designed,
fabricated, and erected to performance standards that will enable the facility to
withstand, without loss of the capability to protect the public, the additional forces that
might be imposed by natural phenomena such as tornados. Furthermore, the system
design basis requirements contained in the Cooper Nuclear Station Updated Safety
Analysis Report, Chapter XII, Section 2.3.3.2.2, Tornado Generated Missiles, specifies
that all Class I structures are designed to provide protection against tornado generated
missiles including:
- A 35-foot long utility pole with a 14-inch butt with an impact velocity of 200 miles
per hour.
- A one-ton missile such as a compact-type automobile with an impact velocity of
100 miles per hour and a contact area of 25 square feet.
- A two-inch extra heavy pipe, 12 feet long.
- Any other missile resulting from failure of a structure or component or one which
has potential of being lifted from storage or working areas at the site.
Additionally, the Cooper Nuclear Station Design Basis for the Diesel Generator Fuel Oil
system includes the following requirements:
- 27 -
- The standby diesel generator system must be capable of withstanding the most
severe conditions anticipated at the location of the plant. The design basis
events are described in IEEE-308-1970, Table I. This table includes postulated
earthquake, wind, hurricane, and tornado effects as natural phenomena design
basis. Additionally, Table I of IEEE-308-1973, lists accident-generated missiles
as one of the events that the emergency diesel system must be designed to
withstand.
- The fuel oil subsystem must provide sufficient fuel to operate the standby diesel
generator under all postulated conditions.
- The safety classification of the essential emergency diesel system including the
diesel fuel oil tank vents is Seismic Class I.
The NRC concluded that the diesel generator fuel oil storage tank Seismic Class I vents
were not assured to be designed, fabricated, and erected to withstand the additional
forces imposed by natural phenomena such as tornados, as required by the licensing
basis stated above. Specifically, the licensees evaluation performed in accordance
with Condition Report CR-CNS-2010-05211 and the associated Engineering
Evaluation EE-10-060 did not adequately demonstrate that the diesel generator fuel oil
storage tank vent lines would maintain its ability to withstand a postulated tornado
missile impact without loss of function. Although the evaluation references the location
of the vents, the area of exposure of the vents to missile impact, and generally discusses
the material composition of the vents and the inferred minimal load transferred to the
diesel generator fuel oil storage tanks, no definitive analytical basis was identified for
concluding that the vent lines would not be damaged by the postulated tornado
generated missile and they would remain functional. While the licensees compensatory
actions dealt with the initial operability condition, the requisite corrective and preventive
measures failed to address the nonconforming design condition, concerning the diesel
generator fuel oil storage tank vents tornado missile protection, initially identified as a
performance deficiency in NRC Component Design Basis Inspection
Report 05000298/2010007.
Based on these reviews, it was concluded that the finding and non-cited violation for
failing to assure that an identified condition adverse to quality was promptly corrected to
meet the requirements in 10 CFR Part 50, Appendix B, Criterion XVI, as documented in
NRC Inspection Report 05000298/2013005, were valid. The failure to perform a proper
engineering evaluation of the diesel generator fuel oil storage tank vents to demonstrate
the ability to perform its specified safety function as required by the licensing bases in
the event of a tornado generated missile has not been documented.
Current Evaluation Results
During the performance of the 2015 Component Design Basis Inspection, the team met
with the licensees engineering and licensing staff to establish the current status of the
diesel generator fuel oil storage tank vents. As a result of these discussions, it was
determined that calculation NEDC 13-046, Diesel Generator Storage Tank Vent Line
- 28 -
Tornado Missile Durability, had been developed to demonstrate that the current design
was acceptable without further action. Specifically, this calculation evaluated the ability
of the diesel generator fuel oil storage tank vents to remain operable following an impact
from design basis tornado generated missiles. As described in NEDC 13-046, the diesel
generator fuel oil storage tank vent lines are part of the fuel oil subsystem of the
emergency diesel generator system. Each component of this system is classified as
Seismic Class 1S. As described in the Updated Safety Analysis Report, Class I
structures are designed to provide protection against tornado generated missiles. The
calculation determines the maximum deflection the vent line can experience without
permanent deformation, the amount of deflection the vent line can experience without
fracture and the amount of force that would be required to fracture the vent line. Forces
from the three Updated Safety Analysis Report specified tornado generated missiles
were determined for comparison to the maximum allowable force the vent line can
withstand. Specifically, using the worst case tornado generated missile, the calculation
concluded that Due to the overwhelming magnitude of the force and the very short
duration of the impact, the vent pipes will shear off or fracture rather than bend and
crimp.
