IR 05000443/2013008
| ML13157A343 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 06/06/2013 |
| From: | Krohn P G Engineering Region 1 Branch 2 |
| To: | O'Keefe M, Walsh K NextEra Energy Seabrook |
| References | |
| IR-13-008 | |
| Download: ML13157A343 (48) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I 2100 RENAISSANCE BOULEVARD, SUITE 100 KING OF PRUSSIA, PENNSYLVANIA 19406-2713 June 6, 2013 Mr. Kevin Walsh Site Vice President, North Region
Seabrook Nuclear Power Plant NextEra Energy Seabrook, LLC
c/o Mr. Michael O'Keefe P.O. Box 300 Seabrook, NH 03874
SUBJECT: SEABROOK STATION, UNIT 1- NRC COMPONENT DESIGN BASES INSPECTION REPORT 05000443/2013008
Dear Mr. Walsh:
On April 26, 2013, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Seabrook Station. The enclosed inspection report documents the inspection results, which were discussed on April 26, 2013, with Mr. Kevin Walsh, Site Vice President, and other members of your staff.
The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license.
In conducting the inspection, the team examined the adequacy of selected components and operator actions to mitigate postulated transients, initiating events, and design basis accidents. The inspection involved field walkdowns, examination of selected procedures, calculations and records, and interviews with station personnel.
This report documents four NRC-identified findings which were of very low safety significance (Green). Three of the findings were determined to involve violations of NRC. However, because of the very low safety significance of the violations and because they were entered into your correction action program, the NRC is treating these violations as non-cited violations (NCV) consistent with Section 2.3.2.a of the NRC Enforcement Policy. If you contest any of the NCVs in this report, 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, D.C. 20555-0001, with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the Seabrook 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 I; and the NRC Resident Inspector at Seabrook Station. In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for the public inspection in the NRC Public Docket Room or from the Publicly Available Records component of NRC'
s document 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/
Paul G. Krohn, Chief Engineering Branch 2 Division of Reactor Safety
Docket No. 50-443 License No. NPF-86
Enclosure:
Inspection Report 05000443/2013008 w/Attachment: Supplemental Information
cc w/encl: Distribution via ListServ
ML13157A343 SUNSI Review Non-Sensitive Sensitive Publicly Available Non-Publicly Available OFFICE RI/DRS RI/DRS RI/DRP RI/DRS NAME KMangan CCahill GDentel PKrohn DATE 6/3/13 6/3/13 6/6/13 6/6/13 i Enclosure U. S. NUCLEAR REGULATORY COMMISSION REGION I Docket No.: 50-443
License No.: NPF-86
Report No.: 05000443/2013008
Licensee: NextEra Energy Seabrook, LCC (NextEra)
Facility: Seabrook Station, Unit 1
Location: Seabrook, NH 03874
Dates: March 25 to April 26, 2013
Inspectors: K. Mangan, Senior Reactor Inspector, Division of Reactor Safety (DRS), Team Leader S. Pindale, Senior Reactor Inspector, DRS J. Richmond, Senior Reactor Inspector, DRS J. Schoppy, Senior Reactor Inspector, DRS W. Sherbin, NRC Mechanical Contractor S. Kobylarz, NRC Electrical Contractor
Approved by: Paul G. Krohn, Chief Engineering Branch 2 Division of Reactor Safety
ii Table of Contents
SUMMARY OF FINDINGS
..........................................................................................................
iii
REPORT DETAILS
................................................................................................................
REACTOR SAFETY
.............................................................................................................. 1
1R21 Component Design Bases Inspection ...................................................................... 1
OTHER ACTIVITIES
........................................................................................................... 25
4OA2 Identification and Resolution of Problems .............................................................. 25
4OA6 Meetings, including Exit .......................................................................................... 26
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
.................................................................................................... A-1
LIST OF ITEMS
OPENED, CLOSED AND DISCUSSED ........................................................ A-1
LIST OF DOCUMENTS REVIEWED
........................................................................................ A-2
LIST OF ACRONYMS
............................................................................................................ A-13
iiiSUMMARY
- OF [[]]
LLC; Seabrook Station, Unit 1; Component Design Bases Inspection.
The report covers the Component Design Bases Inspection conducted by a team of four
NRC contractors. Four findings of very low risk significance (Green) were
identified; three of the findings were considered to be non-cited violations (NCV), and one finding involved not meeting a licensee imposed standard. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter
(IMC) 0609, "Significance Determination Process" (SDP). Findings for which the
NRC management review. The NRC's
program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Proce
ss," Revision 4, dated December 2006.
- A. [[]]
NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
Green. The team identified a finding of very low safety significance involving a non-cited violation (NCV) of the
III, "Design Control," in that, NextEra did not appropriately select and review, for suitability of application, a safety-
related over-current protection device for a safety related power panel (EDE-PP01B).
Specifically, NextEra did not consider the effects of the current-limiter function of safety
related inverters, which supplied the safety related power panel and would limit fault
current at the over-current protection device. As a result, the safety related over-current protective devices would not have prevented a postulated fault of a non-safety related load, supplied from the safety related power panel, from causing a momentary loss of
voltage to the power panel and all associated safety related loads. In response, NextEra
entered the issue into their corrective action program and performed a preliminary
analysis that determined an existing non-safety related fuse would provide adequate over-current protection. NextEra credited the use of this fuse as an interim compensatory measure in their operability assessment in order to conclude the system
was operable. The team determined the analysis and associated assessment were
reasonable.
The finding was more than minor because it was similar to Example 3.j of
- NRC [[]]
IMC 0612, Appendix E, and was associated with the Design Control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the
availability, reliability, and capability of systems that respond to initiating events to
prevent undesirable consequences. The team determined the finding was of very low
safety significance because the issue was a
design or qualification deficiency that did not result in inoperablity of the system. This finding was not assigned a cross-cutting aspect because the underlying cause was not indicative of current performance.
(Section 1R21.2.1.3)
iv Green. The team identified a finding of very low safety significance involving an
CFR Part 50, Appendix B, Criterion III, "Design Control," in that NextEra did not assure the seismic design requirements for the condensate storage tank (CST) were translated into specifications and procedures. Specifically, the team found that
NextEra's seismic design calculations for the CST was based, in part, on a maximum
tank level. The maximum tank level was used to ensure that the floating cover inside the
CST would not strike the top of the tank. NextEra engineers had concluded that this
impact could cause a failure of the
- CST or cover during a seismic event. However, the team identified that the high level alarm and operating procedure limits for the tank were above the level credited in the calculation. Additionally, the team determined that NextEra routinely operated the
CST tank above the maximum level assumed in the
calculation. Following identification NextEra entered the issue into their corrective action
program and proceduralized a lower maximum allowable water level for the CST to prevent a seismically induced impact of the floating cover on the tank. The finding is more than minor because it is associated with the protection against
external factors (seismic event) attribut
e of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the reliability of systems that respond to
initiating events to prevent undesirable consequences. The finding involved the loss or degradation of equipment so a detailed risk
evaluation (DRE) was performed. Based upon the DRE, the finding was determined to be of very low safety significance. This
finding was not assigned a cross-cutting aspect because the underlying cause was not
indicative of current performance. (Section 1R21.2.1.11)
Green. The team identified a finding of very low safety significance, in that NextEra did not perform preventative maintenance (PM) on supplemental emergency power system (SEPS) electrical components as re
quired by the approved engineering design modification for SEPS. As a result, the system's reliability to respond to a loss of off-site power event had not been maintained at a high confidence level, as assumed in
NextEra's design and probabilistic risk analyses. In response, NextEra entered the issue
into their corrective action program, evaluated the effect on equipment reliability for the
never-performed PMs, and implemented an accelerated schedule to complete the missed PM tasks.
