ML15203A042
| ML15203A042 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 07/21/2015 |
| From: | Christine Lipa NRC/RGN-III/DRS/EB2 |
| To: | Bryan Hanson Exelon Generation Co, Exelon Nuclear |
| References | |
| IR 2015008 | |
| Download: ML15203A042 (65) | |
See also: IR 05000454/2015008
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION III
2443 WARRENVILLE RD. SUITE 210
LISLE, IL 60532-4352
July 21, 2015
Mr. Bryan C. Hanson
Senior VP, Exelon Generation Company, LLC
President and CNO, Exelon Nuclear
4300 Winfield Road
Warrenville, IL 60555
SUBJECT: BYRON STATION, UNITS 1 AND 2 - NRC COMPONENT DESIGN BASES
INSPECTION; INSPECTION REPORT 05000454/2015008; 05000455/2015008
Dear Mr. Hanson:
On June 16, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed a Component
Design Bases Inspection at your Byron Station, Units 1 and 2. The purpose of this inspection
was to verify that design bases have been correctly implemented for the selected risk-significant
components, and that operating procedures and operator actions are consistent with design and
licensing bases. The enclosed report documents the results of this inspection, which were
discussed on June 16, 2015, with Mr. B. Currier, and other members of your staff.
This inspection examined activities conducted under your license as they relate to public
health and safety to confirm compliance with the Commissions rules and regulations, and
with the conditions in your license. Within these areas, the inspection consisted of a selected
examination of procedures and representative records, field observations, and interviews with
personnel.
Based on the results of this inspection, the NRC has identified an issue that was evaluated
under the risk Significance Determination Process as having very-low safety significance
(Green). The NRC has also determined that a violation is associated with this issue. This
violation was evaluated in accordance with the NRC Enforcement Policy. The current
Enforcement Policy is included on the NRCs web site at http://www.nrc.gov/about-nrc/
regulatory/enforcement/enforce-pol.html.
B. Hanson
-2-
The violation is cited in the enclosed Notice of Violation (Notice), and the circumstances
surrounding it are described in detail in the subject inspection report. The violation is being
cited in the Notice because Byron Station, Units 1 and 2, failed to restore compliance and failed
to have objective plans to restore compliance in a reasonable period following the NRC
identification of an associated Non-Cited Violation (NCV) on June 15, 2012. The associated
NCV was documented in Inspection Report 05000454/2012007; 05000455/2012007.
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 NRC
review of your response to the Notice will also determine whether further enforcement action is
necessary to ensure compliance with regulatory requirements.
Based on the results of this inspection, the NRC has also determined that six additional
NRC-identified findings of very-low safety significance (Green) were identified. The findings
involved violations of NRC requirements. However, because of their very-low safety
significance, and because the issues were entered into your Corrective Action Program, the
NRC is treating the issues as NCVs in accordance with Section 2.3.2 of the NRC Enforcement
Policy. These NCVs are described in the subject inspection report.
If you contest the subject or severity of the Non-Cited-Violation, you should provide a response
within 30 days of the date of this inspection report, with the basis for your denial, to the
U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC
20555-0001, with copies to the Regional Administrator, Region III; the Director, Office of
Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the
NRC Resident Inspector at the Byron Station.
In addition, if you disagree with the cross-cutting aspect assigned to any finding 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 III, and the NRC Resident
Inspector at the Byron Station.
B. Hanson
-3-
In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public
Inspections, Exemptions, Requests for Withholding, of the NRC's "Rules of Practice," a copy
of this letter, its enclosure, and your response (if any) will be available electronically for public
inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS)
component of the NRC's Agencywide Documents Access and Management System (ADAMS).
ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html
(the Public Electronic Reading Room).
Sincerely,
/RA/
Christine A. Lipa, Chief
Engineering Branch 2
Division of Reactor Safety
Docket Nos. 50-454; 50-455
Enclosures:
(2) IR 05000454/2015008; 05000455/2015008;
cc w/encl: Distribution via LISTSERV
Enclosure 1
Exelon Generation Company, LLC
Docket No. 50-454; 50-455
Byron Station, Units 1 and 2
During an U.S. Nuclear Regulatory Commission (NRC) inspection conducted from
April 20, 2015, through May 22, 2015, a violation of NRC requirements was identified.
In accordance with the NRC Enforcement Policy, the violation is listed below:
Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI,
Corrective Action, states, in part, that measures shall be established to assure that
conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations,
defective material and equipment, and non-conformances are promptly identified and
corrected.
Contrary to the above, from June 15, 2012, to May 22, 2015, the licensee failed to
correct a condition adverse to quality (CAQ). Specifically, on June 15, 2012, the
NRC issued a Non-Cited Violation 05000454/2012007-05; 05000455/2012007-05 for the
failure to provide means to detect and isolate a leak in the emergency core cooling
system within 30 minutes for Byron Station, Units 1 and 2, as described in
Section 6.3.2.5 of the Updated Final Safety Analysis Report which is a CAQ. As of
May 22, 2015, the licensee had not corrected the CAQ in a reasonable time period.
Instead, the licensee created action tracking items to develop a plan to correct the
CAQ, and the associated due date was extended at least eight times.
This violation is associated with a Green Significance Determination Process finding.
Pursuant to the provisions of 10 CFR 2.201, Exelon Generation Company, LLC, 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 III; and the NRC Resident Inspector at the Byron Station,
Units 1 and 2, within 30 days of the date of the letter transmitting this Notice. This reply
should be clearly marked as a Reply to a Notice of Violation; VIO 05000454/2015008-09; 05000455/2015008-09, 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.
2
Because your response will be made available electronically for public inspection in the
NRC Public Document Room or from ADAMS, accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html, to the extent possible, it should not include any
personal privacy, proprietary, or safeguards information so that it can be made available to the
public without redaction. If personal privacy or proprietary information is necessary to provide
an acceptable response, then please provide a bracketed copy of your response that identifies
the information that should be protected and a redacted copy of your response that deletes
such information. If you request withholding of such material, you must specifically identify the
portions of your response that you seek to have withheld and provide in detail the bases for your
claim of withholding (e.g., explain why the disclosure of information will create an unwarranted
invasion of personal privacy or provide the information required by 10 CFR 2.390(b) to support
a request for withholding confidential commercial or financial information). If safeguards
information is necessary to provide an acceptable response, please provide the level of
protection described in 10 CFR 73.21.
In accordance with 10 CFR 19.11, you may be required to post this Notice within two working
days of receipt.
Dated this 21 day of July, 2015.
Enclosure 2
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No:
50-454; 50-455
License No:
Report No:
05000454/2015008; 05000455/2015008
Licensee:
Exelon Generation Company, LLC
Facility:
Byron Station, Units 1 and 2
Location:
Byron, IL
Dates:
April 20, 2015, through June 16, 2015
Inspectors:
N. Féliz Adorno, Senior Reactor Inspector, Lead
B. Palagi, Senior Operations Engineer
D. Betancourt Roldán, Reactor Inspector, Mechanical
M. Jones, Reactor Inspector, Mechanical
A. Greca, Electrical Contractor
J. Leivo, Electrical Contractor
Approved by:
Christine A. Lipa, Chief
Engineering Branch 2
Division of Reactor Safety
SUMMARY ................................................................................................................................ 2
REPORT DETAILS .................................................................................................................... 7
1. REACTOR SAFETY ....................................................................................................... 7
1R21 Component Design Bases Inspection (71111.21) ............................................... 7
4. OTHER ACTIVITIES .....................................................................................................29
4OA2 Identification and Resolution of Problems ..........................................................29
4OA6 Management Meetings ......................................................................................38
SUPPLEMENTAL INFORMATION ............................................................................................. 2
KEY POINTS OF CONTACT .............................................................................................. 2
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED ................................................... 2
LIST OF DOCUMENTS REVIEWED .................................................................................. 3
LIST OF ACRONYMS USED .............................................................................................19
2
SUMMARY
Inspection Report 05000454/2015008; 05000455/2015008, 4/20/2015 - 6/16/2015; Byron
Station, Units 1 and 2; Component Design Bases Inspection.
The inspection was a 3-week on-site baseline inspection that focused on the design of
components. The inspection was conducted by four regional engineering inspectors, and
two consultants. Seven Green findings were identified by the team. Six of these findings were
considered Non-Cited Violations of U.S. Nuclear Regulatory Commission (NRC) regulations
while one of these findings was considered a Notice of Violation of NRC regulations. The
significance of inspection findings is indicated by their color (i.e., greater than Green, or
Green, White, Yellow, Red), and determined using Inspection Manual Chapter (IMC) 0609,
Significance Determination Process, dated April 29, 2015. Cross-cutting aspects are
determined using IMC 0310, Aspects Within the Cross-Cutting Areas effective date
December 4, 2014. All violations of NRC requirements are dispositioned in accordance with
the NRCs Enforcement Policy, dated July 9, 2013. The NRC's program for overseeing the
safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor
Oversight Process, Revision 5, dated February 2014.
NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
Green: The team identified a finding of very-low safety significance (Green), and an
associated cited violation of Title 10, Code of Federal Regulations (CFR), Part 50,
Appendix B, Criterion XVI, Corrective Actions, for the failure to correct a Condition
Adverse to Quality (CAQ). Specifically, on June 15, 2012, the U.S. Nuclear Regulatory
Commission (NRC) issued a Non-Cited Violation (NCV) for the failure to provide means
to detect and isolate a leak in the Emergency Core Cooling System (ECCS) within
30 minutes as described in the Updated Final Safety Analysis Report (UFSAR), which
is a CAQ. As of May 22, 2015, the licensee had not corrected the CAQ. This violation
is being cited because the licensee had not restored compliance, or demonstrated
objective evidence of plans to restore compliance in a reasonable period following the
identification of the CAQ. The licensee captured this finding into their Corrective Action
Program (CAP) to promptly restore compliance.
The performance deficiency was determined to be more than minor because it was
associated with the Mitigating Systems cornerstone attribute of procedure quality, and
affected the cornerstone objective of ensuring the availability, reliability, and capability of
mitigating systems to respond to initiating events to prevent undesirable consequences.
In addition, it was associated with the Barrier Integrity cornerstone attribute of procedure
quality, and affected the cornerstone objective of providing reasonable assurance that
physical design barriers protect the public from radionuclide releases caused by
accidents or events. The finding screened as very-low safety significance (Green)
because it did not result in the loss of operability or functionality, and it did not represent
an actual pathway in the physical integrity of reactor containment. Specifically, the
licensee reasonably demonstrated that an ECCS leak could be detected and isolated
before it could adversely affect long-term cooling of the plant. The team determined that
the associated finding had a cross-cutting aspect in the area of human performance
because the licensee did not use a consistent and systematic approach to make
decisions. Specifically, the creation and management of the associated corrective action
assignments were not consistent with the instructions contained in their CAP procedure.
[H.13] (Section 4OA2.1.b(1))
3
Severity Level IV. The team identified a Severity Level IV NCV of 10 CFR 50.59(d)(1),
Changes, Tests, and Experiments, and an associated finding of very-low safety
significance (Green) for the licensees failure to perform a written safety evaluation that
provided the bases for the determination that a change which resulted in the sharing of
the refueling water storage tanks (RWSTs) of both reactor units did not require a license
amendment. Specifically, the licensee did not evaluate the adverse effect of reducing
reactor unit independence. The licensee captured this issue into their CAP with a
proposed action to revise the associated calculation to remove the dependence on the
opposite unit, and/or review the implications of crediting the opposite unit RWST under
their 10 CFR 50.59 process.
The performance deficiency was more than minor because it was associated with the
Mitigating Systems cornerstone attribute of design control, and affected the cornerstone
objective of ensuring the availability, reliability, and capability of mitigating systems to
respond to initiating events to prevent undesirable consequences. In addition, it was
associated with the Barrier Integrity cornerstone attribute of design control, and affected
the cornerstone objective of providing reasonable assurance that physical design
barriers protect the public from radionuclide releases caused by accidents or events.
In addition, the associated traditional enforcement violation was more than minor
because the team could not reasonably determine that the changes would not have
ultimately required NRC prior approval. The finding screened as very-low safety
significance (Green) because it did not result in the loss of operability or functionality,
and it did not represent an actual open pathway in the physical integrity of the reactor
containment. Specifically, the licensee reviewed the affected calculation and reasonably
determined that enough conservatism existed such that adequate net positive suction
head (NPSH) could be maintained without sharing the RWSTs of both reactor units.
The team did not identify a cross-cutting aspect associated with this finding because it
was confirmed not to be reflective of current performance due to the age of the
performance deficiency. (Section 1R21.5.b(1))
Green. The team identified a finding of very-low safety significance (Green), and an
associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the
licensees failure to translate applicable design basis into Technical Specifications (TSs)
Surveillance Requirement 3.5.4.2 implementing procedures. Specifically, these
procedures did not verify the RWST vent line was free of ice blockage at the locations,
and during all applicable MODEs of reactor operation assumed by the ECCS and
containment spray (CS) pump NPSH calculation. The licensee captured this issue into
their CAP to reconcile the affected procedures and calculation.
The performance deficiency was determined to be more than minor because it was
associated with the Mitigating Systems cornerstone attribute of design control, and
affected the cornerstone objective of ensuring the availability, reliability, and capability of
mitigating systems to respond to initiating events to prevent undesirable consequences.
Additionally, it was associated with the Barrier Integrity cornerstone attribute of design
control, and affected the cornerstone objective of providing reasonable assurance that
physical design barriers protect the public from radionuclide releases caused by
accidents or events. The finding screened as very-low safety significance (Green)
because it did not result in the loss of operability or functionality, and it did not represent
an actual open pathway in the physical integrity of reactor containment. Specifically, the
licensee performed a historical review of the last 3 years of operation, and did not find
any instances in which the vent path temperature fell below 35 degrees Fahrenheit.
4
The inspectors did not identify a cross-cutting aspect associated with this finding
because it was confirmed not to be reflective of current performance due to the age
of the performance deficiency. (Section 1R21.5.b(2))
Severity Level IV. The team identified a Severity Level IV NCV of 10 CFR 50.59(d)(1),
Changes, Tests, and Experiments, and an associated finding of very-low safety
significance (Green) for the licensees failure to perform a written evaluation that
provided the bases for the determination that the changes to the emergency service
water cooling tower (SXCT) tornado analysis as described in the UFSAR did not require
a license amendment. Specifically, the associated 10 CFR 50.59 Evaluation did not
address the introduction of a new failure mode, the resulting loss of heat removal
capacity during worst postulated conditions, and addition of operator actions that have
not been demonstrated can be completed within the required time to restore the required
SXCT heat removal capacity during worst case conditions. The licensee captured this
issue in their CAP with a proposed action to revise the 10 CFR 50.59 Evaluation and
submit a Licensee Amendment Request.
The performance deficiency was determined to be more than minor because it was
associated with the Mitigating Systems cornerstone attribute of protection against
external events, and affected the cornerstone objective of ensuring the availability,
reliability, and capability of mitigating systems to respond to initiating events to prevent
undesirable consequences. In addition, the associated tradition enforcement violation
was determined to be more than minor because the team could not reasonably
determine that the changes would not have ultimately required prior NRC approval.
The finding screened as of very-low safety significance (Green) using a detailed
evaluation because a loss of SXCT during a tornado event would degrade one or more
trains of a system that supports a risk-significant system or function. The bounding
change to the core damage frequency was less than 5.4E-8/year. The team did not
identify a cross-cutting aspect associated with this finding because the finding was not
representative of current performance due to the age of the performance deficiency.
(Section 1R21.5.b(3))
Green. The team identified a finding of very-low safety significance and an associated
NCV of TS 5.4, Procedures, for the failure to maintain emergency operating
procedures (EOPs) for transfer to cold leg recirculation. Specifically, the EOPs for
transfer to cold leg recirculation did not contain instructions for transferring the ECCS
and CS systems to the recirculation mode that ensured prevention of potential pump
damage when the RWST is emptied. The licensee captured this finding into their CAP
to create a standing order instructing operators to secure all pumps aligned to the RWST
when it is emptied, and implement long term corrective actions to restore compliance.
The performance deficiency was determined to be more than minor because it was
associated with the Mitigating Systems cornerstone attribute of procedure quality, and
affected the cornerstone objective of ensuring the availability, reliability, and capability of
mitigating systems to respond to initiating events to prevent undesirable consequences.
In addition, it was associated with the Barrier Integrity cornerstone attribute of procedure
quality, and affected the cornerstone objective of providing reasonable assurance that
physical design barriers protect the public from radionuclide releases caused by
accidents or events. The finding screened as of very-low safety significance (Green)
because it did not result in the loss of operability or functionality of mitigating systems,
represent an actual open pathway in the physical integrity of reactor containment, and
5
involved an actual reduction in function of hydrogen igniters in the reactor containment.
Specifically, the incorrect caution would only be used in the event that transfer to sump
recirculation was not completed prior to reaching tank low-level, or if the RWST suction
isolation valves fail to close. With respect to transfer to sump recirculation prior to
reaching tank low-level, a review of simulator test results reasonably determined that
operators reliably complete the transfer to sump recirculation prior to reaching this set
point. With respect to the failure of the RWST suction isolation valves, a review of
quarterly test results reasonably determined the valves would have isolated the tank
when required. The team did not identify a cross-cutting aspect associated with this
finding because it was not confirmed to reflect current performance due to the age of the
performance deficiency. (Section 1R21.6.b(1))
Green. The team identified a finding of very-low safety significance (Green), and an
associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures,
and Drawings, for the failure to make an operability determination without relying on the
use of probabilistic tools. Specifically, an operability evaluation for an SXCT degraded
condition used probabilities of occurrence of tornado events which was contrary to the
requirements of the licensee procedure established for assessing operability of
structures, systems, and components (SSCs). The licensee captured the teams
concern in their CAP to revise the affected operability evaluation without using
probability of occurrence of tornado events.
The performance deficiency was more than minor because it was associated with the
Mitigating Systems cornerstone attribute of protection against external events, and
affected the cornerstone objective of ensuring the availability, reliability, and capability of
mitigating systems to respond to initiating events to prevent undesirable consequences.
The finding screened as of very-low safety significance (Green) using a detailed
evaluation because a loss of SXCT during a tornado event would degrade one or more
trains of a system that supports a risk-significant system or function. The bounding
change to the core damage frequency was less than 5.4E-8/year. The team determined
that this finding had a cross-cutting aspect in the area of human performance because
the licensee did not ensure knowledge transfer to maintain a knowledgeable and
technically competent workforce. Specifically, the licensee did not ensure personnel
were trained on the prohibition of the use of probabilities of occurrence of an event
when performing operability evaluations, which was contained in licensee procedure
established for assessing operability of SSCs. [H.9] (Section 4OA2.1.b(3))
Cornerstone: Barrier Integrity
Green. The team identified a finding of very-low safety significance, and an associated
NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and
Drawings, for the failure to have procedures to maintain the accuracy within necessary
limits of the instrument loops used to verify compliance with the containment average
air temperature TS limit of 120 degrees Fahrenheit. Specifically, in 2007, the licensee
cancelled the periodic preventive maintenance (PM) intended to maintain the necessary
instrument loops accuracy. The licensee entered this issue into their CAP and
reasonably established that the 120 degrees Fahrenheit limit was not exceeded
by reviewing applicable historical records from 2002 to time of this inspection.