Based on the review of NEDC 13-046, the team determined that the calculation did not
provide an adequate design analysis that would assure that the diesel generator fuel oil
vent lines could maintain an open vent path during a postulated tornado event under all
missile scenarios. Specifically, the calculation failed to provide a bounding analysis that
demonstrated the vent lines would not crimp subsequent to a tornado generated missile
strike from a range of objects which has potential of being lifted from storage or
working areas at the site.
Analysis. The team determined that the failure to evaluate the lack of missile protection
on the emergency diesel generator 1 and 2 fuel storage tank vents was a performance
deficiency. This finding was more than minor because it was associated with the design
control attribute of the Mitigating Systems cornerstone and adversely affected the
cornerstone objective to ensure the availability, reliability, and capability of systems that
respond to events to prevent undesirable consequences. Specifically, the licensee failed
to evaluate a design nonconformance on the emergency diesel generator 1 and 2 fuel
storage tanks for lack of missile protection. 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,
this finding 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.
This finding had a cross-cutting aspect in the area of human performance associated
with conservative bias because individuals failed to use decision making practices that
emphasize prudent choices over those that are simply allowable [H.14].
Enforcement. The team identified a Green, cited violation of 10 CFR Part 50,
Appendix B, Criterion III, Design Control, which states, in part, Design control
measures shall provide for verifying or checking the adequacy of design, such as by the
- 29 -
performance of design reviews, by the use of alternate or simplified calculational
methods, or by the performance of a suitable testing program. Contrary to the above,
since July 2010, the licensee failed to verify the adequacy of the applicable design
control measures. Specifically, the licensee failed to verify the adequacy of design of the
vents for the emergency diesel generator 1 and 2 fuel oil storage tanks to withstand
impact from a tornado driven missile hazard, or to evaluate for exemption from missile
protection requirements using an approved methodology. This finding was entered into
the licensees corrective action program as Condition Report CR-CNS-2015-02366.
VIO 05000298/2015007-04, Failure to Evaluate the Lack of Missile Protection on the
Emergency Diesel Generator 1 and 2 Fuel Oil Storage Tank Vents.
4OA6 Meetings, Including Exit
Exit Meeting Summary
On May 8, 2015, the inspectors presented the inspection results to Mr. O. Limpias,
Vice President-Nuclear and Chief Nuclear Officer, and other members of the licensee
staff. The licensee acknowledged the issues presented. The licensee confirmed that
any proprietary information reviewed by the inspectors had been returned or destroyed.
- 30 -
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
A. Able, Instrumentation and Controls Supervisor, Design Engineering
T. Barker, Manager, Engineering Programs and Components
M. Bergmeier, Operations
D. Buman, Director, Engineering
T. Chard, Manager, Quality Assurance
L. Dewhirst, Manager, Corrective Actions and Assessment
K. Dia, Manager, Systems Engineering
M. Dickerson, Electrical Engineer, Design Engineering
L. DuBois, Emergency Preparedness