The finding was more than minor because, if left uncorrected, it had the potential to lead
to a more significant safety concern. In addition, the finding was associated with the
Procedure Quality and Equipment Performance attributes of the Mitigating Systems
Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. The team determined the finding was of very low safety significance because it was not a design or qualification
deficiency and did not result in the loss of the SEPS system or train function. This
finding had a cross-cutting aspect in the area of Human Performance, Decision Making, because the most recent quarterly system health report (4th quarter 2012) had stated that
- SEPS [[]]
PMs had not been scheduled or perfo
rmed, these reports had been reviewed by NextEra management, however actions were not taken to develop electrical
SEPS. [IMC 0310, Aspect H.1(a) (Section 1R21.2.1.17)
Green. The team identified a finding of very low safety significance involving an
CFR Part 50, Appendix B, Criterion III, "Design Control," in that NextEra did not verify
vthe design basis for the primary component cooling water (PCCW) system had been translated into specifications and procedures. Specifically, the team found that NextEra had produced engineering evaluations and maintenance procedures that allowed a limited amount of leakage past the "B" train PCCW isolation valves. The team noted
NextEra used these documents to conclude that a 2.5 gallons per minute (gpm) leak rate
identified in April 2011 and a 4 gpm leak identified in October 2012 on "B" train valves
were acceptable. The team reviewed the design and licensing basis of the "B" train and
determined the system did not have a safety related refill capability and, therefore, was required to be leak tight. The team determined that, with leakage past the isolation valves, water would need to be added to the system and concluded that following certain
design basis events a safety related refill system would not be available resulting in loss of the PCCW system. Following identification of the issue NextEra entered it into their
corrective action program and evaluated the operability of system, concluding the PCCW system was operable based on recent valve-leakage testing results. The team review of the evaluation determined it to be reasonable. The finding is more than minor because it is associated with the protection against
external factors (seismic event) attribut
e of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding involved the loss or degradation of equipment designed to mitigate a seismic initiating
event and resulted in a
IMC 0609, Appendix A, Exhibit 4.
Based upon the DRE, the finding was determined to be of very low safety significance.
The team determined that this finding has a cross-cutting aspect in the area of Human
Performance, Resources, because NextEra did not ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety. Specifically, engineering evaluations and maintenance procedures associated with PCCW isolation valves did not align with the design and licensing basis requirements for
a leak tight system. H.2(c) (Section 1R21.2.2.1)
B. Licensee-Identified Violations
None
- REPORT [[]]
DETAILS
1. REACTOR SAFETY
Cornerstone: Initiating Events, Mitigating Systems, Barrier Integrity
1R21 Component Design Bases Inspection (IP 71111.21)
.1 Inspection Sample Selection Process
The team selected risk significant components for review using information contained in
the Seabrook Station Probabilistic Risk Assessment (PRA) and the U.S. Nuclear
Regulatory Commission's (NRC) Standardized Plant Analysis Risk (SPAR) model.
Additionally, the Seabrook Station Unit 1
Significance Determination Process (SDP) analysis was referenced in the selection of potential components for review. In general,
the selection process focused on components that had a Risk Achievement Worth
(RAW) factor greater than 1.3 or a Risk Reduction Worth (RRW) factor greater than
1.005. The team also selected components based on previously identified industry operating experience issues and the component contribution to the large early release
frequency (LERF) was also considered. The components selected were located within both safety-related and non-safety related systems, and included a variety of
components such as pumps, breakers, heat exchangers, electrical buses, transformers,
and valves.
The team initially compiled a list of components based on the risk factors previously
mentioned. Additionally, the team reviewed the previous component design bases inspection report (05000443/2007007 and 05000443/2010007) and excluded those
components previously inspected. The team then performed a margin assessment to narrow the focus of the inspection to 18 components and 4 operating experience (OE) samples. One component was selected because it was a containment-related structure, system, and components (SSC) and was considered for LERF implications. The team's evaluation of possible low design margin included consideration of original design
issues, margin reductions due to modifications, or margin reductions identified as a
result of material condition/equipment reliability issues. The assessment also included
items such as failed performance test results, corrective action history, repeated maintenance, maintenance rule (a)1 status, operability reviews for degraded conditions, NRC resident inspector insights, system health reports, and industry operating
experience. Finally, consideration was given to the uniqueness and complexity of the
design and the available defense-in-depth margins.
The inspection performed by the team was conducted in accordance with NRC Inspection Procedure 71111.21. This inspection effort included walkdowns of selected
components, interviews with operators, system engineers and design engineers, and
reviews of associated design documents and calculations to assess the adequacy of the
components to meet design and licensing basis. A summary of the reviews performed
for each component, operating experience sample, and the specific inspection findings identified are discussed in the subsequent sections of this report. Documents reviewed for this inspection are listed in the Attachment.
.2 Results of Detailed Reviews
.2.1 Results of Detailed Component Reviews (18 samples)
.2.1.1 Vital DC Switchgear Bus "11B"
a. Inspection Scope
The team inspected the "11B" vital direct-current (DC) switchgear bus (EDE-SWG-11-B)
to determine if it was capable of meeting its design basis requirements. The team
reviewed the design and operation of the switchgear bus and associated distribution
panels. The review evaluated whether the loading of the DC bus was within equipment ratings and determined whether the bus could perform its design basis function to reliably power the associated loads under worst case conditions. Specifically, the team
reviewed calculations and drawings including voltage drop calculations, short circuit
analyses, and load study profiles to evaluate the adequacy and appropriateness of
design assumptions. The team also reviewed the DC overcurrent protective coordination studies to determine if there was adequate protection for postulated faults
in the DC system.
The team interviewed system and design engineers and walked down the 125 volt direct
current (Vdc) bus and distribution panels to independently assess its material condition and to determine whether the system alignment and operating environment was consistent with design basis assumptions. Finally, the team reviewed corrective action documents and system health reports to det
ermine whether there were any adverse operating trends and to assess NextEra's ability to evaluate and correct problems.
b. Findings
No findings were identified.
.2.1.2 Solid State Protection System "B"
a. Inspection Scope
The team inspected the solid state protection system (SSPS) "B" train control panel and
relays (MM-CP-13) to determine if they were capable of meeting their design basis
requirements. Specifically, the team inspected the design, testing, and operation of the
SSPS and associated relays to determine if they could perform their design basis function to actuate the reactor trip breakers upon a valid reactor trip signal and actuate engineered safety features upon a valid initiation signal. The team reviewed functional
logic diagrams, technical specifications, and vendor specifications to determine the
performance requirements. The team reviewed maintenance, surveillance, and test
procedures to determine whether the established acceptance limits were adequate to
ensure reliable operation and that the equipment performed in accordance with design and licensing basis requirements, industry standards, and vendor recommendations. The team also compared as-found and as-left inspection and test results to the
established acceptance criteria in order to determine if the SSPS logic and relay test results met the established criteria. A
dditionally, the team interviewed system and design engineers and walked down accessible portions of the SSPS system to independently assess the material condition of the system, and to determine if the
system alignment and operating environment were consistent with design assumptions. Finally, the team reviewed corrective action documents and system health reports to
determine if there were adverse trends associated with the SSPS system and to assess NextEra's ability to evaluate and correct problems.
b. Findings
No findings were identified.