6
The performance deficiency was determined to be more than minor because it was
associated with the configuration control attribute of the Barrier Integrity Cornerstone,
and adversely affected the cornerstone objective to ensure that physical design barriers
protect the public from radionuclide releases caused by accidents or events. The finding
screened as very-low safety significance (Green) because it did not represent an actual
open pathway in the physical integrity of reactor containment or involved an actual
reduction in hydrogen igniter function. Specifically, the containment integrity remained
intact and the finding did not impact the hydrogen igniter function. The team determined
that this finding had a cross-cutting aspect in the area of problem identification and
resolution because the licensee did not identify issues completely and accurately in
accordance with the CAP. Specifically, on January 15, 2015, the licensee captured the
lack of periodic PM activities for the containment air temperature instrument loops in the
CAP. However, the licensee failed to completely and accurately identify the issue in that
it was not treated as a CAQ. As a consequence, no corrective actions were
implemented. [P.1] (Section 4OA2.1.b(2))
7
REPORT DETAILS
1.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R21 Component Design Bases Inspection (71111.21)
.1
Introduction
The objective of the Component Design Bases Inspection (CDBI) is to verify that design
bases have been correctly implemented for the selected risk-significant components,
and that operating procedures and operator actions are consistent with design and
licensing bases. As plants age, their design bases may be difficult to determine, and
an important design feature may be altered or disabled during a modification. The
Probabilistic Risk Assessment (PRA) 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.
Specific documents reviewed during the inspection are listed in the Attachment to the
report.
.2
Inspection Sample Selection Process
The team used information contained in the licensees PRA and the Byron Station,
Units 1 and 2, Standardized Plant Analysis Risk (SPAR) Model to identify two scenarios
to use as the basis for component selection. The scenarios selected were a feed and
bleed of the reactor coolant system (RCS), and a loss of ultimate heat sink (UHS).
Based on these scenarios, a number of risk-significant components, including those
with Large Early Release Frequency (LERF) implications, were selected for the
inspection.
The team also used additional component information such as a margin assessment
in the selection process. This design margin assessment considered original design
margin reductions caused by design modification, power uprates, or reductions due to
degraded material condition. Equipment reliability issues were also considered in the
selection of components for detailed review. These included items such as performance
test results, significant corrective actions, repeated maintenance activities, Maintenance
Rule (a)(1) status, components requiring an operability evaluation, U.S. Nuclear
Regulatory Commission (NRC) resident inspector input of problem areas/equipment,
and system health reports. Consideration was also given to the uniqueness and
complexity of the design, operating experience, and the available defense in depth
margins. A summary of the reviews performed and the specific inspection findings
identified are included in the following sections of the report.
The team also identified procedures and modifications for review that were associated
with the selected components. In addition, the team selected operating experience
issues associated with the selected components.
8
This inspection constituted 16 samples (12 components, of which 3 had LERF
implications, and 4 operating experience) as defined in Inspection
Procedure 71111.21-05.
.3
Component Design
a.
Inspection Scope
The team reviewed the Updated Final Safety Analysis Report (UFSAR), Technical
Specification (TS), design basis documents, drawings, calculations and other available
design basis information, to determine the performance requirements of the selected
components. The team used applicable industry standards, such as the American
Society of Mechanical Engineers Code, and Institute of Electrical and Electronics
Engineers Standards, to evaluate acceptability of the systems design. The NRC
also evaluated licensee actions, if any, taken in response to NRC issued operating
experience, such as Information Notices (INs). The review verified that the selected
components would function as designed when required and support proper operation of
the associated systems. The attributes that were needed for a component to perform its
required function included process medium, energy sources, control systems, operator
actions, and heat removal. The attributes to verify that the component condition and
tested capability were consistent with the design bases and appropriate may have
included installed configuration, system operation, detailed design, system testing,
equipment and environmental qualification, equipment protection, component inputs
and outputs, operating experience, and component degradation.
For each of the components selected, the team reviewed the maintenance history, PM
activities, system health reports, operating experience-related information, vendor
manuals, electrical and mechanical drawings, and licensee corrective action documents.
Field walkdowns were conducted for all accessible components to assess material
condition, including age-related degradation, and to verify that the as-built condition was
consistent with the design. Other attributes reviewed are included as part of the scope
for each individual component.
The following 12 components (samples) were reviewed:
Safety Injection Pump (1SI01PB): The team reviewed analyses associated
with inadvertent safety injection (SI) actuation and hydraulic calculations to
assess the pump capability to provide its required accident mitigation function.
The reviewed hydraulic analyses included pump minimum required flow, runout
flow, flow capacity/balance, minimum required net positive suction head (NPSH),
and air entraining vortices. In addition, the team reviewed a sample of operating
procedures associated with pump operation under normal and accident
conditions to assess their consistency with applicable design basis analyses.
The team also reviewed test procedures and completed surveillance tests,
including quarterly and comprehensive in-service testing and flow balances,
to assess the associated acceptance criteria and test results. The team also
reviewed the supporting electrical calculations associated with performance of
the SI pump under design basis conditions. This included review of brake
horsepower requirements for the pump motor, performance under degraded
voltage conditions, and motor protection to assess the capability of the motor to
perform its safety function under design basis conditions. In addition, the team
9
reviewed voltage drop calculations to assess the availability of direct current (DC)
control voltage at the associated bus needed to operate the pump circuit breaker.
The team also performed a non-intrusive visual inspection of the component to
assess overall material condition, configuration, and potential vulnerabilities to
hazards. To assess operating trends and the licensees ability to evaluate and
correct problems, the team reviewed system health reports, selected corrective
action documents, and PM procedures and records.
Pressurizer Power-Operated Relief Valve (1RY456): The team reviewed the
pressure and temperature limit report and calculations associated with the
power-operated relief valve (PORV) lift settings, relief capacity, and set points for
low-temperature overpressure (LTOP) scenarios to assess the PORV capability
to provide its RCS overpressure protection function. The team also reviewed test
procedures and completed surveillances to assess the associated acceptance
criteria and test results. In addition, the team reviewed a sample of associated
operating procedures to assess their consistency with applicable design basis
analyses. The team also reviewed the schematic diagrams for the PORV control
circuit to assess its suitability for bleed-and-feed operation as prescribed by
operating procedures, and to assess the pilot solenoid and position limit switches
qualification for post-accident environmental conditions. The team reviewed
voltage drop calculations to assess the availability of the voltage needed at the
solenoid valve to operate the PORV. The team also reviewed control wiring
schematics and associated instrument loop diagrams to assess the consistency
between operations and system design requirements. This review included a
circuit protection evaluation intended to demonstrate that the containment
electrical penetration was not adversely affected by in-containment faults. The
team also reviewed documentation associated with environmental qualifications
for the postulated containment accident conditions and replacement of
components susceptible to aging. The team reviewed system health reports,
selected corrective action documents, and PM procedures and records to assess
operating trends and the licensees ability to evaluate and correct problems.
Power-Operated Relief Valve Accumulator (1RY32MB): The team reviewed the
accumulator sizing calculation, PORV pressure set point, accumulator stress
analysis, and maximum allowed accumulator leak rate to assess the accumulator
capability to supply the required amount of air pressure and volume to stroke
open its associated PORV on a loss of normal air supply. Additionally, the team
reviewed the design calculation that established the minimum number of PORV
strokes required during certain events, such as LTOP and natural circulation
cooldown. The team also reviewed test procedures and completed surveillances
to assess the associated acceptance criteria and test results. In addition, the
team reviewed a sample of associated operating procedures to assess their
consistency with applicable design basis analyses. Finally, the team reviewed
system health reports, selected corrective action documents, and recent
modifications and operability evaluations to assess operating trends and the
licensees ability to evaluate and correct problems.
Refueling Water Storage Tank (1SI01T): The team reviewed a sample of
associated operating procedures under normal and emergency conditions to
assess their consistency with applicable design basis analyses. The team
also performed a non-intrusive visual inspection of the refueling water storage
10
tank (RWST) to assess overall material condition, configuration, and potential
vulnerabilities to hazards. To assess operating trends, component health, and
the licensees ability to evaluate and correct problems, the team reviewed system
health reports, selected corrective action documents, and recent modifications.
The team reviewed design analyses associated with the ability of the RWST
system to maintain its design function during external events such as tornados
and earthquakes. Additionally, the team reviewed design calculations related to
level set points, temperature limits, and minimum required RWST volume to
mitigate a loss of coolant accident (LOCA), and to support feed-and-bleed
scenarios. The team also reviewed the schematic diagrams and instrument
uncertainty calculations to assess the low-low RWST level signal (i.e., LO-2)
capability to automatically open the containment sump isolation valves
(i.e., 1SI8811A/B) following a LOCA, and its consistency with the associated set
point calculation including instrument uncertainty considerations. To assess
operating trends, component health, and the licensees ability to evaluate and
correct problems, the team reviewed system health reports, selected corrective
action documents, recent modifications, and PM/calibration procedures and
records.
Emergency Service Water Makeup Pump (0SX02PA): The team reviewed
design documents and procedures to assess consistency with vendor
specifications. The team reviewed calculations associated with pump capability
and performance to assess the pump capability to perform its design function of
providing sufficient inventory to the associated Emergency Service Water
Cooling Tower (SXCT) basin under different postulated scenarios. The team
reviewed the water inventory availability from the suction source under routine
service as well as extreme conditions. This review included low and high-river
water levels and temperatures, pump NPSH, pump suction submergence, and
minimum flow protection. The team also reviewed procedures associated with
protection against flooding, seismic, and tornado events since the makeup pump
is credited to some extent during these postulated events. The team also
performed a non-intrusive visual inspection of the pump to assess overall
material condition, configuration, and potential vulnerabilities to hazards.
Work orders and maintenance procedures were reviewed to verify effectiveness
of site maintenance. The team also reviewed test procedures and completed
surveillances to assess the associated acceptance criteria and test results.
To assess operating trends, component health, and the licensees ability to
evaluate and correct problems, the team reviewed system health reports and
selected corrective action documents.
Emergency Service Water Makeup Pump Diesel Engine (0SX02PA-K): The
team reviewed design documents and procedures to assess consistency with
vendor specifications. The team reviewed diesel fuel oil day tank level alarm
response procedures and sizing analyses including the engine diesel fuel oil
consumption rate calculation, tank capacity, vortexing calculation, level
indicators, and alarm setpoint. In addition, the team reviewed the control circuit
electrical diagram to assess the consistency between operations and design
basis requirements. The team also reviewed the set point calculation for the
SXCT basin level switch associated with the starting logic of the diesel engine
to assess consistency between the specified setting and applicable design basis
requirements. In addition, the team reviewed recent level instrument calibration
11
results. The team also reviewed circuit protection and control voltage to assess
the diesel engine capability to start on demand. The inspectors reviewed
completed work orders to assess the as-found and as-left condition of the
diesel engine following recent maintenance activities. The team also reviewed
test procedures and completed surveillances to assess the associated
acceptance criteria and test results. The team also performed a non-intrusive
visual inspection of the engine to assess overall material condition, configuration,
and potential vulnerabilities to hazards. To assess operating trends and the
licensees ability to evaluate and correct problems, the team reviewed system
health reports, selected corrective action documents, modifications, and PM
procedures and records.
Emergency Service Water Cooling Tower (0SX02AA/B and 0SX03CA/H):
The team reviewed design calculations and procedures associated with fan
performance, basin sizing, heat transfer, and makeup requirements during
postulated events including LOCA, tornado, and seismic events. The electrical
calculations associated with fan performance under design basis conditions
were reviewed to assess consistency with the design bases and the motor
capability to perform its specified safety function. This review considered fan
motor brake horsepower requirements, performance under degraded voltage
conditions, and motor protection. The team reviewed voltage drop calculations
to assess the availability of the DC control voltage needed at the associated load
center for the closing and tripping of the cooling tower fan circuit breakers. The
team also reviewed the alternating current (AC) and DC electrical distribution
systems to assess the SXCT capability to perform its specified safety function
assuming a single failure of electrical components. The team also reviewed
control wiring diagrams of the deep well pump and associated control valves to
assess consistency between their operation and design requirements. The team
also performed a non-intrusive visual inspection of the SXCT basin structure, fan
motors, valve houses, and electrical equipment rooms to assess overall material
condition, configuration, and potential vulnerabilities to hazards. The team also
reviewed test procedures and completed surveillances to evaluate the associated
acceptance criteria and test results. To assess operating trends and the
licensees ability to evaluate and correct problems, the team reviewed system
health reports, selected corrective action documents, operability evaluations,
modifications, and PM procedures and records.
4160 Volts Alternating Current Bus 142: The team reviewed voltage drop
calculations to assess the availability of the DC control voltage needed at the
associated bus for the operation of the associated circuit breakers. The team
reviewed calculations associated with load flow, degraded voltage, and protective
settings for selected electrical load paths served by the bus and associated with
the inspection samples to assess the bus capability to support the loads required
safety functions under design basis conditions. The team also performed a
non-intrusive visual inspection of the switchgear to assess overall material
condition, configuration, and potential vulnerabilities to hazards or extreme
service environments. To assess operating trends and the licensees ability to
evaluate and correct problems, the team reviewed system health reports,
selected corrective action documents, and selected PM procedures and records.
12
120 Volts Alternating Current Instrument Bus 111: The team reviewed the DC
voltage drop calculations to assess the availability of the voltage needed for the
proper operation of the associated inverter, including during a loss of AC power.
The team also reviewed the bus loading and breaker ratings to assess the bus
and loads protection against spurious tripping. In addition, the team reviewed a
modification which installed forced air cooling units for the inverter serving the
bus to assess the modification implementation and any potential impact on the
inverter. To assess operating trends and the licensees ability to evaluate and
correct problems, the team reviewed system health reports, selected corrective
action documents, and PM procedures and records for the bus.
125 Volts Direct Current Bus 111: The team reviewed bus loading and short
circuit calculations as well as cable, bus, and circuit breaker ratings to assess
bus and cable capabilities of carrying the maximum anticipated loading and
protection against faulted conditions. The team also reviewed voltage drop and
battery sizing calculations to assess the capability to support momentary and
continuous loading for the duration of the duty cycle during accident conditions
and the loss of all AC power (i.e., station blackout). Additionally, the team
reviewed the battery charger sizing calculation to assess its capability of
maintaining the battery in a charged state and recharging the battery in a timely
manner following a loss of AC power event. The team also reviewed room
heat-up calculations to ensure that the DC components were not adversely
affected by steam line breaks in the turbine building. In addition, the team
reviewed purchase specifications, vendor documents, seismic test reports,
certificate of compliance, and cable separation to assess consistency of the
installed component to the design requirements. For the battery, this review
included an assessment of the inter-cell resistance conformance to voltage drop
calculations. Breaker/fuse coordination was also reviewed to assess the
capability to interrupt overloads and faulted conditions. The team also reviewed
testing procedures and associated recent results, recent system health reports,
molded-case circuit breaker testing, maintenance activities, and recent corrective
action documents to assess component health history.
24 Volts Direct Current Bus 035-2: The team reviewed the sizing calculation for
the diesel start system and the control batteries to assess their capability of
providing adequate voltage to the associated components for the duration of the
duty cycle during accident conditions and loss of all AC power. The team also
reviewed components and wiring schematics related to the diesel start and
control logic to assess the bus capability to perform its intended function.
Additionally, the team reviewed the battery charger sizing calculation to assess
its capability to maintain the batteries in a charged state, and to recharge them in
a timely manner following a loss of AC power event. The team reviewed
purchase specifications, vendor documents, seismic test report, and certificate of
conformance to assess consistency of the installed component to the design
requirements. The team also reviewed testing procedures and associated recent
results, health reports, maintenance activities, and recent corrective action
documents to assess component health history.
480 Volts Alternating Current Motor Control Center 132Z1: The team assessed
conformance to the applicable design and licensing basis by performing an
engineering review of the motor control center (MCC) loading, MCC and control
13
circuits degraded voltage and maximum voltage, electrical protection, and
electrical isolation/physical circuit separation of the MCC from non-safety class
loads. The loads considered during this review were the SXCT riser motor
operated valves (MOVs) (i.e., 0SX163E/F), SXCT makeup MOV (i.e., 0SX157A),
and basin bypass MOV (i.e., 0SX162B). The team reviewed the calculations that
determined minimum terminal voltages for these MOVs to assess consistency
with the associated MOV thrust calculations. The team also reviewed the
thermal overload sizing calculations for these MOV circuits to assess their
protection against premature thermal overload trip and the minimum voltage
calculations for the 120 volts alternating current (VAC) service to the SXCT basin
level control system to assess the availability of the voltage needed for the level
instrumentation under design basis conditions. To evaluate whether there were
adverse operating trends and to assess the licensees ability to evaluate and
correct problems, the team reviewed system health reports, selected corrective
action documents, and PM procedures and records for the MCC.
b.
Findings
(1) Question Regarding the Maximum Wet Bulb Temperature Value Assumed in the
Emergency Service Water Cooling Tower Tornado Analysis
Introduction: The team identified an unresolved item (URI) regarding the maximum
wet-bulb temperature value assumed in the SXCT tornado analysis. Specifically, the
team noted the analysis used a value which was less restrictive than the highest 3-hour
wet-bulb temperature recorded for the site as described in the UFSAR.
Description: In Section 3.5.4 of the UFSAR, Analysis of Missiles Generated by a
Tornado, stated that, An analysis of the UHS cooling capability for a tornado missile
event has been made. It also stated that, A maximum outside air wet-bulb temperature
of 78 degrees Fahrenheit is assumed and is conservatively held constant throughout the
transient. In addition, this UFSAR section stated that, The analysis was performed
using service water cooling tower performance curves generated using the method
described in UFSAR Section 9.2.5.3.1.1.2 [...]. The analysis of the UHS cooling
capability for a tornado missile event was calculation BYR09-002, UHS Capability with
Loss of SX [Emergency Service Water] Fans due to a Tornado Event, which used a
constant maximum outside air wet-bulb temperature value of 78 degrees Fahrenheit
consistent with UFSAR Section 3.5.4.
However, the team noted the assumed maximum outside air wet-bulb temperature value
of 78 degrees Fahrenheit appeared to be inconsistent with the method described in
UFSAR Section 9.2.5.3.1.1.2, Steady State Tower Performance Analysis. Specifically,
it stated that, The design wet-bulb temperature during warm weather operation is
82 degrees Fahrenheit (Refer to UFSAR Section 2.3.1.2.4). In Section 2.3.1.2.4 of
the UFSAR, Ultimate Heat Sink Design, stated that, This analysis [described in
Section 9.2.5.3.1.1] includes scenarios with the highest 3-hour wet-bulb temperature,
82 degrees Fahrenheit, which was recorded on July 30, 1961, at 3:00 pm. This UFSAR
section also stated that, Per Regulatory Guide 1.27, the ultimate heat sink must be
capable of performing its cooling function during the design basis event for this worst
case 3-hour wet-bulb temperature. In addition, it stated, However, the design
operating wet-bulb temperature of the ultimate heat sink is 78 degrees Fahrenheit
(ASHRAE 1 percent exceedance value).
14
This issue is unresolved pending further review by the Office of Nuclear Reactor
Regulation (NRR) of the licensing basis related to the wet-bulb temperature value
applicable for the SXCT tornado analysis, and the team determination of further NRC
actions to resolve the issue. (URI 05000454/2015008-01; 05000455/2015008-01,
Question Regarding the Maximum Wet-Bulb Temperature Value Assumed in the SXCT
Tornado Analysis)
(2) Maximum Wet-Bulb Temperature Value Assumed in Emergency Service Water Cooling
Tower Analysis Was Not Monitored
Introduction: The team identified an URI regarding the lack of monitoring the maximum
wet-bulb temperature value assumed in SXCT analysis. Specifically, the team noted the
maximum wet-bulb temperature value was a critical parameter for the SXCT analyses,
but the licensee had not established a testing program to verify actual values were
bounded.
Description: In Section 3.5.4 of the UFSAR, Analysis of Missiles Generated by a
Tornado, stated that, An analysis of the UHS cooling capability for a tornado missile
event has been made. It also stated that, A maximum outside air wet-bulb temperature
of 78 degrees Fahrenheit is assumed, and is conservatively held constant throughout
the transient.