J. Ehlers, Electrical/Instrumentation and Controls Supervisor, System Engineering
R. Estrada, Manager, Design Engineering
J. Flaherty, Senior Staff Engineer, Licensing
G. Gardener, Supervisor, NSSS
D. Goochman, Manager, Operations
K. Higginbotham, General Manager, Plant Operations
D. Kimball, Director, Nuclear Oversight
O. Limpias, Senior Vice President, Nuclear, and Chief Nuclear Officer
E. Nelson, Supervisor, Emergency Preparedness
T. Ocken, Supervisor, Design Engineering
C. Pelchat, Manager, Projects
R. Penfield, Director, Nuclear Safety Assurance
J. Shaw, Manager, Licensing
J. Stough, Manager, Information Technology
C. Sunderman, Manager, Radiation Protection
K. Tom, Assistant to the Director, Engineering
A. Walters, Manager, Chemistry
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
05000298/2015007-04 VIO Failure to Evaluate the Lack of Missile Protection on the
Emergency Diesel Generator 1 and 2 Fuel Oil Storage Tank
Vents (Section 1R21.4OA2)
Opened and Closed
Failure to Perform an Operability Review of a Condition Report
(Section 1R21.2.7)
Failure to Adequately Maintain Design Modifications to Prevent
Fire Protection System Water Hammer (Section 1R21.3.1)
Failure to Update the Final Safety Analysis Report (FSAR)05000298/2015007-03 NCV
(Section 1R21.4)
Attachment
LIST OF DOCUMENTS REVIEWED
Calculations
Number Title Revision
NEDC 84-050GH Cable Tray Fill Calculation, Tray C35A, Section 134 1
NEDC 84-050GJ Cable Tray Fill Calculation Tray C35A, Section 136 1
NEDC 87-047K Motor Control Center K Load Summary 3
NEDC-86-105B 4160 Volt Switchgear Critical Bus 1F 10
B&R Calculation Cable Sizing Main Motors & Feeders Cable 1
2.05.01
B&R Calculation Cable Sizing summary - 4160 V Motors & Feeders 1
2.05.02
NEDC 91-190 Cable Withstand Evaluation 3
NEDC 00-003 Cooper Nuclear Station Auxiliary Power System Load Flow 8
and Voltage Analysis
NEDC-08-044 Assessment of SSST with Loss of Fan Cooling for PRA 0
NEDC 87-153 SLC Capacity 0
NEDC 87-158 NPSH Calculation for SLC Pumps 0
NEDC 87-167 SLC Operating Pressure with Two Pumps 2
NEDC 88-190 Essential Pump Minimum Flow 0
NEDC 88-286 MSIV Accumulator Capacity Evaluation 0
NEDC 89-1886 CNS Station Blackout (SBO) Condensate Inventory 3
NEDC 91-185 MOV Thermal Overload Heater Sizing 6
NEDC 93-125 Stem Nut Wear Evaluation 0
NEDC 94-142 Core Spray Flows with Minimum Flow Bypass (MFB) Valve 5
Open
NEDC 94-190 Core Spray Pump Miniflow Analysis 0
NEDC 95-003 Determination of Allowable Operating Parameters for CNS 29
NEDC 96-030 SLC Vortex Limit 0
NEDC 97-044A NPSH Margins for the RHR and CS Pumps 4
NEDC 97-071 Design Report RAL-1030 Spherical Disk Piston-Flite flow 0
A-2
Calculations
Number Title Revision
NEDC 00-041 Limiting Component Analysis for Containment Spray Motor- 2
Operated Valves
NEDC 00-042 AC MOV Program MOVs 2
NEDC 00-047 Determination of Allowable Operating Parameters for RHR- 0
MOV-MO38A and RHR-MOV-MO38B
NEDC 00-080 Flood Door Gap Analysis 1
NEDC 00-110 MOV Program Valve Margin Determination 8
NEDC 01-064 8-hour ECST Volume Requirements for an Isolated Reactor 2
NEDC 01-072 ECCS Pump NPSH/ Vortex Limit with Suction from CST 1
NEDC 09-102 Internal Flooding - HELB, MELB, and Feedwater Line Break 1
NEDC 10-053 EDG High/Low Frequency Effect on ECCS Pumps 2
NEDC 87-131C 125 VDC Division 1 Load and Voltage Study 15
NEDC 86-105C CNS DC Short Circuit Study 4
NEDC 89-2163 Control Building HVAC 5
NEDC 86-105D CNS Critical DC Bus Coordination Study 8
NEDC 87-131C Service Test Profile 15