.2.1.3 Vital Alternating Current Power Panel "1B"
a. Inspection Scope
The team inspected the vital alternating current (AC) power panel "1B" (EDE-PP-1-B) to determine if it was capable of meeting its design basis functions. Specifically, the team evaluated the the inverter's capability to provide power to safety-related loads including
the nuclear instrumentation, reactor protection, and the engineered safety features
actuation systems. The team reviewed the loading documentation that determined the
design basis for maximum loading and the inverter equipment vendor ratings for conformance with the design basis. The team also reviewed calculations to determine if the inverter was capable of providing the 120Vac system loads with adequate voltage
during design basis conditions. Additionally, the team reviewed a common mode failure
analysis and the inverter qualification testing in order to determine if there was adequate
clearing for the 120Vac system branch circuits during fault conditions. The team conducted walkdowns at the inverter to assess the observable material condition and to evaluate if the installation was in accordance with manufacturer instructions. The team
also reviewed the operating and surveillance procedures to determine if the 120Vac
system voltage limits were correctly incorporated. Finally, the team reviewed corrective action documents and system health reports to determine whether there were any
adverse operating trends and to assess NextEra's ability to evaluate and correct
problems.
b. Findings
Introduction: The team identified a finding of very low safety significance (Green) involving an
CFR Part 50, Appendix B, Criterion III, "Design Control," in that NextEra's design calculations did not verify that safety related breakers would perform
the intended safety function. Specifically, NextEra's fault analysis for the breaker, that
supplied non-safety related loads from a 120Vac safety related power panels, did not
evaluate the impact the current limiter function of the safety related inverters (which
supplied the power panels) would have on fault current. As a result the safety related breaker over-current protective device was not adequate to prevent a postulated fault
from a non-safety related load from actuating the inverter's current limiter resulting in a loss of voltage to the safety related power panel and associated loads.
Description: The team determined the 120Vac vital power panel, EDE-PP-1B, was powered from inverter EDE-I-1B and noted the panel powered both safety related and
non-safety related loads. The team reviewed Calculation 9763-3-ED-00-46-F, "Failure of
Non-Class 1E Loads on Class 1E Buses," performed to ensure that the failure of
non-safety related loads would not have any detrimental effect on the operation of safety related busses (including panel EDE-PP-1B). The team found the calculation compared
the time-current characteristic curve for inve
rter's overload current sensing relay (OCSR) with the power panel's safety related load circuit breakers that supply non-safety related
loads. The calculation concluded that the panel was adequately protected from non-safety related load faults because there was sufficient coordination between the two over-current protective devices. The team reviewed the calculation for power panel
EDE-PP-1B circuit 10, a non-safety related ground detection panel, to evaluate if the
breaker would clear the fault prior to operation of the
ED-00-68-F, "120VAC Breaker Coordination," which stated that inverter full load output current was 65 amps (A) and the
overload current limit was 150 percent of the full load output current. In addition, the
team found that Appendix C of the calculation described the inverters output response to
an overload or fault for the inverter as follows, "When the full load current is exceeded,
the output voltage will be reduced to maintain the output current less than the current limit value of 150%. Depending on the current magnitude, the output breaker will be tripped based on the time response of the Heinemann relay" (which is the OCSR) "and
the CSRT timer. If the overload/fault clears before the relay times out, the inverter
output will be restored to full voltage." Finally, the team reviewed the vendor technical
manual, "Instrument Bus Inverter Instruction Manual," which stated that the inverter output current would not exceed 150 percent of rated output current in the event of a fault or short circuit.
The team determined that calculation 9763-3-ED-00-46-F did not evaluate the response
of the current limiter function of the inverter on bus voltage with a fault current present.
Because this evaluation was not performed, the power panel's load circuit breakers, although coordinated with the inverter's OCSR relay setting, were not coordinated with the inverter's current limiter. The team further determined that a postulated fault from a
non-safety related load would cause the inverter current limiter to reduce output voltage
to prevent the output current from exceeding 97.5A. The team concluded that, as a
result of a non-safety related load fault, the power panel and all associated safety related loads might be at a reduced voltage for as long as 10 seconds before the safety related breaker supplying the non-safety related load would trip.
Following identification of the issue NextEra performed an evaluation and determined that, if the inverter's output current exceeded the current limited value, the voltage would rapidly collapse. As a result, NextEra concluded that the power panel's safety related
amp circuit breaker would not provide adequat
e over-current protection from a fault on a non-safety related load. NextEra also determined that the lack of adequate
coordination was limited to the ground detector circuit on the bus.
NextEra entered this issue into their corrective action program and performed an operability determination on EDE-PP-1B to assess whether it remained adequately
protected from a non-safety related load fault. NextEra found that a non-safety related
1A fuse located in the ground detection panel could provide adequate over-current
protection and proper coordination with the inverter's current limiter. NextEra noted that, in addition to the fuse, the power feeder cable from the power panel to the ground detection panel was not safety related, but was run in seismically mounted conduit. The
operability evaluation, crediting the fuse as the over-current protective device, concluded
the equipment was operable but non-conforming to licensing and design basis
requirements because of the reliance on the operation of non-safety related components. The team reviewed NextEra's preliminary analysis and operability evaluation and concluded they were reasonable.
Analysis: The team determined that the failure to verify the adequacy of over-current protection to ensure that non-safety load faults could not adversely affect safety related
equipment was a performance deficiency. Specifically, NextEra did not consider the effects of fault current on the current limiter function of inverters which supplied safety related power panels. As a result, a postulated fault of a non-safety related load could
have resulted in a loss of voltage to the safety related power panel and all associated
safety related loads. The finding was more than minor because it was similar to
Example 3.j of
- NRC [[]]
IMC 0612, Appendix E, "Examples of Minor Issues," which determined that calculation errors would be more than minor if, as a result of the errors, there was reasonable doubt on the operability of the component. In addition, the finding
was associated with the Design Control attribute of the Mitigating Systems Cornerstone
and adversely affected the cornerstone objective to ensure the availability, reliability, and
capability of systems that respond to initiating events to prevent undesirable
consequences (i.e., core damage).
The team performed a risk screening, in accordance with IMC 0609, Appendix A,
"Significance Determination Process for Findings At-Power," using Exhibit 2, "Mitigating Systems Screening Questions," and determined the finding was of very low safety
significance (Green) because it was a design or qualification deficiency affecting the over-current protective device that did not result on the loss of operability or functionality of safety related equipment. Specifically, a preliminary analysis determined that existing
fuses, although not qualified as safety related, would provide adequate over-current
protection. This finding did not have a cross-cutting aspect because it was determined
to be a legacy issue and was not considered to be indicative of current licensee
performance.