In Section 9.2.5.3.1.1 of the UFSAR, Design Basis Reconstitution, stated that,
The design basis event for the Byron ultimate heat sink is a LOCA coincident with a
loss-of-off-site power (LOOP) in one unit, and the concurrent orderly shutdown from
maximum power to cold shutdown of the other unit using normal shutdown operating
procedures. It also stated that, The design wet-bulb temperature during warm
weather operation is 82 degrees Fahrenheit (Refer to the UFSAR Section 2.3.1.2.4).
In Section 2.3.1.2.4 of the UFSAR, Ultimate Heat Sink Design, stated that, This
analysis [described in Section 9.2.5.3.1.1] includes scenarios with the highest 3-hour
wet-bulb temperature, 82 degrees Fahrenheit, which was recorded on July 30, 1961, at
3:00 pm.
The analysis of the UHS cooling capability for a tornado missile event was calculation
BYR09-002, UHS Capability with Loss of SX Fans due to a Tornado Event, which used
a constant maximum outside air wet-bulb temperature value of 78 degrees Fahrenheit
consistent with UFSAR Section 3.5.4. The analysis of the UHS cooling capability for a
LOCA coincident with a LOOP was calculation UHS-01, Ultimate Heat Sink Design
Basis LOCA Single Failure Scenarios, which used a constant maximum outside air
wet-bulb temperature value of 82 degrees Fahrenheit consistent with the UFSAR
Section 9.2.5.3.1.1.
However, the licensee had not established a testing program to verify actual
environmental conditions were bounded by these analyses and design basis limits.
In response to the team questions, the licensee stated that this approach was
acceptable because historical data showed wet-bulb temperature had a cyclic nature,
maximum wet-bulb temperature lasted for relatively short durations, and the analyses
assumed constant wet-bulb temperature values.
15
This issue is unresolved pending further NRR review of the acceptability of the
licensee approach to ensure the SXCT analyses bounded actual environmental
conditions, and the team determination of further NRC actions to resolve the issue.
(URI 05000454/2015008-02; 05000455/2015008-02, Maximum Wet-Bulb Temperature
Value Assumed in SXCT Analysis Was Not Monitored)
.4
Operating Experience
a.
Inspection Scope
The team reviewed four operating experience issues (samples) to ensure that NRC
generic concerns had been adequately evaluated and addressed by the licensee.
The operating experience issues listed below were reviewed as part of this inspection:
IN 2013-05, Battery Expected Life and Its Potential Impact on Surveillance
Requirements;
IN 2010-26, Submerged Electrical Cables;
IN 2013-12, Improperly Sloped Instrument Sensing Lines; and
IN 2012-01, Refueling Water Storage Tank Degradation.
b.
Findings
No findings were identified.
.5
Modifications
a.
Inspection Scope
The team reviewed five permanent plant modifications related to selected risk-significant
components to verify that the design bases, licensing bases, and performance capability
of the components had not been degraded through modifications. The modifications
listed below were reviewed as part of this inspection effort:
Engineering Change (EC) 385951, Multiple Spurious Operation - Scenario 14,
1SI8811A/B;
EC396016, Increase U1 Pressurizer PORV Accumulator Tank Operating
Pressure to Increase Number of PORV Open/Close Cycles from Accumulator;
EC388735, Detailed Review of the FC Purification for Use of Non-Safety
Related Portion Connected to Safety Related Piping;
DRP 11-052, Clarify References to RWST Internal Pressure in the ECCS and
the CS Pumps NPHS Analysis; and
EC385829, Tornado Missile Design Basis for the Essential Service Water
16
b.
Findings
(1) Failure to Evaluate the Adverse Effects of Sharing the Refueling Water Storage Tanks of
Both Reactor Units
Introduction: The team identified a Severity Level IV NCV of 10 CFR 50.59(d)(1),
Changes, Tests, and Experiments, and an associated finding of very-low safety
significance (Green) for the licensees failure to perform a written safety evaluation that
provided the bases for the determination that a change which resulted in the sharing of
the RWSTs of both reactor units did not require a license amendment. Specifically,
screening 6E-05-0172, UFSAR Change Package (DRP)11-052, did not address the
reduction in reactor unit independence associated with sharing the RWSTs air space of
both reactor units.
Description: Each reactor unit has one RWST, which supplies borated water to both
trains of the Emergency Core Cooling System (ECCS) and Containment Spray (CS)
systems during the injection phase of a LOCA recovery. The UFSAR Section 6.3,
Emergency Core Cooling System, and UFSAR Section 6.5.2, Containment Spray
Systems, described the NPSH analyses for the ECCS and CS pumps when their
suctions are aligned to their associated RWST. Before November 16, 2005, these
UFSAR sections described the RWST as being under atmospheric pressure during
the injection mode. The licensee changed these UFSAR descriptions following the
discovery that the RWST would not be under atmospheric pressure because the RWST
vent did not have the capacity to prevent vacuum during the high outflow expected
during the injection phase, and the vent vacuum relief device was not safety related.
This discovery was captured in the CAP as AR00239280.
The licensee reviewed this UFSAR change in Title 10, Code of Federal Regulations
(CFR), Part 50.59 screening 6E-05-0172, Clarify References to RWST Internal
Pressure in the ECCS and CS Pumps NPSH Analysis. The screening concluded that
the change did not require a 10 CFR 50.59 safety evaluation and, consequently, NRC
prior approval because the change did not result in an adverse effect to the ECCS and
CS systems. Specifically, the licensee determined the expected vacuum would not
affect the structural integrity of the tank. In addition, the licensee determined in
calculation BYR 04-016, [Residual Heat Removal] RHR, SI, [Chemical and Volume
Control] CV, and CS Pump NPSH during ECCS Injection Mode, that the available
NPSH for the pumps while taking suction from the RWST remained adequate when
considering the expected vacuum.
However, the team noted that revised calculation BYR 04-016 credited the entire RWST
vent line, which was common to the RWSTs of both reactor units. Consequently, the
change credited the free air space of both tanks to mitigate the vacuum expected during
tank drawdown. The team also noted that UFSAR Section 3.1.2.1.5, Evaluation Against
Criterion 5 - Sharing of Structures, Systems, and Components, described those SSCs
important to safety shared by the two reactor units, and the RWSTs were not included
as shared SSCs. Thus, the team noted the licensee implemented a change to the
facility as described in the UFSAR that resulted in a reduction of reactor unit
independence. Changes to the facility as described in the UFSAR that reduce reactor
unit independence adversely impact 10 CFR 50.59 change evaluation criteria because
they result in more than a minimal increase in the likelihood of occurrence of a
malfunction of an SSC important to safety. Since the licensee failed to appropriately
17
evaluate this adverse effect in a 10 CFR 50.59 safety evaluation, the team could not
reasonably determine that the change would not have ultimately required NRC prior
approval.
The licensee captured this issue in their CAP as AR 02496142. The corrective actions
considered at the time of this inspection were to revise calculation BYR04-016 to not
credit the opposite unit RWSTs air space and/or revise 10 CFR 50.59 screening
6E-05-0172 to consider the implications of crediting the opposite unit RWST air space.
The team also noted the licensee did not correctly implement this change into
associated surveillance procedures intended to verify RWST operability. This separate
concern is discussed in detail in Section 1R21.5.b(2) of this report.
Analysis: The team determined that the failure to provide a written evaluation that
provided the bases for the determination that a change which resulted in the sharing of
the RWSTs of both reactor units did not require a license amendment, was contrary to
the requirements of 10 CFR 50.59(d)(1), and was a performance deficiency. The
performance deficiency was more than minor because it was associated with the
Mitigating Systems cornerstone attribute of design control, and affected the cornerstone
objective of ensuring the availability, reliability, and capability of mitigating systems to
respond to initiating events to prevent undesirable consequences. In addition, it was
associated with the Barrier Integrity cornerstone attribute of design control, and affected
the cornerstone objective of providing reasonable assurance that physical design
barriers protect the public from radionuclide releases caused by accidents or events.
Specifically, the change did not ensure the RWST capability to support ECCS and CS
mitigating and barrier functions because it eliminated the capability to achieve the RWST
supporting function while maintaining separation of the reactor units.
In addition, the associated violation was determined to be more than minor because the
team could not reasonably determine the changes would not have ultimately required
NRC prior approval.
Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process
instead of the Significance Determination Process (SDP) because they are considered
to be violations that potentially impede or impact the regulatory process. This violation is
associated with a finding that has been evaluated by the SD, and communicated with an
SDP color reflective of the safety impact of the deficient licensee performance. The
SDP, however, does not specifically consider the regulatory process impact. Thus,
although related to a common regulatory concern, it is necessary to address the violation
and finding using different processes to correctly reflect both the regulatory importance
of the violation and the safety significance of the associated finding.
In this case, the team determined that the finding could be evaluated using the SDP in
accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination
Process by using Attachment 0609.04, Initial Characterization of Findings. Since the
finding impacted the Mitigating Systems and Barrier Integrity cornerstones, the
inspectors screened the finding through IMC 0609 Appendix A, The Significance
Determination Process for Findings At-Power, using Exhibit 2, Mitigating Systems
Screening Questions, and Exhibit 3, Barrier Integrity Screening Questions. The
finding screened as very-low safety significance (Green) because it did not result in the
loss of operability or functionality, and it did not represent an actual open pathway in the
physical integrity of the reactor containment. Specifically, the licensee reviewed
18
calculation BYR 04-016, and reasonably determined that enough conservatism existed
such that adequate NPSH could be maintained without sharing the RWSTs of both
reactor units.
In accordance with Section 6.1.d of the NRC Enforcement Policy, this violation is
categorized as Severity Level IV because the resulting change was evaluated by the
SDP as having very-low safety significance (i.e., Green finding).
The inspectors did not identify a cross-cutting aspect associated with this finding
because it was confirmed not to be reflective of current performance. Specifically, the
finding occurred approximately 10 years ago.
Enforcement: Title 10 CFR 50.59, Changes, Tests, and Experiments, Section (d)(1)
requires, in part, the licensee to maintain records of changes in the facility, of changes in
procedures, and of tests and experiments made pursuant 10 CFR 50.59(c). These
records must include a written evaluation which provides the bases for the determination
that the change, test, or experiment does not require a license amendment pursuant to
Paragraph (c)(2) of this section. Paragraph (c)(2)(ii) states, in part, that a licensee shall
obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed
change, test, or experiment if the change, test, or experiment would result in more than a
minimal increase in the likelihood of occurrence of a malfunction of an SSC important to
safety previously evaluated in the UFSAR. In the UFSAR Sections 6.3 and 6.5 describe
the NPSH evaluations for ECCS and CS pumps when their suctions are aligned to their
associated RWST. Additionally, UFSAR Section 3.1.2.1.5 states that Those systems,
structures, and components important to safety shared by the two units are the ultimate
heat sinks and the associated Byron makeup water systems; various heating, ventilating,
and air conditioning systems within the shared auxiliary and fuel handling building; and a
component cooling heat exchanger which can be valved to serve one unit or the other.
The RWSTs are not included as shared SSCs.
Contrary to the above, on November 16, 2005, the licensee failed to maintain a record
of a change in the facility made pursuant to 10 CFR 50.59(c) that included a written
evaluation which provided the bases for the determination that the change did not
require a license amendment pursuant to 10 CFR 50.90(c)(2). Specifically, the licensee
changed the ECCS and CS pumps NPSH calculation for their injection mode of
operation (i.e., calculation BYR 04-016) to credit the entire vent line common to the
RWSTs of both reactor units and, consequently, the free air space of both tanks to
mitigate the vacuum expected during tank drawdown. However, the licensee failed to
perform a written evaluation that provided the bases for the determination that the
change effect of reducing reactor unit independence by sharing their RWSTs did not
result in more than a minimal increase in the likelihood of occurrence of a malfunction of
the RWSTs and their supported safety systems.
The licensee is still evaluating its planned corrective actions. However, the team
determined that the continued non-compliance does not present an immediate safety
concern because the licensee reasonably determined that the affected analysis
contained enough conservatism such that adequate NPSH could be maintained
without sharing the RWSTs of both reactor units.
19
Because this was a Severity Level IV violation and was entered into the licensee
Corrective Action Program (CAP) as AR 02496142, this violation is being treated
as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy.
(NCV 05000454/2015008-03; 05000455/2015008-03; Failure to Evaluate the Adverse
Effects of Sharing the RWSTs of Both Reactor Units)
The associated finding is evaluated separately from the traditional enforcement violation
and, therefore, the finding is being assigned a separate tracking number.
(FIN 05000454/2015008-04; 05000455/2015008-04; Failure to Evaluate the Adverse
Effects of Sharing the RWSTs of Both Reactor Units)
(2) Failure to Adequately Implement a Design Change Associated with the RWSTs
Introduction: The team identified a finding of very-low safety significance (Green),
and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control,
for the licensees failure to translate applicable design basis into TS Surveillance
Requirement (SR) 3.5.4.2 implementing procedures. Specifically, these procedures
did not verify RWST vent line was free of ice blockage at the locations and during all
applicable MODEs of reactor operation assumed by the ECCS and CS pump NPSH
calculation.
Description: Each reactor unit has one RWST, which supplies borated water to both
trains of the ECCS and CS systems during the injection phase of a LOCA recovery.
The TS 3.5.4, Refueling Water Storage Tank, required the RWSTs to be operable
when their associated reactor unit is in MODEs 1, 2, 3, or 4. A vent line is installed at
the top of each RWST. The vent lines are routed into the auxiliary building where they
connect to a common header which joins to a filtration system. Because the header is
common to both vents, the free air spaces of the RWSTs are communicated via their
vent lines. The vent line portions located between the tanks and the auxiliary building
are exposed to outside ambient conditions. For this reason, TS SR 3.5.4.2 stated,
Verify RWST vent path temperature is 35 degrees Fahrenheit. The associated TS
Basis explained that Heat traced portions of the RWST vent path should be verified to
be within the temperature limit needed to prevent ice blockage and subsequent vacuum
formation in the tank during rapid level decreases caused by accident conditions. The
licensee established procedures 1/2 BOSR 01-1,2,3, Modes 1, 2, and 3 Shiftily and
Daily Operating Surveillance, and 1/2 BOSR 01-4, Mode 4 Shiftily and Daily Operating
Surveillance, as the implementing procedures for SR 3.5.4.2.
Originally, the RWSTs design assumed they were atmospheric tanks by crediting their
associated vent line capability to prevent vacuum during tank drawdown. However, on
November 16, 2005, the licensee implemented a design change to credit the vent lines
capability to communicate the free air space of both tanks following the discovery that
the RWST vents did not have the capacity to prevent vacuum during the high outflow
expected during the injection phase, and the vent vacuum relief devices were not safety
related. This discovery was captured in the CAP as AR00239280.
As a result, calculation BYR 04-016, RHR, SI, CV and CS Pump NPSH during ECCS
Injection Mode, credited the vent lines of both RWSTs to mitigate the vacuum expected
during the drawdown of one tank during accident conditions. However, the team noted
this change was not correctly implemented into procedures 1/2 BOSR 01-1,2,3 and
1/2 BOSR 01-4. Specifically, these procedures were reactor unit specific in that their
instructions only required verifying the RWST vent line portions that were associated
20
with the applicable reactor unit RWST; that is, the portions between the associated
RWST and the auxiliary building. As a consequence, the team was concerned because,
if one vent line is found to be blocked with ice, the procedures would only recognize one
RWST as being inoperable. In addition, the procedures were only implemented when
the associated reactor unit was in MODEs 1, 2, 3, or 4 consistent with the applicability
requirements of TS 3.5.4. Thus, the team was also concerned that a potentially
inoperable condition would not be detected because the procedures would not verify
both vent lines were free of ice blockage when one reaction unit is in MODE 5 or 6
while the other reactor unit is in MODE 1, 2, 3, or 4.
The licensee captured the team concerns in their CAP as AR 02496766. The immediate
corrective action was to verify that outside air temperatures were not forecasted to fall
below 35 degrees Fahrenheit for the foreseeable future. Additionally, the licensee
determined the RWSTs remained operable during the last 3 years by performing a
historical review which did not find instances in which the vent lines temperature fell
below 35 degrees Fahrenheit. The proposed corrective actions to restore compliance
at the time of this inspection included revising the applicable calculations to remove
dependence on the opposite unit, and/or revising the affected procedures to be
consistent with the applicable calculation.
The team also noted the licensee did not perform a written safety evaluation that
provided the bases for the determination that this change, which resulted in a reduction
of reactor unit independence, did not require a license amendment. This separate
concern is discussed in detail in Section 1R21.5.b(1) of this report.
Analysis: The team determined the failure to translate applicable design basis into
TS SR 3.5.4.2 implementing procedures was contrary to 10 CFR Part 50, Appendix B,
Criterion III, Design Control, and was a performance deficiency. The performance
deficiency was determined to be more than minor because it was associated with the
Mitigating Systems cornerstone attribute of design control, and affected the cornerstone
objective of ensuring the availability, reliability, and capability of mitigating systems to
respond to initiating events to prevent undesirable consequences. Additionally, it was
associated with the Barrier Integrity cornerstone attribute of design control, and affected
the cornerstone objective of providing reasonable assurance that physical design
barriers protect the public from radionuclide releases caused by accidents or events.
Specifically, TS SR 3.5.4.2 implementing procedures were inadequate to verify RWST
operability because they did not verify all critical assumptions made by the design
calculations. The RWST supports ECCS, which is a mitigating system, and CS, which
is part of the physical design barrier.
The team determined the finding could be evaluated using the SDP in accordance
with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial
Characterization of Findings. Since the finding impacted the Mitigating Systems and
Barrier Integrity cornerstones, the inspectors screened the finding through IMC 0609,
Appendix A, The Significance Determination Process for Findings At-Power, using
Exhibit 2, Mitigating Systems Screening Questions, and Exhibit 3, Barrier Integrity
Screening Questions. The finding screened as very-low safety significance (Green)
because it did not result in the loss of operability or functionality, and it did not represent
an actual open pathway in the physical integrity of reactor containment. Specifically, the
licensee performed a historical review of the last 3 years of operation and did not find
any instances in which the vent path temperature fell below 35 degrees Fahrenheit.
21
The inspectors did not identify a cross-cutting aspect associated with this finding
because it was confirmed not to be reflective of current performance due to the age
of the performance deficiency. Specifically, the finding occurred approximately
10-years ago.
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires,
in part, that design changes, including field changes, be subjected to design control
measures commensurate with those applied to the original design.
Contrary to the above, on November 16, 2005, the licensee performed a design change
and failed to subject it to design control measures commensurate to those applied to the
original design. Specifically, the licensee changed the ECCS and CS pump NPSH
calculation for their injection mode of operation (i.e., calculation BYR 04-016) to credit
the capability of the vent lines of both RWSTs to support the operability of any one
RWST. However, the design control measures failed to correctly translate the new
design basis into procedures 1/2 BOSR 01-1,2,3 and 1/2 BOSR 01-4 in that they were
not revised to verify the capability of the vent lines of both RWSTs to support the
operability of any one RWST.
The licensee is still evaluating its planned corrective actions. However, the team
determined that the continued non-compliance does not present an immediate safety
concern because outside air temperatures were not forecasted to fall below 35 degrees
Fahrenheit for the foreseeable future. Additionally, a corrective action tracking item was
created to develop compensatory actions if compliance is not restored prior to the next
season when temperatures can potentially decrease below 35 degrees Fahrenheit.