NEDC 91-094 125 VDC/250 VDC Battery Charger Analysis 5
NEDC 91-185 MOV Thermal Overload Heater Sizing 6
NEDC 99-083 Proto Power Calculation 99-056 Service Water Booster Pump 1
System Hydraulic Analysis
NEDC 99-056 Evaluation of Residual Heat Removal Service Water Booster 0
Pump Needed Differential Pressure
NEDC 97-044A Net Positive Suction Head Margins Residual Heat Removal & 4
Core Spray Pumps
NEDC 94-190 Core Spray Pump Miniflow Analysis 0
NEDC 94-142 Core Spray Flows with Min Flow Bypass Valve Open 5
NEDC 93-184 Residual Heat Removal Heat Exchangers Thermal 3
Performance &Tube Plug Margin
NEDC 93-125 Stem Nut Thread Wear Evaluation, Generic Letter 89-10 0
NEDC 00-47 Allowable Operating Parameters RHR-MOV38A & 38B 0
A-3
Condition Reports (CRs)
2009-03704 2015-02061 2010-03876 2015-01268 2011-06146
2013-0088-032 2013-08099 2014-08465 2014-01680 2014-08318
2012-09382 2012-06657 2012-01997 2015-02024 2012-01647
2014-05947 2008-00666 2007-00773 2010-06302 2012-04960
2014-06054 2010-05211 2014-00146
Condition Reports (CRs) Generated during the Inspection
2015-02366 2015-02007 2015-02407 2015-02337 2015-02384
2015-02747 2015-02736 2015-02752 2015-02330 2015-02034
2015-02441 2015-02085 2015-02650 2015-02089 2015-02106
2015-02115 2015-02358 2015-02366 2015-02384 2015-02395
2015-02408 2015-02409 2015-02440 2015-02090 2015-02393
2015-02085
Work Orders
4458028 4699196 4698778 4442920 4699195
4748604 4983674 5072695 11117931 MWR 99-3212
4951675 4999222 5023453 5023550 5025370
5023592 4951419 4801811 4740703
Drawings
Number Title Revision/Date
3001 Main One Line Diagram AC/24
3002, Sh. 1 Auxiliary One line Diagram MCC Z, SWGR Bus 1A, 1B, 1E, AC/51
and Critical Bus 1F 1G
3006, Sh. 5 Auxiliary One line Diagram Starter Racks LZ and TZ, MCCs K, AE/83
L, LX, RA, S, T, TX, X
3127, Sh. 6 Turbine Generator Building Cable tray loading schedule N11
2018 Flow Diagram Turbine Generator Bldg. & Control Bldg. AD/40
Heating and Ventilating Cooper Nuclear Station
932-71212PI, Control Building H & V Unit 1-HV-C-1A N06
Sh. 1C
3752, Sh. 1 Annunciator Loop Diagram ANN-MUX-02 N04
3750, Sh. 1 Annunciator Loop Diagram ANN-MUX-00 N05
A-4
Drawings
Number Title Revision/Date
3157 Reactor Building Elev. 931-6 conduit and Tray plan AG/27
3013 Generator Tripping Schedule 9
3012, Sh. 2 Main three Line Diagram N09
3253, Sh. K1 460V Motor Control Center K Connection Wiring Diagram N19
3253, Sh. DT4 460V Motor Control Centers Wiring Details, Connection N20
Wiring Diagram
E506 Turbine Generator Building Connection Wiring Diagram, N04
Sheet 64
3255, Sh. 38 Control Room-Control Panels Connection Wiring Diagram N11
3156, Sh. 1 Reactor Building, El 903-6 Conduit and Tray Plan AC/35
E501, Sh. 48A Integrated Control Circuit Diagram, SW-MOV-M089B, AC/04
RHR HX B Service Water outlet
3007, Sh. 6 Auxiliary One Line Diagram, Motor Control Centers, N83
E, O, R, RB & Y
INV-3C-70048, Schematic Diagram ARR 130K200F N02
Sh. 2 of 2
3058 DC One Line Diagram 64
152D009 250V & 125V DC Switchgear One Line & Schematic N03
48K7A.STK MCC-K Circuit 7A March 18,
1994
2040, Sh. 2 Flow Diagram Residual Heat Removal System Loop B AB/19
791E264, Sh. 3 Elementary Diagram RCIC System Cooper N21
791E264, Sh. 7 Elementary Diagram RCIC System Cooper N15
791E271, Sh. 3 Elementary Diagram HPCI System Cooper N23
791E271, Sh. 8 Elementary Diagram HPCI System Cooper N20
2049, Sh. 2 Flow Diagram Condensate Supply System AC/38
Procedures
Number Title Revision/Date
7.3.17 4160 Breaker Maintenance 36
7.3.17-1 4160 Breaker Examination 29