Enforcement
III, "Design Control," requires, in part, that design control measures shall provide for verifying the adequacy of design, and for the selection and review for suitability of equipment that is essential to the safety
related functions of the structures, systems, and components. Contrary to the above,
prior to March 29, 2013, NextEra's design control measures had not verified the
adequacy of the design regarding protection of safety related power panels from non-
safety related load faults. Specifically, NextEra did not consider the effects of fault current on the current limiter function of safety related 120 Vac inverters which supplied
safety related power panels. Because this violation was of very low safety significance (Green) and was entered into NextEra's corrective action program (CR 1861161), this
violation is being treated as a non-cited violation (NCV), consistent with Section 2.3.2 of
the NRC's Enforcement Policy. (NCV 05000443/2013008-01, Failure to Verify Adequate Fault Protection for Safety Related Equipment from Non-Safety Related
Load Fault)
.2.1.4 Unit Auxiliary Transformer Breaker to Bus "E6"
a. Inspection Scope
The team inspected the 4.16 kV supply circuit breaker (EDE-BK-ACB-B71) to determine
if it was capable of meeting its design basis functions. Specifically, the team evaluated
whether if the breaker would ensure offsite power would stay connected to the vital bus
when available. The team reviewed applicable portions of the Updated Final Safety Analysis Report (UFSAR), design basis documents (DBD), and drawings to identify the design basis requirements for the breaker. The team reviewed schematic diagrams and calculations for the circuit breaker protective relays to determine whether the circuit
breaker was subject to spurious tripping and was properly coordinated with downstream
non-Class 1E-load feeder breakers. The team also reviewed the bus-transfer schemes and synchronism check relaying for the 4.16 kV breaker to determine whether they would enable continuity of offsite power to the safety buses when available and isolate
the safety bus from the non-safety 4.16 kV system when required. Additionally, the team reviewed maintenance schedules, procedures, and completed work records to
determine whether the breaker was being properly maintained. The team reviewed
corrective action documents and system health reports to determine whether there were any adverse operating trends and to assess NextEra's ability to evaluate and correct problems. Finally, the team performed a visual inspection of the breaker to assess the
material condition and operating environment of the equipment.
b. Findings
No findings were identified.
.2.1.5 Emergency Feedwater Pump Minimum Flow Valve
a. Inspection Scope
The team inspected the emergency feedwater (EFW) minimum flow recirculation valve
(FW-V-347) to determine if it was capable of performing its design basis functions.
Specifically, the team determined if the valve would reposition, as required, to ensure adequate flow was available for the
UFSAR, Technical Specifications (TS), TS Bases, and the in-service
testing (IST) basis documents to identify the design basis requirements of the valve.
The team reviewed periodic motor operated valve (MOV) diagnostic test results and
stroke-timing test data to verify acceptance criteria were met. The team also evaluated whether the MOV safety functions, performance capability, torque switch configuration, and design margins were adequately monitored and maintained in accordance with
generic letter (GL) 89-10 guidance. The team reviewed MOV weak link calculations to
ensure the ability of the valve to remain structurally functional while stroking under
design basis conditions. The team verified that the valve analysis used the maximum differential pressure expected across the valve during worst case operating conditions. Additionally, the team reviewed motor data, degraded voltage conditions, and voltage
drop calculation results to confirm that the MOV would have sufficient voltage and power
available to perform its safety function at degraded voltage conditions. The team
discussed the design, operation, and component history of the valve with engineering
and operations staff to determine performance history and overall component health. The team also conducted a walkdown of the valve to assess its material condition and determine if the installed configuration was consistent with plant drawings, procedures,
and the design basis. Finally, the team reviewed corrective action documents and
system health reports to determine whether there were any adverse operating trends and to assess NextEra's ability to evaluate and correct problems.
b. Findings
No findings were identified.
.2.1.6 Component Cooling Water Bypass Temperature Control Valve
a. Inspection Scope
The team inspected the primary component cooling water (PCCW) bypass temperature control valve (CC-TV-2271-2) to determine if the air operated control valve was capable of performing its design basis functions. Specifically, the team evaluated whether the valve would throttle flow to the
temperature within limits. The team reviewed the UFSAR, TSs, DBD, and drawings to
identify the design basis requirements of the valve. The team identified that the valve
fails closed and evaluated whether the failure mode was consistent with its safety function. The team also evaluated nitrogen backup bottle volume calculations to ensure that sufficient nitrogen would be provided to throttle the valve as required during a
design basis accident. The team evaluated whether the instrument setpoints were
properly translated into system procedures and tests and reviewed completed tests to determine if the results demonstrated component operability. The team reviewed valve and system calculations to determine if the inputs and assumptions were accurate and
justified. The team also conducted several walkdowns of the valve to assess its material
condition and to evaluate if the installed configuration was consistent with the plant
drawings, procedures, and the design basis. Finally, the team reviewed corrective
action documents and system health reports to determine whether there were any adverse operating trends and to assess NextEra's ability to evaluate and correct
problems.
b. Findings
No findings were identified.
.2.1.7 Startup Feed Pump/Emergency Feedwater Cross Connect MOV
a. Inspection Scope
The team inspected the startup feed pump/emergency feedwater cross-connect
FW-V-163) to determine if the valve was capable of performing the function credited in
the PRA and its design basis functions. Specifically, the team evaluated whether the normally closed valve when opened would provide an adequate flow path from the
startup feed pump to the steam generators and provide the required isolation between the feedwater and
UFSAR, TSs, drawings, PRA, and procedures to identify the performance requirements for the valve. The team
reviewed periodic MOV diagnostic test results and stroke-timing test data to verify
acceptance criteria were met. The team evaluated whether the MOV performance
capability, torque switch configuration, and design margins were adequately monitored and maintained in accordance with NextEra's
MOV weak link calculations to ensure the ability of the MOV to remain
structurally functional while stroking under worst case operating conditions. The team
verified the MOV valve analysis used the maximum differential pressure expected
across the valve during worst case operating conditions. Additionally, the motor data,
degraded voltage conditions, and voltage drop calculation results were reviewed to confirm that the MOV would have sufficient voltage and power available to perform its function at degraded voltage conditions. The team discussed the design, operation, and
maintenance of the valves with the system engineer to determine the valves
performance history, maintenance, and overall health. Additionally, the team conducted a walkdown of the valve and associated equipment to assess the material condition of the equipment and to determine if the installed configuration was consistent with the plant drawings and design. Finally, the team reviewed corrective action documents to determine if there were any adverse trends associated with the valves and to assess
NextEra's capability to evaluate and correct problems.
b. Findings
No findings were identified.
.2.1.8 Service Water Pump "D"
a. Inspection Scope
The team inspected the "D" service water (SW) pump (SW-P-41-D) to evaluate if it was capable of performing its design basis functions. Specifically, the team evaluated
whether the
SW system was capable of
transferring the maximum heat loads, from primary and secondary sources in the plant, to the environment. The team reviewed applicable portions of the UFSAR, DBD, and drawings to identify the design basis requirements for the pump. The team evaluated
whether the pump capacity was sufficient to provide adequate flow to the safety-related
components supplied by the
DBAs). The
team reviewed design calculations to assess available pump net positive suction head (NPSH), worst case pump run-out conditions, and to evaluate the capability of the pump to provide required flow to supplied com
ponents. Additionally, the team reviewed the SW pump motor data, degraded voltage conditions, and voltage drop calculations to
confirm that the pump motor would have sufficient voltage and power available to
perform its safety function at degraded voltage conditions. The team reviewed the
IST results and SW system flow verification tests to determine if adequate system flow was available. Specifically, the team reviewed pump data trends for vibration, pump
differential pressure, and flow rate test results to verify acceptance criteria were met and
acceptance limits were adequate. The team ensured changes that impacted flow
requirements to individual SW system loads due to changes in fouling factors, pipe
replacement, modifications, and revised heat load requirements for components were properly evaluated. The team interviewed the system engineer and performed several
walkdowns of the pump to evaluate its material condition and assess the pump's operating environment. The team also reviewed SW intake inspection reports (including
underwater videos) and walked down accessible portions of the SW intake transition
structure to assess the material condition of the SW intake support structures, intake silt/debris loading, and NextEra's configuration control. Finally, the team reviewed corrective action documents and system health reports to determine whether there were
any adverse operating trends and to assess NextEra's ability to evaluate and correct
problems.
b. Findings
No findings were identified.