Because this violation was of very-low safety significance and was entered into the
licensees CAP as AR 02496766, this violation is being treated as a NCV, consistent
with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000454/2015008-05; 05000455/2015008-05; Failure to Adequately Implement a Design Change Associated
with the RWSTs)
(3) Failure to Evaluate the Adverse Effects of Changing the Emergency Service Water
Cooling Tower Tornado Analysis as Described in the Updated Final Safety Analysis
Report
Introduction: The team identified a Severity Level IV NCV of 10 CFR 50.59(d)(1),
Changes, Tests, and Experiments, and an associated finding of very-low safety
significance (Green) for the licensees failure to perform a written evaluation that
provided the bases for the determination that the changes to the SXCT tornado analysis
as described in the UFSAR did not require a license amendment. Specifically,
50.59 Evaluation 6G-11-0041, Tornado Missile Design Basis for the Essential Service
Water Cooling Towers, did not address the introduction of a new failure mode, the
resulting loss of heat removal capacity during worst postulated conditions, and addition
of operator actions that have not been demonstrated can be completed within the
required time to restore the required SXCT heat removal capacity during worst case
conditions.
Description: During the 2005 NRC Safety Systems Design, Performance and
Capability (SSDPC) inspection, the inspectors noted that the UFSAR-described
tornado analysis for the SXCT had not been updated to reflect changes that increased
the heat load. The SSDPC documented this concern as URI 05000454/2005002-07;
22 05000455/2005002-07. In 2007, this URI was subsequently closed to NCV 05000454/
2007004-03;05000455/2007004-03. As a result, on February 14, 2012, the licensee
completed EC 385829, UHS Capability with Loss of SX Fans Due to Tornado Missiles,
to change the UHS tornado missile design basis as described in Revision 7 of the
UFSAR. The EC 385829 evaluated these design basis changes in 10 CFR 50.59 safety
evaluation 6G-11-004, Tornado Missile Design Basis for the Essential Service Water
Cooling Towers, dated February 9, 2012. This 10 CFR 50.59 safety evaluation
concluded that the design basis changes could be implemented without obtaining a
license amendment.
However, the team noted that the licensee did not address the adverse effects of the
changes in the 10 CFR 50.59 safety evaluation. Specifically, the change reduced the
amount of missiles from multiple to single, and changed the SXCT design from
natural draft cooling to mechanical draft cooling (i.e., from passive to active system).
These changes adversely impacted 10 CFR 50.59 change evaluation criteria because
they would result in more than a minimal increase in the likelihood of occurrence of a
malfunction of the SXCT during a tornado event. Specifically:
The change introduced a new failure mode (i.e., fan failures) that was not
bounded by the previous analysis. Specifically, Revision 7 of the UFSAR
Section 3.5.4, Analysis of Multiple Missiles Generated by a Tornado, stated
that the SXCT fans, fan motors, and fan drives were not protected from tornado
missiles. It also stated, An analysis of cooling tower capacity without fans
[emphasis added] has been made. In contrast, this statement was revised to,
An analysis of the UHS cooling capability for a tornado missile event has been
made. The new analysis required multiple operating fans to ensure enough
cooling capacity to mitigate the effects of a single tornado missile. The fans, fan
motors, and fan drives were not modified to add tornado missile protection. In
addition, Revision 7 of the UFSAR Section 9.2.5.3.2, Essential Service Water
Cooling Towers, stated An analysis of the effect of multiple [emphasis added]
tornado missiles on the essential service water cooling towers has been
performed. This statement was revised to delete the word multiple.
Following this revision, the analysis only considered the effects of one
Revision 1 of NEI 96-07, Guidelines for 10 CFR 50.59 Evaluations, which
has been endorsed by the NRC in Regulatory Guide 1.187, Guidance for
Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, stated, in
part, that a change would result in less than a minimal increase in the likelihood
of occurrence of an SSC malfunction provided it satisfies applicable design
basis requirements. In contrast, this change did not satisfy the design basis
requirements for protection against natural phenomena as described in the
USAR Section 3.1.2.1.2, Evaluation Against Criterion 2 - Design Bases for
Protection Against Natural Phenomena. Specifically, Revision 7 and the
revision in effect at the time of this inspection of UFSAR Section 3.1.2.1.2 stated,
The systems, components, and structures important to safety have been
designed to accommodate, without loss of capability [emphasis added], effects of
the design-basis natural phenomena along with appropriate combinations of
normal and accident conditions. However, this change would result in the loss
of SXCT capability to perform its safety function during the worst case conditions
in that the required number of fans would not be available necessitating operator
23
actions to delay shutdown cooling initiation until an adequate number of SXCT
fans are available to support the shutdown cooling heat load and, consequently,
transition to MODE 5 where design basis accidents (DBAs) are not postulated.
The change involved a new operator action that supports the SXCT function
which is not reflected in plant procedures and training programs. Specifically,
UFSAR Section 3.5.4 was revised to credit new operator actions to delay
RHR initiation until an adequate number of SXCT fans are available for shutdown
cooling [emphasis added] and to stagger RHR initiation for the two units.
The revised UFSAR-described analysis assumed For the worst case design
conditions the first unit is assumed to be placed on RHR cooling 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after
the event and the second unit at 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after the event. NEI 96-07 states, in
part, that a new operator action that supports a design function credited in a
safety analysis results in less than a minimal increase in the likelihood of
occurrence of an SSC malfunction provided the action is reflected in plant
procedures and training programs, and these actions have been demonstrated
can be completed in the time required considering the aggregate effects.
However, the licensee had not created procedures and training material to
restore an adequate number of SXCT fans. In addition, the licensee had not
demonstrated that these actions can be completed in the time required
considering the aggregate effects, such as the expected conditions when the
actions are required.
In addition, the change would create a possibility for an SXCT malfunction with a
different result than any previously evaluated in the UFSAR because:
Nuclear Energy Institute (NEI) 96-07 states, A malfunction that involves an
initiator or failure whose effects are not bounded by those explicitly described in
the UFSAR is a malfunction with a different result. In contrast, this change
would result in the loss of SXCT capability to perform its safety function during
the worst case conditions in that the required number of fans would not be
available to support RHR initiation necessitating a delay of RHR initiation until an
adequate number of fans are available. The previous UFSAR-described analysis
assumed the SXCT design remained capable of performing its safety function
during the worst case conditions because it did not require any fans to support
RHR initiation and operation; and
NEI 96-07 stated, An example of a change that would create the possibility for a
malfunction with a different result is a substantial modification that creates a
new or common cause failure that is not bounded by previous analyses or
evaluations. In contrast, this change introduced a new failure that was not
bounded by previous analysis as previously explained.
The licensee captured the team concern in their CAP as AR 2506214 to request a
license amendment. The potential operability implications of this issue are discussed
in Section 4OA2.1.b(3) of this report.
Analysis: The team determined that the failure to perform a written evaluation that
provided the bases for the determination that the changes to the SXCT tornado analysis
as described in the UFSAR did not require a license amendment was contrary to the
requirements of 10 CFR 50.59(d)(1) and was a performance deficiency. The
24
performance deficiency was more than minor because it was associated with the
Mitigating Systems cornerstone attribute of protection against external events, and
affected the cornerstone objective of ensuring the availability, reliability, and capability of
mitigating systems to respond to initiating events to prevent undesirable consequences.
Specifically, the change did not ensure the SXCT reliability and availability during and
following a tornado event because it introduced a new failure mode, and added reliance
on operator actions that have not been demonstrated can be completed in the required
time. The change also did not ensure the SXCT capability to perform its safety function
during the worst case conditions during and following a tornado event in that the
required number of fans would not be available necessitating timely operator action
to restore the required heat removal capability.
In addition, the associated violation was determined to be more than minor because the
team could not reasonably determine the changes would not have ultimately required
NRC prior approval.
Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process
instead of the SDP because they are considered to be violations that potentially impede
or impact the regulatory process. This violation is associated with a finding that has
been evaluated by the SDP, and communicated with an SDP color reflective of the
safety impact of the deficient licensee performance. The SDP, however, does not
specifically consider the regulatory process impact. Thus, although related to a common
regulatory concern, it is necessary to address the violation and finding using different
processes to correctly reflect both the regulatory importance of the violation and the
safety significance of the associated finding.
In this case, the team determined the finding could be evaluated using the SDP in
accordance with IMC 0609, Significance Determination Process, Attachment 0609.04,
Initial Characterization of Findings. Because the finding impacted the Mitigating
System cornerstone, the team screened the finding through IMC 0609, Appendix A, The
Significance Determination Process for Findings At-Power, using Exhibit 2, "Mitigating
Systems Screening Questions." In accordance with Exhibit 2, the team screened the
finding using Exhibit 4, External Events Screening Questions, because the finding
involved the degradation of equipment or function specifically designed to mitigate a
severe weather initiating event. The team conservatively screened the finding as
necessitating a detailed risk evaluation because the loss of UHS during a tornado event
would degrade one or more trains of a system that supports a risk-significant system or
function.
The Senior Reactor Analysts (SRAs) performed a bounding risk evaluation for the delta
core damage frequency (CDF) of tornado missile strike(s) causing a core damage
event at Byron due to damage to the SXCT fans:
The SRAs assumed that a tornado with wind speed exceeding 100 mph would
be required to generate damaging missiles;
The frequency of this tornado for Byron is approximately 1.13E-4/yr from the Risk
Assessment Standardization Project (RASP) website;
25
The tornado missiles were assumed to cause damage and fail an entire set of
SXCT fans in addition to a set of fans that were initially out of service (i.e., 4 fans
conservative assumption); and
The SRAs further assumed that the tornado also caused a severe weather loss
of offsite power event.
The Byron SPAR Model Version 8.27 and Systems Analysis Programs for Hands-on
Integrated Reliability Evaluations (SAPHIRE) Version 8.1.2 software were used by the
SRAs to evaluate the risk significance of this finding. Using the Byron SPAR model, the
Conditional Core Damage Probability (CCDP) (i.e., if the tornado event occurred and
damaged one train of SXCT fans) is approximately 4.8E-4. Thus, a bounding CDF
calculated due to the SXCT vulnerability to missiles is approximately 5.4E-8/yr
(i.e., 1.13E-4/yr x 4.8E-4 = 5.4E-8/yr).
Based on the detailed risk evaluation, the SRAs determined that the finding was of
very-low safety significance (Green). As a result, this violation is categorized as
Severity Level IV in accordance with Section 6.1.d of the NRC Enforcement Policy.
The team did not identify a cross-cutting aspect associated with this finding because the
finding was not representative of current performance. Specifically, the change was
evaluated through the licensee 50.59 process in February 9, 2012.
Enforcement: Title 10 CFR 50.59, Changes, Tests, and Experiments, Section (d)(1)
requires, in part, the licensee to maintain records of changes in the facility, of changes in
procedures, and of tests and experiments made pursuant 10 CFR 50.59(c). These
records must include a written evaluation which provides the bases for the determination
that the change, test, or experiment does not require a license amendment pursuant to
paragraph (c)(2) of this section. Paragraph (c)(2)(ii) states, in part, that a licensee shall
obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed
change, test, or experiment if the change, test, or experiment would result in more than a
minimal increase in the likelihood of occurrence of a malfunction of an SSC important to
safety previously evaluated in the UFSAR. In addition, 10 CFR(c)(2)(vi) states, in part,
that a licensee shall obtain a license amendment pursuant to 10 CFR 50.90 prior to
implementing a proposed change, test, or experiment if the change, test, or experiment
would create a possibility for a malfunction of an SSC important to safety with a different
result than any previously evaluated in the Final Safety Analysis Report (FSAR)as
updated.
The UFSAR Section 9.2.5.3.2 in effect prior to the change implementation stated, An
analysis of the effect of multiple [emphasis added] tornado missiles on the essential
service water cooling towers has been performed. In addition, UFSAR Sections 3.5.4.1
and 9.2.5.3.2 in effect prior to the change implementation stated, An analysis of cooling
tower capacity without fans [emphasis added] has been made. Moreover, UFSAR
Section 3.1.2.1.2 in effect prior to the change implementation and at the time of this
inspection stated, The systems, components, and structures important to safety have
been designed to accommodate, without loss of capability [emphasis added], effects of
the design-basis natural phenomena along with appropriate combinations of normal and
accident conditions.
26
Contrary to the above, on February 9, 2012, the licensee failed to maintain a record of
a change in the facility made pursuant to 10 CFR 50.59(c) that included a written
evaluation which provided the bases for the determination that the change did not
require a license amendment pursuant to 10 CFR 50.59(c)(2). Specifically, the licensee
made changes to the UFSAR-described SXCT tornado analysis and evaluated this
change in 50.59 Evaluation 6G-11-0041. However, this evaluation did not consider
the adverse effects of the introduction of a new failure mode, the resulting loss of heat
removal capacity during worst postulated conditions, and addition of operator actions
that have not been demonstrated can be completed in the required time to restore the
required SXCT heat removal capacity during worst case conditions. As a result, the
evaluation did not provide a basis for the determination that the change did not result in
a more than a minimal increase in the likelihood of occurrence of a malfunction of the
SXCT during and following a tornado event, and would not create a possibility for a
malfunction of the SXCT with a different result than any previously evaluated.
The licensee is still evaluating its planned corrective actions to restore compliance. As
an immediate corrective action, the licensee performed an operability evaluation. At the
time of the CDBI exit meeting on June 16, 2015, the team was still reviewing the revised
operability evaluation with the assistance of NRR.
Because this was a Severity Level IV violation, and was entered into the licensees CAP
as AR 02506214, this violation is being treated as an NCV, consistent with Section 2.3.2
of the NRC Enforcement Policy. (NCV 05000454/2015008-06; 05000455/2015008-06,
Failure to Evaluate the Adverse Effects of Changing the SXCT Tornado Analysis as
Described in the UFSAR)
The associated finding is evaluated separately from the traditional enforcement violation
and, therefore, the finding is being assigned a separate tracking number.
(FIN 05000454/2015008-07; 05000455/2015008-07, Failure to Evaluate the Adverse
Effects of Changing the SXCT Tornado Analysis as Described in the UFSAR)
.6
Operating Procedure Accident Scenarios
a.
Inspection Scope
The team performed a detailed reviewed of the procedures listed below. The
procedures were chosen because they were associated with feed-and-bleed of the
RCS, a loss of UHS, and other aspects of this inspection. For the procedures listed
time critical operator actions were reviewed for reasonableness, in plant action were
walked down with a licensed operator, and any interfaces with other departments were
evaluated. The procedures were compared to the UFSAR, design assumptions, and
training materials to assess consistency.
The following operating procedures were reviewed in detail:
1BFR-H1, Response to Loss of Secondary Heat Sink Unit1, Revision 203;
0BOA PRI-7, Loss of Ultimate Heat Sink Unit 0, Revision 1;
1BOA PRI-7, Essential Service Water Malfunction Unit 1, Revision 106;
1BOA PRI-5, Control Room Inaccessibility, Revision 108;
27
1BOA ELEC-5, Local Emergency Control of Safe Shutdown Equipment,
Revision 106;
1BEP ES-1.3, Transfer to Cold Leg Recirculation Unit 1, Revision 204; and
1BCA-1.2, LOCA Outside Containment Unit 1, Revision 200.
b.
Findings
(1) Failure to Provide Proper Direction for Low Level Isolation of the Refueling Water
Storage Tank in Emergency Operating Procedures
Introduction: The team identified a finding of very-low safety significance (Green), and
an associated NCV of TS 5.4, Procedures, for the failure to EOPs for transfer to cold
leg recirculation. Specifically, Revision 204 of EOPs 1/2BEP ES-1.3, Transfer to Cold
Leg Recirculation, did not contain instructions for transferring the ECCS and CS
systems to the recirculation mode that ensured prevention of potential pump damage
when the RWST is emptied following a LOCA.
Description: Procedures 1/2BEP ES-1.3 were established as the implementing EOPs
for transferring ECCS and CS system suction from the RWST to containment sump
recirculation. These EOPs were intended to be consistent with the technical guidelines
of Westinghouse Owners Group Guidelines (WOG) Emergency Response Guideline
(ERG) ES-1.3, Transfer to Cold Leg Recirculation, dated April 30, 2005. The technical
guideline of WOG ERG ES-1.3 included the following caution statement: Any pumps
taking suction from the RWST should be stopped if RWST level decreases to (U.03).
The ERG defined (U.03) as RWST empty alarm set point in plant specific units. It also
stated, Based on pump suction piping configuration, the plant specific value of (U.03)
may need to consider the possibility of vortexing and air entrainment. The ERG basis
for this caution stated, Any pumps taking suction from the RWST must be stopped
when the level in the tank reaches the empty alarm set point in order to prevent loss of
suction flow and potential pump damage. The licensee established 9 percent RWST
level as the empty alarm set point to prevent air-entraining vortices and ensured
adequate pump NPSH.
In 1996, the licensee changed EOPs 1/2BEP ES-1.3 to include a deviation to this ERG
caution. Specifically, the revised EOP caution stated Any pumps taking suction from
the RWST should be stropped if level drops to 9 percent, unless a flow path also exists
from the CNMT [containment] sump. The EOP deviation document stated This will
allow continuing with switchover without securing pumps if an acceptable flow path
exists. It also stated CNMT pressure should isolate the RWST flow path once aligned
to the sump. However, the licensee did not perform any evaluation to support this
rationale.
The team was concerned because the revised caution did not assure to prevent air
entrainment into the piping system to avoid ECCS and CS pump air binding and/or
cavitation leading to potential damage. The licensee captured the team concern in
their CAP as AR 02495580. The immediate corrective action was to create a standing
order instructing operators to secure all pumps aligned to the RWST when it reaches
9 percent level. The proposed corrective actions to restore compliance at the time of
this inspection included performing a detailed engineering analysis of the hydrodynamic
fluid mechanics with a dual suction source option or removing the dual suction source
option.
28
Analysis: The team determined that the failure to maintain an EOP for transfer to cold
leg recirculation was contrary to TS 5.4, Procedures, and was a performance
deficiency. The performance deficiency was determined to be more than minor because
it was associated with the Mitigating Systems cornerstone attribute of procedure quality,
and affected the cornerstone objective of ensuring the availability, reliability, and
capability of mitigating systems to respond to initiating events to prevent undesirable
consequences. In addition, it was associated with the Barrier Integrity cornerstone
attribute of procedure quality, and affected the cornerstone objective of providing
reasonable assurance that physical design barriers protect the public from radionuclide
releases caused by accidents or events. Specifically, failure to maintain an EOP for
transfer to cold leg recirculation does not ensure that air entrainment into the piping
system is prevented. As a consequence, the availability, reliability, and capability of
the ECCS pumps to meet their mitigating function are not ensured. Similarly, the
performance deficiency does not provide reasonable assurance the CS pumps would
remain capable of supporting the reactor containment barrier function.
The team determined the finding could be evaluated using the SDP in accordance
with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial
Characterization of Findings. Because the finding impacted the Mitigating Systems
and Barrier Integrity cornerstones, the team screened the finding through IMC 0609,
Appendix A, The Significance Determination Process for Findings At-Power, using
Exhibit 2, Mitigating Systems Screening Questions, and Exhibit 3, Barrier Integrity
Screening Questions. The finding screened as of very-low safety significance (Green)
because it did not result in the loss of operability or functionality of mitigating systems,
represent an actual open pathway in the physical integrity of reactor containment, and
involved an actual reduction in function of hydrogen igniters in the reactor containment.
Specifically, the incorrect caution would only be used in the event that transfer to sump
recirculation was not completed by 9 percent tank level or if the RWST suction isolation
valves fail to close. With respect to transfer to sump recirculation by 9 percent tank
level, this is a time critical operator action that is tested and verified periodically on the
plant simulator. A review of these simulator test results reasonably determined that
operators reliably complete the transfer to sump recirculation prior to reaching this set
point. With respect to the failure of the RWST suction isolation valves, these valves are
test quarterly to demonstrate operability. A review of these test results for the last
3 years reasonably determined the valves would have isolated the tank when required.