7.3.17.4 4160V Vacuum Bottle Breaker Maintenance 1
3.11 Vendor Manual Control and Use 26
A-5
Procedures
Number Title Revision/Date
0-EN-OE-100 Operating Experience Program 16CS
14.35.1 Electrical Equipment Instrument Calibration 10
7.3.41 Examination and High-pot testing of Non-Segregated 10
Buses and Associate Equipment
2.2.38 HVAC Control Building 40
7.3.2.1 Westinghouse DB-50 Breaker Maintenance and Testing 19
7.3.13 Motor Control Center Examination and Maintenance 22
2.1.11.1 Turbine Building Data 148
0.5.OPS Operations Review of Condition Reports/Operability 53
Determination
6.SWBP.201 Surveillance Procedure SW-MO-89A/B Full Stroke 6
Operability (IST)
3-EN-DC-126 Engineering Calculation Process 3C2
2.3-C-2 Operator Observation and Action Startup Transformer 45
Trouble Panel Window C-2/F-9
14.11.16 IAC Procedure Yokogawa Recorder DX Series Calibration 73
Check
MNT118-00-00 Lesson on Calibration Tool Issues 07
2.2.15 Startup Transformer 54
6.EE.610 Off-Site AC Power Alignment 37
0-CNS-LI-102 Corrective Action Process 0
7.0.2 Preventive Maintenance Program Implementation 53
7.3.2 DC DB-25 and DB-50 Fused Disconnect Testing and 22
Maintenance
7.3.39 Inspection of 125/250 VDC Buses and Switchgear A and B 4
7.3.27.1 125V Station Battery Equalizing Charge 13
7.3.31.3 125V/250V Battery Terminal Cleaning and Torqueing 14
2.2.25.1 125 VDC Electrical System (Div. 1) 19
7.3.1.6 125/250 VDC Station Battery Charger Protective Relays 18
Testing and Calibration
7.3.14 Thermal Examination of Plant Components 10
7.3.23.6 Battery Charger Clean and Inspect 1
2.0.1.3 Time Critical Operator Action Validation 4
A-6
Procedures
Number Title Revision/Date
2.1.20.3 RPV Refueling Preparation (Wet Lift of Dryer and Separator) 52
2.2.99 Supplemental Diesel Generator System 5
3-EN-DC-304 MOV Thrust/Torque Setpoint Calculations 1C0
5.8.6 RPV Flooding Systems (Table 6) 32
5.8.7 Primary Containment Flooding/Spray Systems 30
6.1CS.101 Core Spray Test Mode Surveillance Operation (IST) (DIV 1) 28
6.2CS.101 Core Spray Test Mode Surveillance Operation (IST) (DIV 2) 25
6.MS.201 Main Steam Isolation Valve Operability Test (lST) 17
6. PC.513 Main Steam Local Leak Rate Tests 24
6.SDG.101 SDG Test Mode Surveillance Operation 4
6.SLC.101 SLC Pump Operability Test 23
6.SLC.102 SLC Test Mode Surveillance Operation (IST) 27
7.2.24.2 MSIV Speed Adjustment 2
89-176 MSIV Closing Test with Instrument Air Valved Out May 25, 1989
O-BARRIER Barrier Control Process 16
O-BARRIER- Control Building 5
CONTROL
5.3SBO Station Blackout 34
5.8 Emergency Operating Procedures (EOPs) 36
5.8, 1A - RPV Control 16
Attachment 1
5.8, 6A - RPV Pressure (Failure to Scram) / Reactor Power 16
Attachment 1 (Failure to Scram)
5.8, 7A - RPV Level (Failure to Scram) 17
Attachment 1
5.8, 3A - Primary Containment Control 15
Attachment 1
Attachment 2
5.8.8 Alternate Boron Injection and Preparation 16
2.1.5 Reactor Scram 71
5.8, Stop and Prevent Hard Card 36
Attachment 4
A-7
Procedures
Number Title Revision/Date
5.8, Failure to Scram Actions Hard Card 36
Attachment 6
2.1.22 Recovering From A Group Isolation 59
0.29.1 License Basis Document Changes 34
0.29.2 USAR Control and Maintenance 19
0-EN-LI-103 Operating License Amendments 7C0
2.0.1.3, Time Critical Operator Action Validation May 1, 2015
Attachment 1
2.0.1.3, Time Critical Operator Action Validation July 21, 2014
Attachment 1
Vendor Documents
Number Title Revision/Date
VM-1188 Vendor Manual 125 & 250 Volt Batteries and Chargers 12
0109D4798 Bus Duct Arrangement 01