.2.1.9 Primary Component Cooling Water Pump "D"
a. Inspection Scope
The team inspected the "D"
CC-P-11-D) to determine if it was capable of
meeting its design basis function. Specifically, the team evaluated the ability of the
- PCCW system to provide cooling water to essential components under normal, transient, and accident conditions. The team evaluated whether the pump capacity was sufficient to provide adequate flow to the safety-related components supplied by the system during
DBAs. The team also reviewed calculations for NPSH to ensure that the pump could
successfully operate under the most limiting conditions. The team reviewed drawings, calculations, hydraulic analyses, procedures, system health reports, and the system DBD to ensure consistency with design and licensing bases requirements. Additionally,
the motor data, degraded voltage conditions, and voltage drop calculation results were
reviewed to confirm that the pump motor would have sufficient voltage and power
available to perform its safety function at degraded voltage conditions. The team also
reviewed completed pump surveillance tests to ensure pump performance and procedure acceptance criteria were consistent with system flow calculations. The team walked down the PCCW pumps and accessible portions of the system and reviewed the
system health report and maintenance records to assess NextEra's configuration control,
operating environment of the pumps, and the
system's overall material condition. Finally, the team reviewed corrective action documents and system health reports to determine whether there were any adverse operating trends and to assess NextEra's ability to evaluate and correct problems.
b. Findings
No findings were identified (see Section 1R21.2.2.1 for PCCW-related finding).
.2.1.10 Containment Enclosure Fan "A"
a. Inspection Scope
The team inspected the "A" containment enclosure fan (EAH-FN-5-A) to determine if it was capable of meeting its design basis functions. Specifically, the team evaluated whether the fan capacity was sufficient to provide adequate flow for heat removal from
safety-related components during design basis events. The team reviewed design
documents and drawings in order to determine the minimum fan flow requirements required to provide adequate cooling. Additionally, the motor data, degraded voltage conditions, and voltage drop calculation results were reviewed to confirm that the fan
would have sufficient voltage and power available to perform its safety function at
degraded voltage conditions. The team reviewed inspection and testing procedures to
evaluate whether appropriate maintenance activities were being performed and
reviewed past test results to determine if the fan was capable of removing the required heat load. The team conducted a walkdown of the fan and associated ventilation equipment and interviewed engineers regarding the maintenance and operation of the
fan, in order to assess the material condition of the system. Finally, the team reviewed
corrective action documents and system health reports to determine whether there were
any adverse operating trends and to assess NextEra's ability to evaluate and correct
problems.
b. Findings
No findings were identified.
2.1.11 Condensate Storage Tank
a. Inspection Scope
The team inspected the
CO-TK-25) to determine if it was capable of meeting its
design basis function. Specifically, the team evaluated whether the tank was adequately designed to provide the required quantity of water for the
- EFW system during design basis events. The team reviewed the design, testing, inspection, and operation of the
CST, and associated tank level instruments to evaluate whether the tank could perform
its design basis function as the water source for the emergency feedwater pumps.
Specifically, the team reviewed design calculations, drawings, and vendor specifications (including tank sizing and level uncertainty analysis, and pump vortex calculations) to evaluate the adequacy and appropriateness of design assumptions and operating limits. Seismic design documentation was reviewed to evaluate whether CST design
assumptions were consistent with limiting seismic conditions. The team interviewed
system and design engineers, reviewed instrument test records, and tank inspection results to determine whether maintenance and testing was adequate to ensure reliable operation. Additionally, the review evaluated whether those activities were performed in accordance with regulatory requirements, industry standards, and vendor
recommendations. The team also conducted a walkdown of the tank area to
independently assess the material condition of the CST and associated instrumentation.
Finally, the team reviewed corrective action documents and system health reports to determine whether there were any adverse operating trends and to assess NextEra's ability to evaluate and correct problems.
b. Findings
Introduction: The team identified a non-cited violation of
III, "Design Control," in that NextEra did not assure the seismic design requirements for the CST were translated into specifications and procedures.
Specifically, the team determined that NextEra routinely operated the CST tank at a
water level above that credited in the CST seismic design calculation developed to
ensure the tank would be available following a seismic event. Additionally, the team
determined that the CST high level alarm was set above this limit.
Description: The team found that the design and licensing basis of the
EFW pumps for
design basis events. The team noted that the tank is equipped with a floating stainless
steel cover; the cover is designed to freely move up and down; and a foam rubber gasket is mounted on the circumference of the plate in contact with the tank wall.
The team reviewed calculation
CST Seismic Report," Revision T that
evaluated, in part, the potential for the floating cover to impact the CST roof as a result
of water movement and wave action during a seismic event. NextEra had determined
that during a safe shutdown earthquake (SSE), the wave height could be over three feet. The team determined that, in order to ensure the integrity of the tank and cover, the calculation assumed that the CST water level was maintained at a height such that the
cover could not reach the tank roof during the event. The team then reviewed the tank level instrument setpoint calculation to evaluate the adequacy of the tank high level alarm and found that the calculation set the high level alarm (including instrument
uncertainty) at a level where the
CST roof
which was consistent with seismic design calculation. The team found that with
instrument uncertainty factored into the level setpoint, the high level alarm setpoint was
either 385,760 gallons or 391,150 gallons (depending upon which level instrument is used) in order to assure the limit established in the design calculations was maintained.
Following a review of the alarm setpoint maintenance and testing procedures, however,
the team determined that the procedure setpoints corresponded to a tank level of
401,300 gallons. The team also reviewed plant operating procedures for the
CST to the high level setpoint. The team discussed operation of the CST with plant operations personnel, who stated
that the tank is usually maintained at a level between 380,000 and 400,000 gallons of
water. The team reviewed the previous year's tank level records which confirmed the
tank was routinely operated above the limit established in the design calculation. Finally, the team noted during a walkdown in the control room that the tank level was above the calculation limit. Subsequent to the team's questions, NextEra entered the issue into
their corrective action program; developed a night order that procedurally limited the tank level to below the seismic calculation level limit; and performed evaluations to determine
if an impact from the cover on the roof of the tank affect the performance of safety
related component.
The team reviewed the evaluations and concluded that it was reasonable to assume that
the tank and steel cover would be unaffected during the event, however, the team
determined that the cover gasket was not designed or tested to be capable of
withstanding an impact on the
- CST and could not conclude that it would be unaffected by a seismic event. The team further determined that if the gasket failed it would sink and possibly impact the operation of the
EFW system.