The team did not identify a cross-cutting aspect associated with this finding because the
finding was not representative of current performance. Specifically, the inadequate
caution had been added to 1/2BEP ES-1.3 in 1996.
Enforcement: In TS Section 5.4.1b states, in part, that written procedures shall be
established, implemented, and maintained covering the EOPs required to implement the
requirements of NUREG-0737 and NUREG-0737, Supplement 1, as stated in Generic
Letter (GL) 82-33, Section 7.1. NUREG-0737, Supplement 1, Section 7.1.c, states,
Upgrade EOPs to be consistent with Technical Guidelines and an appropriate
procedure Writers Guide. The applicable technical guideline contained in WOG ERG
ES-1.3, Transfer to Cold Leg Recirculation, dated April 30, 2005, stated, Any pumps
taking suction from the RWST should be stopped if RWST level decreases to (U.03).
The ERG defined (U.03) as RWST empty alarm set point in plant specific units. It also
stated, Based on pump suction piping configuration, the plant specific value of (U.03)
may need to consider the possibility of vortexing and air entrainment.
29
The licensee established Revision 204 of 1/2BEP ES-1.3, Transfer to Cold Leg
Recirculation, as the implementing procedures for WOG ERG ES-1.3 to specify the
actions required for transfer to containment sump recirculation. In addition, the licensee
established 9 percent RWST level as the empty alarm set point, in part, to prevent air
entrainment.
Contrary to the above, between 1996 to at least May 4, 2015, the licensee failed to
maintain a written procedure covering the EOPs required to implement the requirements
of NUREG-0737 and NUREG-0737, Supplement 1, as stated in GL 82-33, Section 7.1.
Specifically, the licensee did not upgrade EOPs 1/2BEP ES-1.3 to be consistent with the
technical guideline contained in WOG ERG ES-1.3 in that the EOPs did not instructed
operators to stop any pumps taking suction from the RWST if level decreases below the
9 percent RWST empty alarm set point when a flow path from the containment sump
existed.
The licensee is still evaluating its planned corrective actions. However, the team
determined that the continued non-compliance does not present an immediate safety
concern because the licensee created a standing order instructing operators to secure
all pumps aligned to the RWST when it reaches 9 percent level.
Because this violation was of very-low safety significance, and was entered into the
licensees CAP as AR 02495580, this violation is being treated as an NCV, consistent
with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000454/2015008-08; 05000455/2015008-08, Failure to Provide Proper Direction for Low Level Isolation of
4.
OTHER ACTIVITIES
4OA2 Identification and Resolution of Problems
.1
Review of Items Entered Into the Corrective Action Program
a.
Inspection Scope
The team reviewed a sample of the selected component problems identified by the
licensee, and entered into the CAP. The team reviewed these issues to verify an
appropriate threshold for identifying issues, and to evaluate the effectiveness of
corrective actions related to design issues. In addition, corrective action documents
written on issues identified during the inspection were reviewed to verify adequate
problem identification and incorporation of the problem into the CAP. The specific
corrective action documents sampled and reviewed by the team are listed in the
attachment to this report.
The team also selected three issues identified during previous CDBIs to verify that
the concern was adequately evaluated and corrective actions were identified and
implemented to resolve the concern, as necessary. The following issues were reviewed:
NCV 05000454/2012007-01; 05000455/2012007-01, Non-Conforming 480/120
VAC Motor Control Contactors;
NCV 05000454/2012007-03; 05000455/2012007-03, Non-Conservative
Calibration Tolerance Limits for Electrical Relay Settings; and
30
NCV 05000454/2012007-05; 05000455/2012007-05, Failure to Provide Means
to Detect Leak in Emergency Core Cooling Flow Path.
b.
Findings
(1) Failure to Promptly Correct an NRC-Identified Non-Cited Violation Associated with the
Capability to Detect and Isolate Emergency Core Cooling System Leakage
Introduction: A finding of very-low safety significance (Green), and an associated cited
violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, was
identified by the team for the failure to correct a condition adverse to quality (CAQ).
Specifically, on June 15, 2012, the NRC issued an NCV for the failure to provide means
to detect and isolate a leak in the ECCS within 30 minutes as described in the UFSAR,
which is a CAQ. As of May 22, 2015, the licensee had not corrected the CAQ.
Description: On June 15, 2012 the NRC identified that the licensee had failed to provide
a means to detect and isolate a leak in the ECCS flow path within 30 minutes, as
described in UFSAR 6.3.2.5, System Reliability. Specifically, UFSAR 6.3.2.5 stated,
in part, that the design of the auxiliary building and related equipment was based upon
handling of leaks up to a maximum of 50 gallons per minute (gpm). In addition, it stated
Means were provided to detect and isolate such leaks in the emergency core cooling
flow path within 30 minutes. The 2012 CDBI team identified that the licensee had failed
to provide a means to detect and isolate an ECCS leak within 30 minutes. This issue
was documented as NCV 05000454/2012007-05; 05000455/2012007-05, Failure to
Provide Means to Detect Leak in ECCS Flow Path, in Inspection Report (IR) 05000454/
2012007; 05000455/2012007.
The licensee captured this NCV in their CAP as AR 01378257 and AR 01398434. The
assigned corrective action tracking item (CA) was AR01378257-04, which stated:
Investigate the bases/sources of the values assigned to the single failure
(50 gpm and 30 minutes), including whether there is a commitment associated.
Create additional corrective actions (CA type) as necessary. If UFSAR change is
determined feasible, include an action to determination of the impact of the leak
duration lasting longer than 30 minutes on flood level inside containment and the
Auxiliary Building.
The CA due date was extended eight times and, eventually, the CA was downgraded to
an action tracking item (ACIT) because the licensee recognized that it did not correct the
issue. Procedure PI-AA-125, Corrective Action Program Procedure, defined ACIT as
Action items that are completed to improve performance, or correct minor problems that
do not represent CAQ. On February 18, 2015, the licensee discovered that a new CA
type assignment was not generated to address the NCV following the AR 01378257-04
downgrade from a CA to an ACIT type. This was inconsistent with step 4.5.2 of
procedure PI-AA-125 in that it required, in part, the creation of a CA for any planned
action necessary to correct a CAQ. This discovery was captured in the CAP as
AR 02454767. The associated CA assignment stated:
Design Engineering will determine if UFSAR section 6.3.2.5 requires revision
using the information provided in IR 01378257 and IR 1398434. If it is concluded
a revision is required, an additional CA to track the change will be created.
31
During this inspection period, the team noted that the actions assigned by this CA were
similar to those of AR 01378257-04, which the licensee had previously determined
did not correct the NCV. The team was concerned because, as of May 22, 2015, the
licensee failed to restore compliance and failed to have objective plans to restore
compliance in a reasonable period following the NRC identification of the NCV on
June 15, 2012.
The licensee captured the teams concern in their CAP as AR 02501454 to promptly
restore compliance. As an immediate corrective action, the licensee reasonably
determined ECCS remained operable by reviewing procedures and calculations.
Specifically, the licensee reasonably determined procedures used when responding to
postulated events would direct operators to detect and isolate an ECCS leak before it
could adversely affect the system mitigating function or result in a radionuclide release
in excess of applicable limits.
Analysis: The team determined that the failure to correct an NRC-identified NCV
associated with the capability to detect and isolate ECCS leakage, which is a CAQ, was
contrary to 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, and was a
performance deficiency. The performance deficiency was determined to be more than
minor because it was associated with the Mitigating Systems cornerstone attribute of
design control, and affected the cornerstone objective of ensuring the availability,
reliability, and capability of mitigating systems to respond to initiating events to prevent
undesirable consequences. In addition, it was associated with the Barrier Integrity
cornerstone attribute of design control, and affected the cornerstone objective of
providing reasonable assurance that physical design barriers protect the public from
radionuclide releases caused by accidents or events. Specifically, the failure to detect
and isolate a leak in the ECCS flow path within 30 minutes could compromise long term
cooling, adversely affecting its capability to mitigate a DBA. In addition, a detection and
isolation time greater than the time assumed by the design basis for an ECCS leak
following an accident would result in greater radionuclide release to the auxiliary
building, and the environment and, thus, does not assure that physical design barriers
protect the public from radionuclide releases caused by accidents or events.
The team determined the finding could be evaluated using the SDP in accordance
with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial
Characterization of Findings. Because the finding impacted the Mitigating Systems
and Barrier Integrity cornerstones, the team screened the finding through IMC 0609,
Appendix A, The Significance Determination Process for Findings At-Power, using
Exhibit 2, Mitigating Systems Screening Questions, and Exhibit 3, Barrier Integrity
Screening Questions. The finding screened as very-low safety significance (Green)
because it did not result in the loss of operability or functionality, and it did not represent
an actual pathway in the physical integrity of reactor containment. Specifically, the
licensee reasonably demonstrated that an ECCS leak could be detected and isolated
before it could adversely affect long-term cooling of the plant.
The team determined that the associated finding had a cross-cutting aspect in the area
of human performance because the licensee did not use a consistent and systematic
approach to make decisions. Specifically, the licensee downgraded the original CA to
an ACIT without creating a new CA, which was inconsistent with the instructions
contained in procedure PI-AA-125. Additionally, when the licensee subsequently
discovered a CA type assignment was not created to address the NCV, the licensee
32
created a CA assignment to track actions that were similar to those tracked by the ACIT,
which was inconsistent with the licensee previous determination that those actions did
not correct the NCV. [H.13]
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action,
states, in part, that measures shall be established to assure that conditions adverse to
quality, such as failures, malfunctions, deficiencies, deviations, defective material and
equipment, and non-conformances are promptly identified and corrected.
Contrary to the above, from June 15, 2012, to at least May 22, 2015, the licensee failed
to correct a CAQ. Specifically, on June 15, 2012, the NRC issued NCV 05000454
/2012007-05;05000455/2012007-05 for the failure to provide means to detect and
isolate a leak in the ECCS within 30 minutes for Byron Station, Units 1 and 2, as
described in UFSAR Section 6.3.2.5, which is a CAQ. As of May 22, 2015, the
licensee had not corrected the CAQ in a reasonable period. Instead, the licensee
created ACTI to develop a plan to correct the CAQ, and the associated due date was
extended at least eight times.
The licensee is still evaluating corrective actions. However, the team determined that
the continued non-compliance does not present an immediate safety concern because
the licensee reasonably demonstrated that a leak could be detected and isolated before
it could adversely affect long-term cooling of the plant or result in a radionuclide release
in excess of applicable limits.
This violation is being cited as described in the Notice, which is enclosed with this IR.
This is consistent with the NRC Enforcement Policy, Section 2.3.2.a.2, which states, in
part, that the licensee must restore compliance within a reasonable period of time (i.e., in
a timeframe commensurate with the significance of the violation) after a violation is
identified. The NRC identified NCV 05000454/2012007-05; 05000455/2012007-05 on
June 15, 2012, and documented it in IR 05000454/2012007. The team determined that
the licensee failed to restore compliance within a reasonable time following issuance of
this NCV and failed to have objective plans to restore compliance. (VIO 05000454
/2015008-09;05000455/2015008-09, Failure to Promptly Correct an NRC-Identified
NCV Associated with the Capability to Detect and Isolate ECCS Leakage)
(2) Failure to Maintain the Accuracy of the Instrument Loops Used to Verify Compliance with
the Containment Average Air Temperature Technical Specification Limit
Introduction: The team identified a finding of very-low safety significance (Green), and
an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions,
Procedures, and Drawings, for the failure to have procedures to maintain the accuracy
within the necessary limits of instrument loops used to verify compliance with the
containment average air temperature TS limit of 120 degrees Farhenheit. Specifically,
in 2007, the licensee cancelled the periodic PMs intended to maintain the instrument
accuracy necessary for verifying compliance with the limiting condition for operation
(LCO) of TS 3.6.5, Containment Air Temperature.
Description: The team reviewed selected corrective action documents initiated by the
licensee as a result of their recent Focused Self-Assessment titled, Readiness Review
for 2015 NRC Component Design Basis Inspection. The reviewed corrective action
document sample included AR 02437973. This corrective action document was initiated
on January 15, 2015, in part, for the discovery that the four instrument loops used for
33
determining containment average air temperature (i.e., loops 1/2VP-030, 1/2VP-031,
1/2VP-032, and 1/2VP-033) were removed from the PM Program in 2007 via Service
Request 47654. The corrective action document also noted that the PMs were last
performed in 2001 for 1VP-030; 2002 for 1/2VP-031, 1/2VP-032, 2VP-030, and
2VP-033; and 2009 for 1VP-033.
This corrective action document created an ACIT to determine if the PMs should be
reestablished. Procedure PI-AA-125, Corrective Action Program Procedure, defined
ACIT as Action items that are completed to improve performance, or correct minor
problems that do not represent CAQ. On March 3, 2015, the ACIT concluded that there
was no need to reestablish the PMs due to the instrument loop reliability, previous
calibration history, loop design, redundancy, and daily monitoring which the licensee
believed would notice instrument drift. However, the team noted that TS SR 3.6.5.1
required verifying containment air temperature is less than 120 degrees Fahrenheit
by averaging the instrument readings and, thus, instrument reading variability was
expected. In addition, the team noted the licensee had not established a variability limit
(i.e., acceptance criteria) among the instrument loops and relied on operator judgment to
identify adverse drifts.
The team was concerned because these instrument loops were not maintained to
ensure their accuracy was within the necessary limits to verify compliance with the
containment average air temperature TS limit of 120 degrees Fahrenheit. Containment
average air temperature is an initial condition used in DBA analyses, and is an important
consideration in establishing the containment environmental qualification operating
envelope for both pressure and temperature. This TS limit ensures that initial conditions
assumed in these analyses are met during unit operations.
The licensee captured the teams concern in their CAP as AR 02502846. As an
immediate corrective action, the licensee reasonably established that the 120 degrees
Fahrenheit limit was not exceeded by reviewing applicable historical records from 2002
to time of this inspection. The proposed corrective action to restore compliance at the
time of this inspection was to reconstitute PM procedures for these instrument loops to
assure they are maintained.
Analysis: The team determined that the failure to have procedures to maintain the
accuracy within necessary limits of the instrument loops used during SR 3.6.5.1 was
contrary to 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and
Drawings, and was a performance deficiency. The performance deficiency was
determined to be more than minor because it was associated with the configuration
control attribute of the Barrier Integrity Cornerstone, and adversely affected the
cornerstone objective to ensure that physical design barriers protect the public from
radionuclide releases caused by accidents or events. Specifically, the failure to have
procedures to maintain the accuracy of the containment air temperature instrumentation
loops within necessary limits does not ensure the instrument loop accuracy is
maintained such that SR 3.6.5.1 activities are effective at verifying compliance with the
containment average air temperature TS limit. As a result, the potential exists for an
inoperable condition to go undetected.
The team determined the finding could be evaluated using the SDP in accordance
with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial
Characterization of Findings. Because the finding impacted the Barrier Integrity
34
cornerstone, the team screened the finding through IMC 0609, Appendix A, The
Significance Determination Process for Findings At-Power, using Exhibit 3, Barrier
Integrity Screening Questions. The finding screened as of very-low safety significance
(Green) because it did not represent an actual open pathway in the physical integrity of
reactor containment or involved an actual reduction in hydrogen igniter function.
Specifically, the containment integrity remained intact and the finding did not impact
the hydrogen igniter function.
The team determined that this finding had a cross-cutting aspect in the area of problem
identification and resolution because the licensee did not identify issues completely and
accurately in accordance with the CAP. Specifically, on January 15, 2015, the licensee
captured the lack of periodic PM activities for the containment air temperature instrument
loops in the CAP. However, the licensee failed to completely and accurately identify the
issue in that it was not treated as a CAQ. As a consequence, no corrective actions were
implemented. [P.1]
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures,
and Drawings, requires, in part, that activities affecting quality be prescribed by
documented procedures of a type appropriate to the circumstances and be
accomplished in accordance with these procedures.
Contrary to the above, since 2007 to at least May 22, 2015, the licensee failed to have a
procedure for maintaining the accuracy within the necessary limits of the instrument
loops used while implementing SR 3.6.5.1. Specifically, in 2007, the licensee cancelled
the PMs intended to maintain the instrument loops accuracy necessary for verifying
compliance with LCO 3.6.5 limit.
The licensee is still evaluating its planned corrective actions. However, the team
determined that the continued non-compliance does not present an immediate safety
concern because containment average air temperature readings were significantly lower
than the associated TS limit, and are reasonably expected to maintain that margin in the
foreseeable future based on past performance.
Because this violation was of very-low safety significance, and was entered into the
licensees CAP as AR 02502846, this violation is being treated as an NCV, consistent
with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000454/2015008-10; 05000455/2015008-10, Failure to Maintain the Instrument Loops Used to Verify
Compliance with the Containment Average Air Temperature TS Limit)
(3) Operability Evaluation Relied on Probabilities of Occurrence of the Associated Event
Introduction: The team identified a finding of very-low safety significance (Green),
and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions,
Procedures, and Drawings, for the failure to make an operability determination without
relying on the use of probabilistic tools. Specifically, an operability evaluation related to
an SXCT degraded condition used probabilities of occurrence of tornado events which
was contrary to the requirements of Revision 16 of procedure OP-AA-108-115,
Description: Revision 7 of UFSAR Section 3.5.4, Analysis of Multiple Missiles
Generated by a Tornado, stated that the SXCT fans, fan motors, and fan drives were
not protected from tornado missiles. It also stated that An analysis of cooling tower
35
capacity without fans has been made. In addition, it stated that Using the most
conservative design conditions, it is predicted if the plant is shut down under non-LOCA
conditions with loss of offsite power, the temperature of the service water supplied to
the plant will not exceed 110 degrees Farhernheit. However, during the 2005 NRC
SSDPC inspection, the inspectors noted that this analysis had not been updated to
reflect changes that increased the heat load. The SSDPC documented this concern as
URI 05000454/2005002-07; 05000455/2005002-07. In 2007, this URI was subsequently
closed to NCV 05000454/2007004-03; 05000455/2007004-03. As a result, on February
14, 2012, the licensee completed EC 385829, UHS Capability with Loss of SX Fans
Due to Tornado Missiles, to change the UHS tornado missile design basis to require
a minimum of two SXCT fans and motors for cooling following a tornado event. The
change did not include adding tornado protection to the fans, fan motors, and fan drives.
On August 9, 2013, the licensee initiated corrective action document IR 01545153 for
the NRC discovery that the associated written safety evaluation intended to provide
the bases for the determination that this change did not require a license amendment
failed to consider the change adverse effects. On August 14, 2013, the licensee
initiated corrective action document AR 1546621 to address the associated technical
implications. This corrective action document resulted in Revision 0 of Operability
Evaluation 13-007, Ultimate Heat Sink Capability with Loss of Essential Service Water
Cooling Tower Fans, intended to reasonably demonstrate UHS operability until
corrective actions to restore compliance were implemented.
During this inspection period, the CDBI team noted that Operability Evaluation 13-007
relied on the probability of occurrence of a tornado. Specifically, it stated The UHS is
capable of providing the required cooling because, given a tornado strike under the
design conditions in the UFSAR, the probability of occurrence is less than the
acceptance criteria of 10E-7 /year in SRP 2.2.3. It also stated that The software used
to determine the missile hit probability is called [Tornado Missile Risk Evaluation
Methodology] TORMIS. In addition, it stated that The software uses site specific
factors such as predicted tornado characteristics, tornado occurrence rates, building
layout, potential missile sources and types, missile distribution and the number of
potential missiles. The supporting analysis used the UFSAR Section 2.3.1.2.2,
Tornadoes and Severe Winds. tornado probability of occurrence value of 21E-4 per
year.