022-3-R-0558, Power and Control Circuits line-up 04 Unit-1 2
Sh. 22
IC1000-K240- Siemens System and Service Instruction Manual - Vacuum
A164-X-4AUS Circuit Breaker (vehicle) Type GEH 4.16KV-250MVA,
4.16KV-250MVA upgraded to 350MVA
E50001-F710- Siemens Instruction Manual Type 3AH3 and 3AHc-Vacuum
A251-V1-4A00 Circuit Breaker operator modules 4.16KV to 38KV
GEK-41905 GE Instruction and Recommended Parts for Maintenance
Magne-Blast Circuit Breaker Type AMH-4.76-250-2D 1200 &
2000 Amperes with ML-13 Mechanism
GEK-88771-D GE Instruction Magne-Blast Circuit Breaker Type AMH-4.76-
250-0D AMH-4.76-250-1D
F-1329-D-0460 Connection Diagram, LT IB/24/30MVA OA/FA/FA N01
791E262, Sh. 1 Standby Liquid Control, System N17
791E252, Sh. 1 Nuclear Boiler Process Inst. N12
TR-109641 Guidance on Routine Preventive Maintenance for Magne-Blast October 1998
Circuit Breaker, Supplement to N
VM-0986 Limitorque Valves Composite manual 33
CD 7.4.1.7-7 ABB High Accuracy Voltage Relays ITE-27N Undervoltage A
Relay; ITE-59N Overvoltage Relay
A-8
Vendor Documents
Number Title Revision/Date
IB 7.4.1.7-7 ABB Single Phase Voltage Relays Type 27N High Accuracy D
Undervoltage Relay; 59N High Accuracy Overvoltage Relay
IB 7.4.1.7-7 ABB Single Phase Voltage Relays Type 27N High Accuracy E
Undervoltage Relay; 59N High Accuracy Overvoltage Relay
VMCF 9 3-228 FPE Transformer Installation, Operations and Maintenance
Instructions IN-T-415
Design Basis Documents
Number Title Date
DCD-05 DC Electrical System Design Criteria Document EEDC1 February 2,
2009
DCD-04 AC Electrical System Design Criteria Document EEAC1 October 30,
2014
DCD-12 Core Spray System - Design Criteria Document October 30,
2014
DCD-19 Standby Liquid Control (SLC) System - Design Criteria January 22,
Document 2010
Miscellaneous
Number Title Revision/Date
13-004 Engineering Evaluation - Electrical Bus Outage Maintenance 1
Plans - 24 Month Refueling Cycle Review
EC-4899459 1200 A, 4160V Vacuum Bottle Circuit Breaker Replacement 0
10776733 Notification - Evaluate for Preventive maintenance December 8,
2010
2LE6SJ Project - Hitachi Overhaul AMH 4,76-250 1200 Amp Circuit 0
Breaker S/N:0224A6208-001
061-15288 NLI Overhaul AMH-4.76-250-1D 1200Amp Circuit Breaker August 16,
S/N:224A6204-008 2011
LO 2014-0130 CDBI Focused Self-Assessment Report January 16,
2015
800000042081 Operation 0050, Transformer General Maintenance
IEEE Std 279- IEEE Standard: Criteria for Protection Systems for Nuclear
1971 Power Generating Stations
CED 1998- Nitrogen Cushion Installation into Fire Protection System High 1
0060 Points
A-9
Miscellaneous
Number Title Revision/Date
CED 6029940 Supplemental Diesel Generator May 25, 2010
Core Spray System Health Report January, 2015
Main Steam System Health Report January, 2015
Standby Liquid Control System Health Report January, 2015
EE 01-030 Flood Door Gap Analysis 0
ER 2015-011 Sensitivity Analysis of Diesel Generator Storage Tank Vent 0
Function Following a Tornado Missile Strike
SIL No. 482 MSIV Closure Testing Requirement February 22,
1989
NUMARC 87- Guidelines and Technical Bases for NUMARC Initiatives 1
00
OPL-3A Input Parameters Verification For ATWS Analyses (Cycle 27) 0
TAC NO. Cooper Nuclear Station - Issuance of Amendment Re: January 31,
ME5287 Technical Specification 3.4.3 To Reduce The Number of Safety 2012
Relief Valves Required To Be Operable For Overpressure
Protection
SW06 Simulator Cause and Effect Malfunction - Service Water 01.00
Leakage in Control Building Basement
A-10