Analysis: Failure to operate the CST in accordance with water level limits determined by seismic analysis is a performance deficiency. The finding was more than minor because
it was similar to Example 5.a. of
- NRC [[]]
IMC 0612, Appendix E, "Examples of Minor Issues," in that design calculations required the high level alarm to limit CST level, the alarm was set incorrectly, and the alarm was restored to service. Additionally, the
finding was associated with the Design Control attribute of the Mitigating Systems
Cornerstone and adversely affected the objective to ensure the availability, reliability,
and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the finding in accordance with IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations,"
and screened this external event issue under Mitigating Systems. In accordance with
IMC 0609, Exhibit 2.B, the finding involved the loss or degradation of equipment or function specifically designed to mitigate a seismic initiating event and triggered the use
of Exhibit 4, "External Events Screening Questions." The "yes" response to question 1 of Exhibit 4 directed a detailed risk evaluation (DRE) because if the gasket failed it could
degrade one or more trains of a system that supports a risk significant system or function.
Since a Seabrook specific seismic risk model has not been developed, a Region I Senior
Reactor Analyst (SRA) conducted a detailed analysis utilizing the External Initiator Risk
Informed Inspection Notebook for Salem Generating Salem, Table 5.3.1 as a surrogate.
This was deemed appropriate since both units are four loop Westinghouse plants with
large dry containments. Additionally, the safety functions employed to mitigate a design basis
EFW suctions and the failure
modes of the tank cover gasket, the failure probability of the EFW system was increased
from 1E-4 to 1E-1. This was deemed to be cons
ervative since there are two "T" suctions in the tank, the material would sink and the path to the pumps experiences several elevation changes which would make debris transport difficult. In addition no credit was given for off-site power. This was also conservative since the fragility analysis in the
Individual Examination for External Events (IPEEE) for Seabrook, Table 3.7, predicts a
greater than 50 percent success for offsite power give the SSE. Given these bounding
assumptions the change in core damage frequency was E-7. The dominant sequence was a
EFW and the failure to successfully conduct reactor feed and bleed operations. Large early release was determined not to
be applicable in accordance with Inspection Manual Chapter 0609, Appendix H,
Containment Integrity Significance Determination Process, Table 5.1. As a result the
finding was of very low safety significance (Green).
This finding was not assigned a cross-cutting aspect because the underlying cause was a legacy issue and not indicative of current performance.
Enforcement:
III, "Design Control," states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, prior to April 15, 2013, NextEra did not provide
the appropriate operating water level limit for the CST high level alarm or operating
procedures to prevent the tank's floating cover from impacting the seismic Class I CST
during a seismic event. However, because this finding is of very low safety significance
and was entered into the licensee's corrective action program (AR 01865544), the violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the
NCV 05000443/2013008-02, Condensate Storage Tank
Water Level Above Limits of Seismic Qualification) .2.1.12 Primary Component Cooling Water Heat Exchanger 17B Outlet Pipe Vacuum Breaker
a. Inspection Scope
The team inspected the PCCW heat exchanger "17B" service water outlet pipe vacuum
breaker (SW-V-175) to determine if it was capable of meeting its design basis function.
Specifically, the team evaluated whether the valve design and capacity was sufficient to provide adequate protection against a postulated water hammer in the heat exchanger's SW system discharge piping. The team reviewed design documents, in-service test
(IST) program documents, and drawings to evaluate the ability of the valve to provide adequate water hammer protection. The team also reviewed inspection and testing procedures to evaluate whether appropriate maintenance activities were being
performed and reviewed past test results to determine if the valve demonstrated
acceptable performance. The team conducted a walkdown of the valve and associated
equipment and interviewed engineers regarding valve maintenance and operation to
assess the material condition of the valve. Finally, the team reviewed corrective action documents and system health reports to ev
aluate whether there were any adverse operating trends and to assess NextEra's ability to evaluate and correct problems.
b. Findings
No findings were identified.
.2.1.13 Motor-Driven Emergency Feedwater Pump
a. Inspection Scope
The team inspected the motor-driven
FW-P-37-B) to determine if it was
capable of meeting its design basis functions. Specifically, the team evaluated whether the pump was capable of providing adequate flow to the steam generators during DBAs.
The team reviewed the
- EFW system hydraulic model and the design basis hydraulic analysis/calculations to verify that required total developed head (
TDH), NPSH, and pump run-out conditions had been properly evaluated under all DBA conditions. Additionally, the motor data, degraded voltage conditions, and voltage drop calculation
results were reviewed to confirm that the pump motor would have sufficient voltage and
power available to perform the intended safety function at degraded voltage conditions.
The team reviewed system operating procedures to ensure they were consistent with the
design requirements. The team also reviewed pump IST procedures, test results, and trends in test data to determine if pump performance was consistent with design basis
assumptions. Additionally, IST acceptance criteria were reviewed to verify appropriate
correlation to accident analyses requirements. Seismic design documentation was
reviewed to evaluate whether pump design was consistent with limiting seismic
conditions. The team also conducted a detailed walkdown of the pump and support systems to determine the material condition of the components and to ensure adequate configuration control. Finally, the team reviewed corrective action documents and
system health reports to evaluate whether there were any adverse operating trends and
to assess NextEra's ability to evaluate and correct problems.
b. Findings
No findings were identified.
.2.1.14 Emergency Diesel Generator "A" Mechanical Systems
a. Inspection Scope
The team inspected the "A" emergency diesel generator (EDG) (DG-1-A) mechanical
systems to determine if they were capable of supporting their design basis functions.
Specifically, the team evaluate whether the mechanical support systems for the
EDG could provide power to 4.16 kV electrical bus "E5" during normal operation, operational transients, and design basis accidents. The
team selected the EDG engine, fuel oil system, air start system, lube oil system, and
jacket water cooling system for an in-depth review. The team reviewed the UFSAR, the
TSs, operating procedures, and DBD to identify the design basis requirements for these
systems. The team also reviewed EDG surveillance test results, equipment operator logs, and operating procedures to ensure that the mechanical support systems were
operating as designed and within their vendor design limits. The team reviewed fuel oil
consumption calculations to verify TS requirements were adequate to meet design basis
loading conditions. The team reviewed lube oil sample and chemistry results to assess whether NextEra had performed timely analysis for wear and trending, identified potential adverse trends, and to determine if proper lubrication of system components
was being performed. The team reviewed the
EDG surveillance tests, and preventive maintenance (PM) activities for the lube oil and fuel oil filters to
ensure that NextEra replaced the filters prior to any adverse impact on engine operation.
The team also conducted several detailed walkdowns of the EDG and its support systems (including control room instrumentation) to visually inspect the physical/material condition, to assess the operating environment and potential hazards, and to ensure
adequate configuration control. Finally, the team reviewed corrective action documents and system health reports to evaluate whether there were any adverse operating trends and to assess NextEra's ability to evaluate and correct problems.
b. Findings
No findings were identified.