Procedure OP-AA-108-115, Operability Determinations, Section 4.5.13, Use of PRA,
stated:
PRA is a valuable tool for evaluating accident scenarios because it can consider
the probabilities of occurrence of accidents or external events. Nevertheless, the
definition of operability is that the SSC must be capable of performing its
specified function or functions, which inherently assumes that the event occurs
and that the safety function or functions can be performed. Therefore, the use of
PRA or probabilities of occurrence of accidents or external events is not
consistent with the assumption that the event occurs, and is not acceptable for
making operability decisions.
Thus, the team determined that the use of TORMIS, the probability for occurrence of
tornados, and the probabilities of missile strikes was not acceptable and contrary to
licensee procedure OP-AA-108-115. The team, in consultation with NRR, also
36
determined that this procedure requirement was consistent with Attachment C.06 of
NRC IMC 0326, Operability Determinations & Functionality Assessments for Conditions
Adverse to Quality or Safety, which was established to assist NRC inspectors review of
licensee determinations of operability and resolution of degraded or nonconforming
conditions.
In addition, the team noted that Byron had not obtained NRC approval for the site
specific use of TORMIS as stated in Regulatory Issue Summary (RIS) 2008-14, Use of
TORMIS Computer Code for Assessment of Tornado Missile Protection. Specifically,
the RIS stated that The initial use of the TORMIS methodology as described in this
RIS requires a license amendment in accordance with 10 CFR 50.59(c)(2)(viii) and
subsequent revision to the plant licensing basis because it is a Departure from the
method of evaluation described in the FSAR, as updated, used in establishing the
design bases or in the safety analysis as defined in 10 CFR 50.59(a)(2).
The team was concerned because Operability Evaluation 13-007 did not reasonably
demonstrate the degraded UHS would be capable of performing its function following a
tornado event. The licensee captured the team concern in their CAP as AR 2504624 to
revise Operability Evaluation 13-007 without using PRA tools.
Analysis: The team determined that the failure to make an operability determination
without relying on the use of probabilistic tools was contrary to licensee procedure
OP-AA-108-115 and was a performance deficiency. The performance deficiency was
determined to be more than minor because it was associated with the Mitigating
Systems cornerstone attribute of protection against external events, and affected the
cornerstone objective of ensuring the availability, reliability, and capability of mitigating
systems to respond to initiating events to prevent undesirable consequences.
Specifically, failure to perform an adequate operability evaluation does not ensure the
SXCT would remain capable of performing its safety function, and had the potential to
allow an inoperable condition to go undetected.
The team determined the finding could be evaluated using the SDP in accordance
with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial
Characterization of Findings. Because the finding impacted the Mitigating System
cornerstone, the team screened the finding through IMC 0609, Appendix A, The
Significance Determination Process for Findings At-Power, using Exhibit 2, "Mitigating
Systems Screening Questions." In accordance with Exhibit 2, the team screened the
finding using Exhibit 4, External Events Screening Questions, because the finding
involved the degradation of equipment or function specifically designed to mitigate a
severe weather initiating event. The team conservatively screened the finding as
necessitating a detailed risk evaluation because the loss of UHS during a tornado event
would degrade one or more trains of a system that supports a risk-significant system or
function.
The SRAs performed a bounding risk evaluation for the CDF of tornado missile
strike(s) causing a core damage event at Byron due to damage to the SXCT fans:
The SRAs assumed that a tornado with wind speed exceeding 100 mph would
be required to generate damaging missiles.
The frequency of this tornado for Byron is approximately 1.13E-4/yr from the
RASP website;
37
The tornado missiles were assumed to cause damage and fail an entire set of
SXCT fans in addition to a set of fans that were initially out of service (i.e., 4 fans
- conservative assumption); and
The SRAs further assumed that the tornado also caused a severe weather loss
of offsite power event.
The Byron SPAR Model Version 8.27 and SAPHIRE Version 8.1.2 software were used
by the SRAs to evaluate the risk significance of this finding. Using the Byron SPAR
model, the CCDP (i.e., if the tornado event occurred and damaged one train of SXCT
fans) is approximately 4.8E-4. Thus, a bounding CDF calculated due to the SXCT
vulnerability to missiles is approximately 5.4E-8/yr (i.e., 1.13E-4/yr x 4.8E-4 = 5.4E-8/yr).
Based on the detailed risk evaluation, the SRAs determined that the finding was of very
low safety significance (Green).
The team determined that this finding had a cross-cutting aspect in the area of human
performance because the licensee did not ensure knowledge transfer to maintain a
knowledgeable and technically competent workforce. Specifically, the licensee did not
ensure personnel were trained on the prohibition of the use of probabilities of occurrence
of an event when performing operability evaluations, which was contained in procedure
OP-AA-108-115. [H.9]
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures,
and Drawings, requires, in part, that activities affecting quality be prescribed by
documented procedures of a type appropriate to the circumstances and be
accomplished in accordance with these procedures.
The licensee established Revision 16 of procedure OP-AA-108-115, Operability
Determinations, as the implementing procedure for assessing operability of SSCs, an
activity affecting quality. Section 4.5.13, Use of Probabilistic Risk Assessment, stated
[] the use of PRA or probabilities of occurrence of accidents or external events is not
consistent with the assumption that the event occurs, and is not acceptable for making
operability decisions.
Contrary to the above, on August 20, 2013, the licensee failed to follow Section 4.5.13 of
procedure OP-AA-108-115. Specifically, the licensee used a PRA tool (i.e., TORMIS)
and probabilities of occurrence of an external event (i.e., tornado) when making an
operability decision related to the SXCT degradation when mitigating tornado events.
Establishing a reasonable expectation of operability is an activity affecting quality.
As an immediate corrective action, the licensee revised the affected operability
evaluation without using PRA tools. At the time of the CDBI exit meeting on
June 16, 2015, the team was still reviewing the revised operability evaluation with
the assistance of NRR.
Because this violation was of very-low safety significance and was entered into the
licensees CAP as AR 2504624, this violation is being treated as an NCV, consistent
with Section 2.3.2 of the NRC Enforcement Policy. (NCV 05000454/2015008-11; 05000455/2015008-11, Operability Evaluation Relied on Probabilities of Occurrence of
the Associated Event)
38
4OA6 Management Meetings
.1
Interim Exit Meeting Summary
On May 22, 2015, the team presented the inspection results to Mr. R. Kearney, and
other members of the licensee staff. The licensee acknowledged the issues presented.
The inspectors had outstanding questions that required additional review and a follow-up
exit meeting.
.2
Exit Meeting Summary
On June 16, 2015, the team presented the inspection results to Mr. B. Currier, and other
members of the licensee staff. The licensee acknowledged the issues presented. The
team asked the licensee whether any materials examined during the inspection should
be considered proprietary. Several documents reviewed by the team were considered
proprietary information and were either returned to the licensee or handled in
accordance with NRC policy on proprietary information.
ATTACHMENT: SUPPLEMENTAL INFORMATION
Attachment
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
R. Kearney, Site Vice President
T. Chalmers, Plant Manager
C. Keller, Engineering Director
B. Currier, Senior Manager of Design Engineering
D. Spitzer, Regulatory Assurance Manager
J. Cunzeman, Mechanical/Structural Design Manager
A. Corrigan, NRC Coordinator
U.S. Nuclear Regulatory Commission
C. Lipa, Chief, Engineering Branch 2
J. Ellegood, Chief, Reactor Projects Branch 3 (Acting)
N. Féliz Adorno, Senior Reactor Inspector
C. Zoia, Senior Resident Inspector (Acting)
J. Draper, Resident Inspector
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened 05000454/2015008-01; 05000455/2015008-01
Question Regarding the Maximum Wet Bulb
Temperature Value Assumed in the SXCT Tornado
Analysis (Section 1R21.3.b(1))05000454/2015008-02; 05000455/2015008-02
Maximum Wet Bulb Temperature Value Assumed in
SXCT Analysis Was Not Monitored
(Section 1R21.3.b(2))05000454/2015008-03; 05000455/2015008-03
Failure to Evaluate the Adverse Effects of Sharing the
RWSTs of Both Reactor Units (Section 1R21.5.b(1))05000454/2015008-04; 05000455/2015008-04
Failure to Evaluate the Adverse Effects of Sharing the
RWSTs of Both Reactor Units (Section 1R21.5.b(1))05000454/2015008-05; 05000455/2015008-05
Failure to Adequately Implement a Design Change
Associated with the RWSTs (Section 1R21.5.b(2))05000454/2015008-06; 05000455/2015008-06
Failure to Evaluate the Adverse Effects of Changing the
SXCT Tornado Analysis as Described in the UFSAR
(Section 1R21.5.b(3))05000454/2015008-07; 05000455/2015008-07
Failure to Evaluate the Adverse Effects of Changing the
SXCT Tornado Analysis as Described in the UFSAR
(Section 1R21.5.b(3))05000454/2015008-08; 05000455/2015008-08
Failure to Provide Proper Direction for Low Level
Isolation of the RWST in EOPs (Section 1R21.6.b(1))05000454/2015008-09; 05000455/2015008-09
Failure to Promptly Correct an NRC-Identified NCV
Associated with the Capability to Detect and Isolate
ECCS Leakage (Section 4OA2.1.b(1))
2 05000454/2015008-10; 05000455/2015008-10
Failure to Maintain the Instrument Loops Used to Verify
Compliance with the Containment Average Air
Temperature TS Limit (Section 4OA2.1.b(2))05000454/2015008-11; 05000455/2015008-11
Operability Evaluation Relied on Probabilities of
Occurrence of the Associated Event
(Section 4OA2.1.b(3))
Closed 05000454/2015008-03; 05000455/2015008-03
Failure to Evaluate the Adverse Effects of Sharing the
RWSTs of Both Reactor Units (Section 1R21.5.b(1))05000454/2015008-04; 05000455/2015008-04
Failure to Evaluate the Adverse Effects of Sharing the
RWSTs of Both Reactor Units (Section 1R21.5.b(1))05000454/2015008-05; 05000455/2015008-05
Failure to Adequately Implement a Design Change
Associated with the RWSTs (Section 1R21.5.b(2))05000454/2015008-06; 05000455/2015008-06
Failure to Evaluate the Adverse Effects of Changing the
SXCT Tornado Analysis as Described in the UFSAR
(Section 1R21.5.b(3))05000454/2015008-07; 05000455/2015008-07
Failure to Evaluate the Adverse Effects of Changing the
SXCT Tornado Analysis as Described in the UFSAR
(Section 1R21.5.b(3))05000454/2015008-08; 05000455/2015008-08
Failure to Provide Proper Direction for Low Level
Isolation of the RWST in EOPs (Section 1R21.6.b(1))05000454/2015008-10; 05000455/2015008-10
Failure to Maintain the Instrument Loops Used to Verify
Compliance with the Containment Average Air
Temperature TS Limit (Section 4OA2.1.b(2))05000454/2015008-11; 05000455/2015008-11
Operability Evaluation Relied on Probabilities of
Occurrence of the Associated Event
(Section 4OA2.1.b(3))
3
LIST OF DOCUMENTS REVIEWED
The following is a list of documents reviewed during the inspection. Inclusion on this list does
not imply that the NRC inspectors reviewed the documents in their entirety, but rather, that
selected sections of portions of the documents were evaluated as part of the overall inspection
effort. Inclusion of a document on this list does not imply NRC acceptance of the document or
any part of it, unless this is stated in the body of the inspection report.
CALCULATIONS
Number
Description or Title
Revision
4391/19D-11
Sizing of Replacement Battery Charger for Diesel Driven
Pumps
0
BYR08-035
Essential Service Water Cooling Tower Basin Level
Indication Uncertainty Analysis
0
BYR12-070
Auxiliary Building Environment following a High Energy Line
Break in the Turbine Building
2
BYR12-072
Thermal Endurance Evaluation of the Safety Related
Electrical Equipment in the Essential Service Water (SX)
Cooling Tower Switchgear Rooms
0
BYR97-193
Battery Duty Cycle and Sizing for the Byron Diesel Driven
Auxiliary Feedwater Pumps and the Byron Diesel Driven
Essential Service Water Makeup Pumps
1-1E
BYR97-205
125VDC Battery Charger Sizing Calculation
2
BYR97-204
125 VDC Battery Sizing Calculation
3-3K
BYR97-224
125Vdc Voltage Drop Calculation
4-4A
BYR97-226
125 V DC System Short Circuit Calculation
4
BYR97-239
SX Cooling Tower Basin Level Auto Start Level Set Point
Analysis
1
BYR97-336
SX Cooling Tower Basin - Time to Reach the Low Level
Alarm Set Point
1
BYR2000-136
Voltage Drop Calculation for 4160V Switchgear Breaker
Control Circuits
1
BYR2000-191
Voltage Drop Calculation for 480V Switchgear Breaker
Control Circuits
0 -0C
4391/19-AN-3
Protective Relay Settings for 4.16 kV ESF Switchgear
16
19-AQ-24
Voltage Drop on 480-120V AC Control Transformer Circuits
8
19-AQ-63
Division Specific Degraded Voltage Analysis
7A
19-AQ-69
Evaluation of the Adequacy of the 120 Vac Distribution
Circuit at the Degraded Voltage Setpoint
16
19-AQ-75
Essential Service Water Cooling Tower 480V Buses
Maximum Voltage
1
19-AU-4
480 V Unit Substation Breaker and Relay Settings
19
19-G-1
Cable Ampacity
2
19-T-5
Diesel Generator Loading During LOOP/LOCA
7
BYR01-068
Environmental Parameters of EQ Zones
2
BYR01-084
Generic Thermal Overload Heater Sizing Calculation for
Motor Operated Valves
000
4
CALCULATIONS
Number
Description or Title
Revision
BYR01-095
Motor Operated Valves (MOV) Actuator Motor Terminal
Voltage and Thermal Overload Sizing Calculation - Essential
Service Water (SX) System
1
BYR06-111
Model APT-30K-11 SXCT Fan Blade Pitch Setting
1
BYR12-042
Essential Service Water Discharge Header Temperature
Indication Uncertainty
0
BYR95-005
120 VAC Instrument Bus/SSPS Cabinet Fuse Sizing and
Coordination
0
BYR96-128
Refueling Water Storage Tank (RWST) Level Alarm
Bistables and Level Indication Accuracy
2
DIT BB-EPED-
0189