.2.1.15 4.16kV Vital Bus "E6"
a. Inspection Scope
The team inspected the 4.16kV vital bus "E6" (EDE-SWG-6) to determine if it was
capable of meeting its design basis functions. Specifically, the team evaluated whether the bus was capable of transferring supplied power to downstream loads during a
UFSAR, DBD, and drawings to identify the
design basis requirements for the bus. The team reviewed selected calculations for the electrical distribution system's load flow/voltage drop, degraded voltage protection,
short-circuit protection, and coordination to verify the adequacy and appropriateness of
design assumptions in the calculations. The team's review evaluated whether bus and breaker capacity would be exceeded and determined if bus voltages remained above minimum acceptable values under design basis conditions. The team also reviewed
switchgear protective device settings and breaker ratings to ensure that selective coordination was adequate for the protection of connected equipment during short-circuit conditions and when non-Class 1E loads on the
power system were postulated to fail. The team reviewed the preventive maintenance inspection and testing procedure and
associated test results to ensure that breakers were maintained in accordance with
industry and vendor recommendations. The team also reviewed calculations to
determine if adequate voltage would be available for the breaker closure and opening control circuit components and the breaker spring charging motors. Additionally, the team performed a visual inspection of observable portions of the safety-related 4.16 kV
switchgear to assess the installed configuration, material condition, environmental
condition, and potential vulnerability to hazards. Finally, the team reviewed corrective
action documents and system health reports to evaluate whether there were any adverse operating trends and to assess NextEra's ability to evaluate and correct
problems.
b. Findings
No findings were identified.
.2.1.16 Residual Heat Removal Heat Exchanger "A"
a. Inspection Scope The team inspected the "A" residual heat removal (RHR) heat exchanger (RH-E-9-A) to determine if it was capable of meeting its design basis function. Specifically, the team
evaluated the ability of the heat exchanger to adequately remove decay heat following a
postulated accident. The team reviewed applicable portions of the UFSAR, DBD, and
drawings to identify the design basis requirements for the heat exchanger. The team also reviewed design calculations to evaluate the capability of the heat exchanger to transfer the required heat load during normal operations and postulated accident
conditions. Additionally, the team reviewed the design and procedural controls for the
control valves associated with the heat exchanger to determine if required RHR system flow and temperature would be maintained under design conditions. The team
interviewed system and design engineers, reviewed corrective action documents, and performed a walkdown of the heat exchanger to assess the material condition of the equipment. Finally, the team reviewed co
rrective action documents and system health reports to evaluate whether there were any adverse operating trends and to assess
NextEra's ability to evaluate and correct problems.
b. Findings
No findings were identified.
.2.1.17 Supplemental Emergency Power System "B" Electrical
a. Inspection Scope The team inspected the "B" supplement
al emergency power system (SEPS) (SEPS-DG-2B), including auxiliary support systems, to determine if it was capable of meeting its licensing basis and
SEPS to supply power to safe shutdown loads following a loss of off-site power coincident with a failure of one EDG. The team reviewed loading analysis and
voltage regulation calculations to determine if appropriate design assumptions had been
translated into the SEPS design specificati
ons and operating instructions. The team reviewed analyses and surveillance testing to assess the SEPS's electrical capabilities under required operating conditions. Additionally, the team reviewed overcurrent protection, coordination, and short-circuit calculations to evaluate whether the SEPS
diesel generator was adequately protected with properly set protective devices during
test mode and emergency operation under worst case fault conditions. The team also
reviewed preventive maintenance records and surveillance test results to evaluate whether the test results satisfied the established acceptance criteria and were consistent with design assumptions. The team performed independent walkdowns of the SEPS
and associated support equipment to assess the installation, configuration control,
material condition, and potential vulnerability to external hazards. Finally, system health
reports and component maintenance histories were reviewed to evaluate whether there
were any adverse operating trends and to assess NextEra's ability to evaluate and
correct problems.
b. Findings
Introduction: The team identified a finding of very low safety significance (Green), in that NextEra did not perform
SEPS electrical components as required by the approved engineering design modification for SEPS. Specifically, 4kV breakers,
480VAC breakers, protective relays, power distribution components, and battery
chargers had not been maintained in accordance with vendor recommendations,
industry standards, or Seabrook maintenance practices since initial installation (2003 to
2005). As a result, the system's reliability to respond to a loss of off-site power event had not been maintained at a high confidence level, as assumed in NextEra's design and probabilistic risk analyses.
Description: The team evaluated
- PM [[to verify that its capacity, capability, and reliability were maintained consistent with design and licensing bases assumptions. The team's evaluation of the scope of routine functional tests determined the testing did not include all of the critical components within the]]
- SEPS system that had to function, in order to deliver power to the Seabrook safety buses. The team also reviewed
PM tasks performed on SEPS equipment and found that a
significant number of critical components did not have any PM tasks assigned, scheduled, or performed and they did not have any inspections, tests, or calibrations since initial installation (2003-2005).
The team's review of the SEPS modification found that the following equipment had
been installed:
two diesel generators (DGs) 4kV tie-bus load bank underground power cable from the local tie-bus to a manual transfer switch power cables from manual transfer switch to safety bus-5 and -6 4kV breakers on the local tie-bus and safety bus-5 and -6 support systems (e.g., engine starting batteries, tie-bus switchgear battery, battery charges, and tie-bus protective relays)
The team's review of functional testing on the equipment identified that load tests at
different load ratings on a monthly, annual, and biennial frequency were performed.
However, the team determined that these tests were performed using the local load bank; as a result, the power circuit from the local tie-bus to the plant safety bus-6 had not been energized or tested since initial site acceptance testing, in 2005, and the power
circuit from the SEPS manual transfer switch to safety bus-5 had never been energized.
The team also determined that PM tasks had not been performed on many of these
untested components including the 4kV breakers (local tie-bus and safety bus), 480 Vac breakers, protective relays, battery charges, manual transfer switch, or underground
power cables.
The team reviewed
DCR002 and
determined that, as part of the engineering approved modification, NextEra had identified that numerous PM tasks were required in order to satisfy vendor recommendations, industry standards, and Seabrook maintenance practices (i.e., site
specific standards). As part of the modification process, NextEra had issued corrective
action program items CR-05-13019, CR-05-13020, and CR-05-13021, which were
created to ensure that appropriate PM tasks were developed and scheduled. The team
determined that the corrective actions were never completed.
The team also reviewed NextEra's safety assessment of risk and the associated revision to the plant's
SEPS modification. The
team determined that the risk assessments were based on a SEPS failure rate model
that reflected generic diesel generator failure data from NUREG/CR-6928, "Industry-Average Performance for Components and Initiating Events at Nuclear Power Plants." The team determined that these failure rates were based in part on the completion of
typical maintenance practices as recommended by industry standards and emergency diesel generator vendors. Therefore, the team concluded that SEPS availability and
reliability had not been maintained at a high confidence level, so as to validate and align
with NextEra's design and PRA assumptions.
In response, NextEra entered this issue into their corrective action program, evaluated
the effect on equipment reliability for the never performed PMs, and implemented an accelerated schedule to create and complete the missed PM tasks. During the team's
last week on-site, a number of first time
SEPS 4kV breakers on the plant safety buses. The team noted that no problems were identified that would have prevented proper operation of the equipment. Finally,
the team reviewed NextEra's functionality assessment of the SEPS system availability, created to allow Nextera to continue crediting the system for risk reduction analysis, and
concluded that it was reasonable.
Analysis: The team determined that the failure to perform PM tasks, as required by an approved design modification, without a technical evaluation of the impact of delayed
maintenance on the equipment, was a performance deficiency. Specifically, as part of
the
PM tasks
were required to be created and implemented. However, the
- PM tasks were never created. As a result, preventative maintenance that NextEra had determined to be necessary to maintain
SEPS reliability had not
been performed in the 8 year period since installation. The finding was more than minor because, if left uncorrected, it had the
potential to lead to a more significant safety concern (e.g., an on-demand failure to run).