Design Information Transmittal: Minimum Starting/Running
Voltages for Essential Motors
5/14/93
DIT BB-EXT-
0406
Design Information Transmittal: Essential Service Water
Cooling Tower Fan Motors [starting duty]
12/9/92
DIT-BRW-2002-
033
Design Information Transmittal: Basis for EDG loading
10/15/02
SI-90-01
Minimum Containment Flood Level
11
BYR04-016
RHR, SI, CV, and CS Pump NPSH During ECCS Injection
Mode
2
BYR14-053
Pressurizer PORV Air Accumulator Tank Requirements
0
BYR06-029
Byron/Braidwood SI/RHR/CS/CV system hydraulic analysis
in support of GSI-191
5
BYR06-058
NPSHA for RHR & CS Pumps During Post-LOCA
Recirculation
0
BYR07-055
Determination of the Correlation for the Critical Submergence
Height (Vortexing) for the RWST
0
SM-SI0930
RWST Level
D
SITH-1
Refueling Water Storage Tank (RWST) Level Set points
8
CN-RRA-00-47
Byron/Braidwood Natural Circulation Cooldown TREAT
Analysis for RSG and Uprating Programm
3
CN-RRA-00-47
Byron/Braidwood Natural Circulation Cooldown TREAT
Analysis for RSG and Uprating Program
4
CQD-200074
PORV Accumulator Tank
Z2
8.1.16
Refueling Water Storage Tanks Analysis and Design
5
BYR97-287
Determination of RWST Free Air Volume above Maximum
RWST Water Level
2
SM-SI0930
RWST Level
D
SM-SI0931
RWST Level
D
SM-SI0932
RWST Level
D
SM-SI0933
RWST Level
D
ATD-0062
Heat Load to the Ultimate Heat Sink During a Loss of
Coolant Accident
5
BYR03-131
Evaluation of UHS Make Up for CST-based Cooldown Profile 1
BYR05-018
Tornado Missile Risk Assessment of Vulnerable Targets of
Essential Service Water Cooling Towers
0
BYR06-111
Model APT-30K-11 SXCT Fan Blade Pitch Setting
1
5
CALCULATIONS
Number
Description or Title
Revision
BYR09-002
UHS Capability with Loss of SX Fans due to a Tornado
Event
1
BYR09-002
UHS Capability with Loss of SX Fans due to a Tornado
Event
1
BYR97*239
SX Cooling Tower Basin Level Auto Start Setpoint Error
Analysis
1
BYR97-034
Essential Service Water Cooling Tower Basin Minimum
Volume Versus Level and Minimum
Usable Volume Calculation
0a
BYR97-034
Essential Service Water Cooling Tower Basin Minimum
Volume Versus Level and Minimum
Usable Volume Calculation
0A
BYR97-127
Byron Ultimate Heat Sink Cooling Tower Performance
Calculations
1
BYR97-134
Heat Load on the UHS - 2 Unit Shutdown
3
BYR97-366
SX Cooling Tower Basin - Time to Reach the Low Level
Alarm Set Point
1
BYRO8-035
Essential Service Water Cooling Tower Basin Level
Indication Uncertainty Analysis
0
NED-M-MSD-
009
Byron Ultimate Heat Sink Cooling Tower Basin Temperature
Calculation: Part IV
8B
NED-M-MSD-
014
Byron Ultimate Heat Sink Cooling Tower Basin Makeup
Calculation
9
UHS-01
Ultimate Heat Sink Design Basis LOCA Single Failure
Scenarios
4
SL-101
ELMS-AC Report: Running Voltage Summary, Division 12
1/21/15
SL-102
ELMS-AC Report: Short Circuit Summary for High Voltage
Buses
1/21/15
SL-109
ELMS-AC Report: Connection Loading, Division 12
1/21/15
SL-112
ELMS-AC Report: Single Bus Summary, Bus 142
4/20/15
CORRECTIVE ACTION DOCUMENTS Generated Due to the Inspection
Number
Description or Title
Date
2015 CDBI - Design Analysis Inconsistency Identified
4/21/15
NRC CDBI: Loose Parts Found During Walkdown of RWST
4/22/15
CDBI - Bucket Collecting Diesel Fuel Drips from 0DO088A
4/22/15
CDBI - SX Make-Up Pump Temperature Recorder Panel
Memory Full
4/22/15
CDBI - Outdated Information in SystemIQ
4/22/15
NRC ID: Jumpers Not Readily Available for 1/2BOA PRI-5
4/22/15
Negative Vibration Reading on Idle 0E SXCT Fan
4/22/15
CDBI - ID 1RY456 WO As-Found Not as Expected, No IR
Written
4/24/15
CDBI - Issues Identified in Calculation BYR 97-224
4/30/15
CDBI - Issue Identified in Calculation 19-AQ-69
5/1/15
CDBI Question Related to BEP ES-1.3 Cold Leg
Recirculation
5/4/15
6
CORRECTIVE ACTION DOCUMENTS Generated Due to the Inspection
Number
Description or Title
Date
CDBI - FC Purification Flow Not Considered in RWST NPSH
Calc
5/4/15
CDBI - NRC Identified Issues in BYR97-193
5/5/15
CDBI - 50.59 and DRP did not explicitly evaluate GDC 5
5/5/15
NRC CDBI - Error Discovered in EACE Investigation
5/6/15
CDBI - RWST Calc May Lead to Inconsistent Application of
TS
5/6/15
NRC CDBI: Procedure Enhancement for ECCS Flow
Balancing
5/6/15
CDBI Deficiency Identified - THD Testing for Instrument
Inverter
5/8/15
Lightning Rod on SX Cooling Tower Bent; Clarify Inspection
WO Instructions
5/8/15
CDBI 2015 - VTIP for Containment DP Has Limited Lead
Length
5/15/15
CDBI - CA Created for NCV Does Not Resolve Issue
5/15/15
No Routine PM on Containment Temperature Loops
5/19/15
CDBI Concern Regarding Op Eval 13-007
5/22/15
CDBI - TS Clarification Needed for Transition to LTOPs
5/22/15
2012 50.59 for SXCT Tornado Analysis
5/19/15
CORRECTIVE ACTION DOCUMENTS Reviewed During the Inspection
Number
Description or Title
Date
Design of RWST Vacuum Relief System
2/15/05
RWST Vent / Vacuum Breaker Design Basis Issues
7/27/04
0A SX M/U PP Failures
2/13/09
0A SX MU Pump Did Not Stop When Local CS Taken to Off
2/17/09
1DC08E Battery, 1DC08E 123 Bus and DC 123 Batt Low
4/8/10
DC Bus 123 Low Voltage
9/21/10
Megger Test of Submerged Cable (1SX172)
4/20/11
Check/Adjust Charger 123 Float Voltage
5/17/11
0A SX MU PP Failed to Start at the Desired Setpoint SPC
9/15/11
0A SX M/U PP Battery Bank Test
1/25/12
Replace Breaker for MCC 035-2-C5 (0CW03PC-C)
5/4/12
CDBI ESF MCC Contactors not Tested at Assumed Pickup
Volt
5/18/12
CDBI Follow-up on MCC Contactors (IR 1368220)
6/11/12
NRC CDBI - Protective Relay Setting Tolerances
6/12/12
Need Engineering to Evaluate Test Frequency
6/15/12
Action Tracking Needed for Size 3 and 4 Contactors
6/22/12
The Station 111 ESF Battery Needs to Be Replaced in
B1R19
7/11/12
The Station 112 ESF Battery Needs to be Replaced in
B1R19
7/11/12
Protective Relay Tolerances Require Fleet Review
7/19/12
7
CORRECTIVE ACTION DOCUMENTS Reviewed During the Inspection
Number
Description or Title
Date
NRC IDD CDBI Green NCV Non-Conforming 480/120 VAC
Motor Contactors
6/15/12
NRC CDBI Green NCV Non-Conservative Cal Tolerance for
Elec Relays
6/15/12
Engineering Evaluate Frequency of Battery Capacity Test
9/16/12
0A SX Makeup Pump Failed to Auto Start per 0BOSR 7.9.6-
1
4/16/13
Breaker Will Not Reset During Oden Testing
5/29/13
0A SX M/U PP Had To Be Tripped During Monthly Run
10/10/13
Loss of Instrument Bus 111
11/21/13
Need New Contingency Work Order ofr Instrument Inverter
111
11/23/13
NRC ID - PCM Template/Vendor Manual Recommendation
11/26/13
0A SX Makeup Pump Auto Start Level Setpoint
1/23/14
Erratic Reading on Ammeter (111-IP001) for Inverter 111
4/30/14
Specific Gravity of Battery Cell Still Low After Equalize
5/10/14
0A SX MU PP Trouble Alarm Continues to Alarm
7/10/14
Gradual Float Current Trend on 111 Battery Charger
4/15/14
0SX02PA Kept Running
11/5/14
Pump As Found Condition/Dry Start Improvement
Opportunity
11/25/14
Thermography Needed on FRT for Instrument Inverter 111
11/29/14
0A SX MU Failed Surveillance
2/5/15
Replace Breaker for MCC 132Z1-A4 (0SX157A)
12/8/11
NER-NC-10-008-Y - Buried Cable
4/14/10
B2F26 Bus 142 Undervoltage Relay
2/3/12
Safety-Related Cable Vault 1M1G(1G1) Inspection - Repairs
9/5/12
MCC 132Z1-A5 Tripped Out of Tolerance
9/24/12
Safety-Related Cable Vault 1J2 Inspection - Repairs
10/12/12
Operating Experience Applicable to Byron (SXCT Fan
Reverse Rotation)
12/2/13
Degraded Voltage Relay Target did not Change State
2/25/14
Step Change Identified in Unit 1 Containment Air
Temperature in PI
4/16/14
Safety Related Cable Vault PM and Engineering Inspections
7/30/14
Cable Vault PM and Engineering Inspections
1/14/15
CDBI FASA - Review of Robinson and Wolf Creek Findings
1/15/15
RWST Vent/Vacuum Breaker Design Basis Issue
7/27/04
U-1 RWST level
4/30/12
U-1 RWST on FC Purification
5/2/12
U-1 RWST level loss During Purification
5/3/12
NRC Information Notice 2012-01: Seismic Considerations -
Principally Issues Involving Tanks
5/9/12
NRC CDBI Green NCV-Leak Detection for ECCS Flowpath
Lacking
6/15/12
CDBI, Question about ECCS leakage
6/15/12
8
CORRECTIVE ACTION DOCUMENTS Reviewed During the Inspection
Number
Description or Title
Date
Review of Braidwood IR 1459353 Pzr PORV Accumlator
Press
1/23/13
1B PZR PORV Accum Failed Decay Test
3/19/14
NOS ID: No CA to Correct an NRC NCV
2/18/15
IR298958
SSD&PC: Inaccurate Setpoints Referenced in BYR97-034
6/30/05
Inadequate 50.59 for EC 385829 (SXCT Tornado Missiles)
8/14/13
AR295141295141
Ssd&pc Question on Tornado Anaylsis Supporting UFSAR
Stmnt
1/28/05
Clarification Needed on UHS Passive Failure Design
7/1/14
NRC Question and Feedback on UHS Temperature Analysis 10/3/13
UFSAR Section 2.4.11.6 Needs Revision
7/1/14
Recommendation from UHS Assessment
7/1/14
Inadequate 50.59 for EC 385829
2/9/12
Failed Spider Bearing on 0A SX Makeup Pump
11/4/14
Obsolete SX Makeup Pump D/O Storage Tank Level
Indicator
9/28/11
Review of Flow Anomaly On 0B SX Makeup
1/14/15
0A SX MU Failed Surveillance
2/5/15
DRAWINGS
Number
Description or Title
Revision
S-529
Essential Service Cooling Tower Drainage Duct Plan,
Section Details
H
6E-0-4030SX09
Schematic Diagram - Essential Service Water Make-up
Pump 0A 0SX02PA
P
6E-0-4030SX23
Schematic Diagram - Essential Service Water Make-up
Pump 0A Control Cabinet (Diesel Driven) 0SX02JA
S
6E-0-4030SX24
Schematic Diagram - Essential Service Water Make-up
Pump 0A Control Cabinet (Diesel Driven) 0SX02JA
F
6E-0-
4030CW11
Schematic Diagram - Essential service Water Cooling Tower
0A & 0B Well Water Make-up Valves 0CW100A & B
D
6E-0-
4030WW01
Schematic Diagram - Deep Well Pump 0A - 0WW01PA
M
6E-0-
4030WW02
Schematic Diagram - Deep Well Pump 0B - 0WW01PB
H
6E-0-
4030WW05
Schematic Diagram - Essential service Water Cooling Tower
0A & 0B Circulating Water Make-up Valves 0WW019A & B
E
Station One Line Diagram
P
Station Key Diagram
O
Single Line Diagram - 120V AC ESF Instrument Inverter Bus
111 and 113, 125V DC ESF Distribution Center 111
K
Byron - Unit 1 - Key Diagram 480V ESF Substation Bus
131X (1AP10E)
M
Key Diagram - 125V DC ESF Distribution Center Bus 111
(1DC05E) Part 1
M
9
DRAWINGS
Number
Description or Title
Revision
Key Diagram - 125V DC ESF Distribution Center Bus 111
(1DC05E) Part 2
G
Key Diagram - 125V DC Non Safety Related Distribution
Panel 113 (1DC05EB)
K
6E-1-4030DC05
Schematic Diagram - 125 VDC ESF Distribution Center, Bus
111, Part 1, 1DC05E
U
6E-1-4030IP01
Schematic Diagram 7.5KVA Fixed Frequency Inverter for
Instrument Bus 111 (1IP05E)
0
6E-1-4030RC31
Schematic Diagram - Reactor Coolant System High Pressure
& Low Temperature Control & Alarms
G
6E-1-4030RH02
Schematic Diagram - Residual Heat Removal Pump 1B -
1RH01PB
N
6E-1-4030RY14
Schematic Diagram - Pressurizer Pressure & Level Control
Safety Related & Non-Safety Related (Div 12)
F
6E-1-4030RY17
Schematic Diagram - Pressurizer Power Relief Valves -
1RY455A & 1RY456; Pressurizer Relief Tank Primary Water
Supply Isolation Valve - 1RY8030; Pressurizer Relief Tank
Drain Isolation Valve 1RY8031
V
6E-1-4031RC26
Loop Schematic Diagram - Reactor Coolant System Cold
Overpressurization System Control 1A & 1D Control Cabinet
5 & 6
S
6E-1-4031RY15
Loop Schematic Diagram - Pressurizer Pressure & Level
Control Cabinet 6 (1PA06J) Part 1
O
6E-1-4031RY19
Loop Schematic Diagram - Pressurizer Pressure Safety
Valve Discharge Temp & Pressure Control (ITE-0464)
Control Cabinet 7 (1PA07J)
F
M-42 Sh. 6
Diagram of Essential Service Water
M-60 Sh. 5
Diagram of Reactor Coolant
M-2042 Sh. 5
P&ID/C&I Diagram ESS Service Water System - SX
F
Duct Runs, Outdoor Plan, Southeast Area
Duct Runs, Outdoor Plan, Southwest Area
Y
Duct Runs, Sections
F
Electrical Installation, ESW Cooling Tower 0A Plan -
Switchgear Room, Elev. 874-6
AZ
6E-0-3502CT1
Conduit Tabulation, ESW Cooling Tower 0A Plan -
Switchgear Room, Elev. 874-6
T
6E-0-3502D01
Electrical Installation, ESW Cooling Tower 0A Switchgear
Room Partial Plans and Sections
N
Electrical Installation, ESW Cooling Tower 0B Plan -
Switchgear Room, Elev. 874-6
BN
6E-0-3507CT1
Conduit Tabulation, ESW Cooling Tower 0B Plan -
Switchgear Room, Elev. 874-6
Y
6E-0-3507D01
Electrical Installation, ESW Cooling Tower 0B Switchgear
Room Partial Plans and Sections
W
6E-0-4030SX01
Schematic Diagram, Essential Service Water Cooling Tower
0A, Fan 0A
V
Duct Run Routing Outdoor - West of Station
10
DRAWINGS
Number
Description or Title
Revision
6E-0-4030SX02
Schematic Diagram, Essential Service Water Cooling Tower
0A, Fan 0B
U
6E-0-4030SX03
Schematic Diagram, Essential Service Water Cooling Tower
0A, Fan 0C
U
6E-0-4030SX04
Schematic Diagram, Essential Service Water Cooling Tower
0A, Fan 0D
W
6E-0-4030SX05
Schematic Diagram, Essential Service Water Cooling Tower
0B, Fan 0E
V
6E-0-4030SX06
Schematic Diagram, Essential Service Water Cooling Tower
0B, Fan 0F
W
6E-0-4030SX07
Schematic Diagram, Essential Service Water Cooling Tower
0B, Fan 0G
W
6E-0-4030SX08
Schematic Diagram, Essential Service Water Cooling Tower
0B, Fan 0H
W
Station One Line Diagram
P
Key Diagram, 4160V ESF Switchgear Bus 142
J
Key Diagram, 480V ESW Cooling Tower ESF MCC 132Z1
R
Key Diagram, 120 Vac Instrument Bus 111
W
Relaying & Metering Diagram, 4160 ESF Switchgear Bus
142
U
6E-1-
4030AP115
Schematic Diagram, Tripping Circuit, 480V ESW Cooling
Tower MCC 131Z1A, 132Z1A
A
6E-1-4030RY17
Schematic Diagram, Pressurizer Power Relief Valve 1RV456
V
6E-1-4030SI02
Schematic Diagram, Safety Injection Pump 1B
N
6E-1-4030SI14
Schematic Diagram, Containment Sumps 1A and 1B
Isolation Valves SI8811A & B
Q
6E-1-4031VP11
Loop Schematic Diagram [containment inside/outside
differential pressure]
K
M-61, Sh. 1B
Diagram of Safety Injection
AX
M-136, Sh. 1
Diagram of Safety Injection
BB
M-63, Sh. 1A
Diagram of Fuel Pool Cooling and Clean up
BI
S-1404
Refueling Water Storage Tank Sections & Details
I
M-60, Sh. 8
Diagram of Reactor Coolant
Sh. 1
Pressurizer PORV Air Relief Valve
0
M-60, Sh.5
Diagram of Reactor Coolant
10 CFR 50.59 DOCUMENTS (Screenings/Safety Evaluations)
Number
Description or Title
Date
DCP 9600355 ESW Cooling Tower Basin Level Switch
7/3/97
Tornado Missile Design Basis for the Essential Service
Water Cooling Tower
0
Tornado Missile Design Basis for the Essential Service
Water Cooling Towers
2/9/12
Multiple Spurious Operation - Scenario 14, 1SI8811A/B
12/9/11
UFSAR Change Package (DRP)11-052
11/16/05
11
10 CFR 50.59 DOCUMENTS (Screenings/Safety Evaluations)
Number
Description or Title
Date
Increase Pressurizer PORV tank Operating Pressure to
Increase Margin for PORV Operation (Unit 1)
0
Technical Requirements Manual (TRM) Revision to Delete
TLCO 3.4.a, Pressurizer Safety Valves-Shutdown
9/19/00
MISCELLANEOUS
Number
Description or Title
Date or
Revision
IST Program Plan - Service Water System
8/26/14
Standing Order 15-020
Emergency Operating Procedure Cold Leg Recirc.
5/15/15
DW-09-004
ERG Feedback
2/27/09
Stewart & Stevenson Certificate of Conformance for Battery
Chargers Serial No. 2165, 2167, 2170, 2174, 4 Batterrie 20
Cells/Set and 8 Battery Racks, Purchase Order No. 203731
11/4/81
06EN003246
FLT Series Flex Switch - Flow, Level, Temperature Switch
Monitor
2
01492090-03
Level 3 OPEX Evaluation - NRC IN 2013-05: Battery
Expected Life and Its Potential Impact on Surveillance
Requirements
5/16/13
CQD-009436
Seismic Qualification Test Report for Nife Ni-Cad Batteries
H-410 (1,2 AF01EA-A, EA-B, EB-A, EB-B/0SX02EA, EB-A,
EC-A, ED-A
8/17/83
CQD-012527
Review of Seismic Qualification Test Report for Battery
Chargers (1&2 DC03E, 04E)
10/2/13
CQD-049161
Justification for the Application of Permatex Form A Gasket
with EPT Diaphragms
1
CQD-200164
Dynamic Qualification of Battery Chargers 0SX02EA-1
through 0SX02ED-1; 1,2AF01EA-1 and 1,2AF01EB-1
5/29/86
NEC-06-6066
Procurement of Safety Related 125 Volt Batteries
B
604990-70-F1
Reliance Electric Dimension Sheet [SX Cooling tower fan
motor data sheet]
4/4/78
EQ-GEN023
EQ Binder for NAMCO EA180 limit switches
13
EQ Evaluation - Pressurizer PORV Diaphragm Design
Pressure
0
EQER-06-98-
002
EQ Evaluation for PORVs 1(2) FSV-RY-455A & 1(2)FSV-RY-
456
2/29/99
Low Temperature Protection (LTOP) System Evaluation for
Byron and Braidwood Units 1 and 2 Measurement
Uncertainty Recapture (MUR) Power Uprate Program
9/7/10
Simulator Work
Request 13961
PZR PORV Testing reveals lower than design flow
4/25/12
Byron Unit 1 Pressure and Temperature Limits Report
3/14
Summary of the Design and Licensing Basis for Inadvertent
ECCS Actuation at Power
0
12
MODIFICATIONS
Number
Description or Title
Date or
Revision
Ultimate Heat Sink Capability with Loss of Essential Service
Water Cooling Tower Fans
2
UHS Capability with Loss of SX Fans Due to Tornado
Missiles
2/14/12
M6-1(2)-87-142
Install Fan Cooling to Instrument Power Inverter Cubicles
10/17/90
Multiple Spurious Operation - Scenario 14, 1SI8811A/B
12/9/11
Detailed Review of FC Purification System for Use of Non
Safety Related Portion Connected to Safety Related Piping
0
Increase U1 Pressurizer PORV Accumulator Tank Operating
Pressure to Increase number of PORV Open/Close Cycles
from Accumulator
0
OPERABILITY EVALUATIONS
Number
Description or Title
Date 13-001
Capacity of the Pressurizer PORV Air Accumulator During
Natural Circulation Cooldown
5 13-007
Ultimate Heat Sink Capability with Loss of Essential Service
Water Cooling Tower Fans
1
PROCEDURES
Number
Description or Title
Revision
1BOA PRI-5
Control Room Inaccessibility
108
1BOA ELEC-5
Local Emergency Control of Safe Shutdown Equipment
106
0BOA PRI-7
Loss of Ultimate Heat Sink Unit 0
1
1BOA PRI-7
Essential Service Water Malfunction Unit 1
106
1BEP ES-1.3
Transfer to Cold Leg Recirculation Unit 1
204
1BCA-1.2
LOCA Outside Containment Unit 1
200
Operator Response Time Program
3
OP-BY-102-106
Operator Response Time Program at Byron Station
7
1BOA S/D-2
Shutdown LOCA Unit 1
105
1BOSR XRS-Q1 Unit One Remote Shutdown Panel Quarterly Surveillance
13
Response to Loss of Secondary Heat Sink Unit1
203
0BHSR 8.4.2-1
Unit Zero Comprehensive Inservice Testing (IST)
Requirements for Essential Service Water Makeup Pump 0A
8
0BHSR SX-1
Unit 0 0A SX Makeup Pump Battery Bank A Capacity Test
0
0BHSR SX-5
0A SX Makeup Pump Battery Bank D Capacity Test
0
0BISR 7.a.4-200 Calibration of Essential Service Water Cooling Tower Basin
0A Level Switch (SX)
7
0BOSR Z.7.a.2-
1
Unit Common Deepwell Pump Operability Monthly
Surveillance
1
0BOSR 7.9.6-1
Essential Service Water Makeup Pump 0A Monthly
Operability Surveillance
32
0BVSR SX-1
Unit 0 0A SX Makeup Pump Battery Bank A Capacity Test
3
0BVSR SX-4
Unit 0 0A SX Makeup Pump Battery Bank D Capacity Test
3
0BVSR WW-1
Biennial Deep Well Pump Structure Inspection
2
1BHSR 8.4.2-1
Unit 1 Bus 111 125V Battery Charger Operability
1
13
PROCEDURES
Number
Description or Title
Revision
1BHSR 8.4.3-1
Unit 1 125 Volt Battery Bank 111 Service Test
3
1BHSR 8.6.6-1
Unit 1 Battery 111 125 Volt Battery Bank 5 Year Modified
Performance Test
0 & 2
1BHSR AF-1AA
Unit 1 1B Diesel Aux Feed Pump Battery Bank A Battery A
(1AF01EA-A) Capacity Test
1
1BOA ELEC-1
Loss of DC Bus Unit 1
103
1BOSR 8.4-1
125V DC Bus 111 Load Shed When Cross-Tied to DC Bus
211
12
2BHSR 8.4.2-1
Unit 2 Bus 211 125V Battery Charger Operability
1
BISR 3.1.10-206 Pressurizer Pressure Protection Channel II (RY) Test Report
Package)
8
BISR 3.1.10-207 Pressurizer Pressure Protection Channel III (RY) Test Report
Package)
8
BISR 4.12.8-200 Wide Range Reactor Coolant Pressure Loop 1A Hot Leg
(RC)
7
BOP-AP-93
MCC 035-2 Outage
1
BOP SX-3
Essential Service Water Make-up Pump Startup
30
BOP SX17
Shutdown of SX Makeup Pump Battery Chargers
3
BOP SX18
Placing the SX Makeup Pump Battery Chargers in
Operation/Equalize
8
Control and Tracking of Electrical Load Changes
4
ER-AA-310-
1004
Maintenance Rule - Performance Monitoring
13
MA-BY-026-
1001
Seismic Housekeeping
2
MA-BY-721-060
125 Volt Battery Bank 18 Month Surveillance
11
MA-BY-721-061
125 Volt Battery Bank Quarterly Surveillance
12 & 15
MA-BY-723-053
Station Battery Charger 18 Month Surveillance
18
MA-BY-723-
053-001
0B SX Makeup Pump A Battery Charger 0SX02EA Battery
Charger Test
0
MA-BY-723-
053-002
0B SX Makeup Pump D Battery Charger 0SX02ED Battery
Charger Test
1
MA-BY-723-
053-003
0B SX Makeup Pump B Battery Charger 0SX02EB Battery
Charger Test
0
MA-BY-723-
053-004
0B SX Makeup Pump C Battery Charger 0SX02EC Battery
Charger Test
1
MA-BY-723-054
Nickel Cadmium Battery Bank Surveillance
14
0BHSR SX-3
Annual Surveillance for Essential Service Water Cooling
Tower Fan Motors
2
0BOSR 7.9.4-1
ESW Cooling Tower Fan Monthly Surveillance
6
1BOSR IP-R1
Instructions to Cycle Instrument Bus 111 Distribution Panel
0
1BOSR 3.2.9-1
Train A Manual Safety Injection Initiation and Manual Phase
A Initiation Surveillance
22
1BOSR 8.9.1-2
Unit 1 ESF Onsite Power Distribution Weekly Surveillance
Division 12
10
Adjusting Reactive Load
12
14
PROCEDURES
Number
Description or Title
Revision
Cable Condition Monitoring Program
1
Electrical Testing of AC Motors Using Baker Instrument
Advanced Winding Analyzer
3
Preventative Maintenance on Westinghouse Type DHP 4kv,
6.9kv, and 13.8kv Circuit Breakers
8
Plant Shutdown and Cooldown
68
Startup of the Purification System to Purify or Recirculate the
Refueling Water Storage Tank
13
1BEP ES-0.2
Natural Circulation Cooldown Unit 1
202
BAR 1-12-C4
RCS Press High at Low Temp
2
1BOSR 5.C.3.1
Safety Injection System Cold Leg Flow Balance
3
2BOSR 0.1-4
Unit 2 Mode 4 Shiftly and Daily Operating Surveillance
25
1BOSR 0.1-
1,2,3
Unit 1 Mode 1,2,3 TRM and Tech Spec and Non Tech Spec
Data Sheet D5
56
BIP 2500-088
Calibration of Refueling Water Storage Tank Outlet
Temperature Loop (SI)
5
1BOSR
5.5.8.SI.5-2C
Unit 1 Comprehensive Inservice Testing (IST) Requirements
for Safety Injection Pump 1SI01PB
5
1BOSR
5.5.8.SI.5-2a
Unit 1 Group A Inservice Testing (IST) Requirements for
Safty Injection Pumps 1SI01PB
1
0BOSR NLO-
Non-Licensed Operator TRM, ISFSI, and NPDES Data Daily
Logs
18
1BGP 100-5
Plant Shutdown and Cooldown
68
BOP SX-T2
SX Basin Level Tree
5
BOP SX-11
SXCT Fan Startup
9
BOP SX-12
Makeup to an Essential Service Water Mechanical Draft
10
0BOA ENV-1
Adverse Weather Conditions
114
1BOA PRI-5
Control Room Inaccessibility
108
1BOA ELEC-5
Local Emergency Control of Safe Shutdown Equipment Unit
1
106
Reactor Trip or Safety Injection
207
1BEP ES-0.1
Reactor Trip Response
203
1BEP ES-0.2
Natural Circulation Cooldown
202
BOP RH-6
Operation of the RH System In Shutdown Cooling
46
Oversight and and Control of Operator Burdens
2
BOP CC-1
Component Cooling Water System Startup
12
SURVEILLANCES (Completed)
Number
Description or Title
Date or
Revision
0BHSR SX-1
0A SX Makeup Pump Battery Bank A Capacity Test
6/14/12
0BHSR SX-5
0A SX Makeup Pump Battery Bank D Capacity Test
9/14/12
0BISR 7.a.4-200 Calibration of Essential Service Water Cooling Tower Basin
0A Level Switch (SX)
8/7/14
0BOSR
5.5.8.SX.5-1c
0SX02PA Comprehensive IST Req for SX Makeup Pump
2/5/15
15
SURVEILLANCES (Completed)
Number
Description or Title
Date or
Revision
0BOSR 7.9.6-1
0A SX Makeup Pump Operability Surveillance
3/12/13
0BOSR 7.9.6-1
0A SX Makeup Pump Operability Surveillance
2/4/15
0BOSR 7.9.6-1
0A SX Makeup Pump Battery Bank A Capacity Test
3/11/15
0BVSR SX-1
0A SX Makeup Pump Battery Bank A Capacity Test
10/17/06
0BVSR SX-4
0A SX Makeup Pump Battery Bank D Capacity Test
6/19/06
1BHSR 8.4.2-1
Unit 1 Bus 111 125V Battery Charger Operability Test
11/8/11
1BHSR 8.4.2-1
Unit 1 Bus 111 125V Battery Charger Operability Test
9/17/13
1BHSR 8.4.3-1
111 A Train 125V Battery Bank Service Test
3/20/14
1BHSR 8.6.6-1
111 A Train 125V Battery Bank 5Yr Capacity Test
4/1/08
1BHSR 8.6.6-1
111 A Train 125V Battery Bank 5Yr Capacity Test
9/11/12
BISR 3.1.10-206 Pressurize Pressure Protection Channel 2 Loop 1RY-0456
4/6/15
BISR 3.1.10-207 Pressurizer Pressure Protection Channel 3 Loop 1RY-0457
4/13/15
BISR 4.12.8-200 Cal of Wide Range RC Pressure Loop 1A Hot Leg 1P-406
4/28/14
M A-BY-721-
060
125 Volt Battery Bank Quarterly Surveillance
9/11/12
M A-BY-721-
060
125 Volt Battery Bank Quarterly Surveillance
3/20/14
M A-BY-721-
061
125 Volt Battery Bank 18 Months Surveillance
9/16/12
M A-BY-721-
061
125 Volt Battery Bank 18 Months Surveillance
3/22/14
M A-BY-721-
061
125 Volt Battery Bank 18 Months Surveillance
9/15/14
M A-BY-721-
061
125 Volt Battery Bank 18 Months Surveillance
12/16/14
MA-BY-723-053
EM 18 Month Battery Charger Surveillance - 0B SX M/U
Pump 0B Batt Chgr # 0SX02EB-1
1/15/13
MA-BY-723-053
EM 18 Month Battery Charger Surveillance - 0A SX M/U
Pump 0A Batt Chgr # 0SX02EA-1
2/6/14
MA-BY-723-053
EM 18 Month Battery Charger Surveillance - 0A SX M/U
Pump 0D Batt Chgr # 0SX02ED-1
8/5/14
MA-BY-723-053
EM 18 Month Battery Charger Surveillance - 0B SX M/U
Pump 0C Batt Chgr # 0SX02EC-1
3/27/15
MA-BY-723-
053-001
0B SX Makeup Pump A Battery Charger 0SX02EA Battery
Charger Test
2/4/14
MA-BY-723-
053-002
0B SX Makeup Pump D Battery Charger 0SX02ED Battery
Charger Test
8/6/14
MA-BY-723-
053-003
0B SX Makeup Pump B Battery Charger 0SX02EB Battery
Charger Test
1/15/13
MA-BY-723-
053-004
0B SX Makeup Pump B Battery Charger 0SX02EC Battery
Charger Test
3/27/15
MA-BY-723-054
Quarterly 24 VDC NiCad Battery Surveillance M/U Diesel
SX- 0SX02ED-A
8/5/14
MA-BY-723-054
Quarterly 24 VDC NiCad Battery Surveillance M/U Diesel
SX- 0SX02EA-A
9/5/14
16
SURVEILLANCES (Completed)
Number
Description or Title
Date or
Revision
MA-BY-723-054
Quarterly 24 VDC NiCad Battery Surveillance M/U Diesel
SX- 0SX02EA-A
10/30/14
MA-BY-723-054
NiCad Battery Surveillance M/U Diesel SX- 0SX02E
11/6/14
Unit 1 Pressurizer PORV Accumulator Press Decay Test
3/19/14
SI pump ECCS Flow Balance Test (After System Alteration)
10/5/14
OP 2BOSR 5.C.3-2 Unit 2 SI to HL Flow Balance
4/2/10
Unit 1 Safety Injection System Hot Leg Flow Balance
9/4/09
SI pump ECCS Flow Balance Test (After System Alterations)
9/4/09
1SI01PB Comprehensive IST RQMTS For Safety Injection
Pump
1/28/14
1SI01PB Comprehensive IST RQMTS For Safety Injection
Pump
8/9/12
STT/PIT For 1RY455A and 1RY456
9/28/12
STT/PIT For 1RY455A and 1RY456
2/7/14
PMID 140860
0BOSR 7.9.6-1 0A SX Makeup Pump Operability Review
4/18/13
TRAINING DOCUMENTS
Number
Description or Title
Date or
Revision
BY 14-2-2
Requalification Simulator Scenario Guide
1
10-1-5
Requalification Simulator Scenario Guide
0
P1-SPBY-1401
BEP-1, BEP-2
2
OPBYLLORT5
BFR H, Heat Sink Series
8/28/13
WORK DOCUMENTS
Number
Description or Title
Date or
Revision
00961518
Replace Entire Solenoid to Meet EQ Requirements - EM
ASCO Solenoid Valve Replacement (EQ) - 1FSV-RY456-2
4/1/08
01057719
Test All MCC Breakers in This MCC in a Bus Outage -
Assembly 480V RSH MCC 035-2
5/2813
01094421
Replace Float and Equalize Voltage Adjustment
11/29/11
01490541
111 A Train 125 V Battery Charger Operability Test
9/18/13
01536066
Essential Service Water Cooling Tower Level 0SX-064 IM
Calibration
3/3/14
01558514
B1R19 Replace 111 ESF Batteries
3/29/14
01578627
Test Replace Actuator Hose 1RY456
3/14/14
01599481
Calibration of Wide Range RC Pressure Loop 1A Hot Leg
Pressure Loop 1RC-0406
4/28/14
01600072
Clean/Inspect/Check Connections on DC Bus/Panel 111 and
Perform Therm. on Distr. Panel Breakers
3/30/14
01621944
Support Diver Insp./Cleaning RSH South 0B Intake/SED PM
ID 30
6/25/13
01652815
211 A Train 125 V Battery Charger Operability Test
5/14/14
17
WORK DOCUMENTS
Number
Description or Title
Date or
Revision
01017127
Perform Dynamic Baker Testing - 1SI01PB Motor
8/26/08
01085998
Perform Static Baker Test and MA-AA-723-310 Inspection of
SX Cooling Tower Fan Motor 0SX03CC
4/27/09
01117942
PM for 4kV Bus 142, breaker ACB 1425Z
9/21/09
01119375
Lightning Protection System 5 Year Inspection [Includes
Document 1 attachment to WO]
11/18/09
01120491
PM for 4kV Bus 142, breaker ACB 1424
9/29/09
01129028
Inspection of SX Cooling Tower Fan Motor 0SX03D
10/28/09
01136617
PM for 4kV Bus 142, breaker ACB 1422
3/15/09
01141049
Perform Static Baker Test and MA-AA-723-310 Inspection of
SX Cooling Tower Fan Motor 0SX03CB
3/19/10
01216011
Perform Dynamic Baker Testing - 1SI01PB Motor
8/26/10
01258194
Calibration of OLS-XS097
1/6/11
01265167
PM for 4kV Bus 142, breaker ACB 1421
10/26/11
01287321
Inspection of SX Cooling Tower Fan Motor 0SX03CE
9/1/11
01299949
Containment Inside/Outside DP Loop 1VP-231
6/30/11
01343409
Inspection of SX Cooling Tower Fan Motor 0SX03CH
11/21/11
01367641
PM for 4kV Bus 142, breaker ACB 1SI01PB
2/21/12
01372340
PM for 4kV Bus 142, breaker ACB 1422
11/11/12
01382271
Perform Static Baker Test and MA-AA-723-310 Inspection of
SX Cooling Tower Fan Motor 0SX03CC
6/12/12
01384474-01
Inspection of SX Cooling Tower Fan Motor 0SX03CF
11/26/12
01380551-01
Inspection of SX Cooling Tower Fan Motor 0SX03CA
6/8/12
01393782
Inspection of SX Cooling Tower Fan Motor 0SX03CG
10/30/11
01401180
Calibration of OLS-XS097
8/24/12
01419437
PM for 4kV Bus 142, breaker ACB 1425Z
9/23/12
01419758
Test All MCC 132Z1 Breakers - Oden Testing
9/23/12
01420365
PM for 4kV Bus 142, breaker ACB 1424
9/23/12
01421751
Unit 1 Train A Manual SI and Manual Phase A Initiation
Surveillance
9/11/12
01433378-01
Inspection of SX Cooling Tower Fan Motor 0SX03CD
3/12/13
01453350
Containment Inside/Outside DP Loop 1VP-231
3/19/15
01471461
Calibration of OLS-XS096
9/6/11
01473594-01
Perform Static Baker Test and MA-AA-723-310 Inspection of
SX Cooling Tower Fan Motor 0SX03CB
5/17/13
01480666-01
Testing of Power Cables 2AP178
4/20/13
01486337
Calibration of OLS-XS096
2/8/13
01538412
PM for 4kV Bus 142, breaker ACB 1423
11/30/13
01564018-01
Testing of Power Cables 1AP178 (North SX towers)
3/18/14
01569220
Calibration of OLS-XS097
6/2/14
01585654-02
Testing of Power Cables 2AP183 (Bus 242, Cubicle 20)
10/6/14
01615167
Calibration of OLS-XS096
8/8/14
01621573-01
Perform Surveillance of SX Cooling Tower Fan Motor
0SX03CE
9/16/14
01639602
PM for 4kV Bus 142, breaker ACB 1421
11/19/14
18
WORK DOCUMENTS
Number
Description or Title
Date or
Revision
01644724-01
Perform Surveillance of SX Cooling Tower Fan Motor
0SX03CH
11/20/14
01652671
PM for 4kV Bus 142, breaker ACB 1SI01PB
3/29/15
01667453
Calibration of 1SX-015 Loop
2/17/15
01680518
Calibration of 1SX-016 Loop
3/31/15
01543156
Calibration of 2SX-015 Loop
2/12/14
01716477
Calibration of 2SX-016 Loop
3/23/15
01734645-01
SX Cooling Tower Fan Motor Surveillance - 0SX03CG
11/4/14
01734645-02
SX Cooling Tower Fan Motor Surveillance & Triannual
Inspection - 0SX03CG
11/5/14
01760801
PM for 4kV Bus 142, breaker ACB 1423
1/30/15
01805922
ESW Cooling Tower Fan Monthly Surveillance
3/10/15
01419750
Replace Actuator Diaphragm
9/20/12
01515448
Refueling Water Storage Tank Outlet Temp LOOP 1SI-058
2/24/14
01186461
Refueling Water Storage Tank Outlet Temp LOOP 1SI-058
4/21/10
01544629
Calibration of Refueling Water Storage Tank (RWST) level
9/20/13
01374939
Calibration of Refueling Water Storage Tank (RWST) level
2/28/12
00915331
Minor Leakage from 0A WW Pump Well Head
8/20/08
00768385
0B WW PP 10 Year Rebuild
11/09/06
01754077
Received 0A SX Make Up Pp Trouble alarm
7/17/14
00921203
SXCT Fan Assembly Replacement EC 356417
8/23/12
00921198
SXCT Fan Assembly Replacement EC 356417
1/10/07
01634644
Replace Start Contactor Relay K1B at 0SX02PA-B
4/17/13
01682260
Support Diver Insp/Cleaning SXCT South 0B Basin
10/31/14
01691008
Support Diver Insp/Cleaning SXCT South 0A Basin
11/14/14
19
LIST OF ACRONYMS USED
Delta Core Damage Frequency
Alternating Current
Action Tracking Item
Agencywide Document Access Management System
CA
Corrective Action Tracking Item
Corrective Action Program
Conditional Core Damage Probability
Component Design Bases Inspection
CFR
Code of Federal Regulations
Containment
CV
Chemical and Volume Control
Design Basis Accident
Direct Current
Division of Reactor Projects
Division of Reactor Safety
EC
Engineering Change
Emergency Operating Procedure
Emergency Response Guideline
Final Safety Analysis Report
gpm
Gallons per Minute
IMC
Inspection Manual Chapter
IN
Information Notice
IR
Inspection Report
LCO
Limiting Condition for Operation
Limited Liability Corporation
Loss of Coolant Accident
Low Temperature Overpressure Protection
Motor Control Center
Motor-Operated Valve
Non-Cited Violation
NEI
Nuclear Energy Institute
Net Positive Suction Head
NRC
U.S. Nuclear Regulatory Commission
Nuclear Reactor Regulation
Publicly Available Records System
Preventive Maintenance
Power-Operated Relief Valve
Risk Assessment Standardization Project
Regulatory Issue Summary
Refueling Water Storage Tank
20
Systems Analysis Programs for Hands-on Integrated Reliability Evaluations
Significance Determination Process
Safety Injection
Standardized Plant Analysis Risk
SR
Surveillance Requirement
Senior Reactor Analyst
System, Structure, and Component
SSDPC
Safety Systems Design, Performance and Capability Inspection
Emergency Service Water
SXCT
Emergency Service Water Cooling Tower
Tornado Missile Risk Evaluation Methodology
TS
Technical Specification
Updated Final Safety Analysis Report
Unresolved Item
VAC
Volts Alternating Current
VDC
Volts Direct Current
Westinghouse Owners Group
B. Hanson
-3-
In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public
Inspections, Exemptions, Requests for Withholding, of the NRC's "Rules of Practice," a copy
of this letter, its enclosure, and your response (if any) will be available electronically for public
inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS)
component of the NRC's Agencywide Documents Access and Management System (ADAMS).
ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html
(the Public Electronic Reading Room).
Sincerely,
/RA/
Christine A. Lipa, Chief
Engineering Branch 2
Division of Reactor Safety
Docket Nos. 50-454; 50-455
Enclosures:
(2) IR 05000454/2015008; 05000455/2015008;
cc w/encl: Distribution via LISTSERV
DISTRIBUTION w/encl:
Kimyata MorganButler
RidsNrrDorlLpl3-2 Resource
RidsNrrPMByron Resource
RidsNrrDirsIrib Resource
Cynthia Pederson
DRPIII
DRSIII
ROPreports.Resource@nrc.gov
ADAMS Accession Number ML15203A042
Publicly Available
Non-Publicly Available
Sensitive
Non-Sensitive
To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copy
OFFICE
RIII
RIII
RIII
RIII
NAME
MJones for NFeliz-Adorno:cl
CLipa:
DATE
07/21/15
07/21/15
OFFICIAL RECORD COPY