In addition, the finding was associated with the Procedure Quality and Equipment Performance attributes of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that
respond to initiating events to prevent undesirable consequences (i.e., core damage).
The team performed a risk screening, in accordance with IMC 0609, Appendix A,
"Significance Determination Process for Findings At-Power," using Exhibit 2, "Mitigating Systems Screening Questions," and determined the finding was of very low safety significance (Green) because it was not a design or qualification deficiency and did not result in the loss of the SEPS system or train function.
This finding had a cross-cutting aspect in the area of Human Performance, Decision
Making, because the most recent quarterly system health report (4th quarter 2012) had stated that
- SEPS [[]]
PMs had not been scheduled
or performed, these reports had been reviewed by NextEra management, however actions were not taken to develop electrical
SEPS. [IMC 0310, Aspect H.1(a)
Enforcement: This finding did not involve a violation of regulatory requirements because
- SEPS was not safety related and was not relied upon to respond to design basis accidents. Because this finding did not involve a violation of regulatory requirements and was very low safety significance, it was identified as a finding (
FIN). NextEra
entered this issue into their corrective action program (CR 1862758) took immediate
corrective action to evaluate the effect on equipment reliability for the never-performed PMs, and implemented an accelerated schedule to complete the missed PM tasks.
(FIN 05000443/2013008-03, Failure to Perform Preventative Maintenance on the
Supplemental Emergency Power System)
.2.1.18 Vital Unit Substation 52
a. Inspection Scope
The team inspected the 480 Vac vital unit substation (EDE-US-52) to determine if it was
capable of performing its design basis functions. Specifically, the team evaluated
whether the bus and associated supply transformer were capable of transferring supplied power to downstream loads following a DBA. The team reviewed electrical distribution calculations including load flow, voltage drop, short-circuit, and electrical
protection coordination to evaluate the adequacy and appropriateness of design
assumptions. The team also determined if substation capacity and voltages remained
within acceptable values under design basis conditions. The team reviewed the electrical overcurrent protective relay settings for the substation supply breaker and selected load center breakers to determine if the trip setpoints would ensure the ability of
the supplied equipment to perform its design basis safety function and provide adequate
load center protection during fault conditions. Additionally, the team reviewed system maintenance test results, interviewed system and design engineers, and conducted field walkdowns to verify that equipment alignment, nameplate data, and breaker positions were consistent with design drawings and to assess the material condition of the load
center. Finally, the team reviewed corrective action documents and system health reports to evaluate whether there were any adverse operating trends and to assess
NextEra's ability to evaluate and correct problems.
b. Findings
No findings were identified.
.2.2 Review of Industry Operating Experience and Generic Issues (5 samples)
The team reviewed selected OE issues for applicability at the Seabrook Station. The
team performed a detailed review of the OE issues listed below to evaluate if NextEra had appropriately assessed potential applicability to site equipment and initiated
corrective actions when necessary.
.2.2.1 NRC Information Notice 2011-14, Component Cooling Water System Gas Accumulation
and Other Performance Issues
a. Inspection Scope
The team assessed NextEra's applicability review and disposition of
IN) 2011-14. This IN discussed industry operating experience regarding air
intrusion into component cooling water (CCW) systems, as well as other CCW system
performance issues including protection
from high energy line breaks (HELBs) and seismic events. The team reviewed the Seabrook PCCW system operating, fill and vent,
and alarm response procedures to verify that procedures adequately addressed the concerns identified in the
PCCW piping and head tanks; reviewed PCCW system corrective action
condition reports (CRs); reviewed the
- PCCW system design basis and modification history; and interviewed design engineers to independently verify that the
PCCW system was adequately designed to ensure protection from the design basis events postulated
in the IN. Finally, the team reviewed NextEra procedures developed to respond to a loss
of PCCW inventory event by reviewing operating procedures, interviewing operators,
and conducting a walkthrough of time-critical PCCW emergency makeup strategies to
determine if the procedures and actions were adequate to mitigate the postulated loss of inventory and were consistent with licensing basis documents.
b. Findings
Introduction: The team identified a finding of very low safety significance involving a non-cited violation of
III, "Design Control," because NextEra did not verify the design basis for the PCCW had been translated into
specifications and procedures. Specifically, NextEra incorrectly determined that leakage
past isolation valves separating the safety and non-safety portions of the system was
acceptable contrary to the system design which required the "B" PCCW system to be leak tight because a safety related refill capability was not available.
Description: The team reviewed several
PCCW system (CC-V-447 and 448). The team found that these
valves were credited as the isolation between the safety related and the non-safety
related portions of the PCCW system. The team determined that during the April 2011 and September 2012 refueling outages NextEra leak tested the isolation valves using approved work order instructions and adjusted the valves, as required, below the
leakage limits identified in the instructions. The team noted that in April 2011 the V447
leakage was 2.5 gpm and V448 leakage was 4.5 gpm. Corrective action was taken on
V448 to reduce leakage to 0.0 gpm and V447 was not adjusted. The team noted that in September 2012 the V447 leakage was 4.0 gpm and V448 leakage was 0.0 gpm - corrective actions were taken to reduce V447 leakage to zero.
The team reviewed the licensing basis of the system to determine if leakage past these valves was acceptable. The team reviewed
- NRC [[]]
NUREG 0800, Standard Review Plan
Section 9.2.2, and found it stated, "cooling water systems that are closed loop systems are reviewed to ensure that the surge tanks have sufficient capacity to accommodate
expected leakage from the system for 7 days or that a seismic source of makeup can be made available within a time frame consistent with the surge tank capacity." The team
then reviewed the
PCCW trains were equipped with a surge tank to provide makeup capability for the system and safety related makeup sources for the trains were not required because the trains were described in the licensee's submittal to be leak tight.
To address the differences between the current testing limits and the original design and
licensing basis, NextEra provided the team with several engineering evaluations that
were used as the basis to develop the acceptable system leak rates listed in the work order. The team reviewed engineering evaluation RES-93-500 which determined a total leakage of 2.5 gpm per PCCW loop was acceptable. The analysis was based on the low
level alarm and time available (~ 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) for operators to provide alternate makeup to the respective
- PCCW loop. The team noted that the evaluation stated, "the capability to connect to the makeup sources and deliver makeup to the
PCCW head tank should be
developed." The team then reviewed NextEra's response to NRC Information Notice
98-25, "Loss of Inventory from Safety-Related Closed-Loop Cooling Water Systems."
The team found the evaluation stated UFSAR change 99-043 credited the installed cross
connection from
FP booster pump as the seismic
makeup source for both PCCW trains and stated that, "this capability is considered to be
part of the current licensing and design basis for
CR 573970) which evaluated and approved the limited use of a 10 gpm leakrate as acceptance criteria in the work order instructions used to leak-check the
V447 and V448 valves.
To verify the adequacy of the refill lineup credited in the evaluations, the team completed
an alignment walkdown with the abnormal operating procedure used to refill the PCCW system with a NextEra equipment operator. The team determined the required procedure steps could be accomplished within 30 minutes, however, the team found that
the
header, and the lineup was dependent on the availability of the "A" train vital bus.
Therefore, the team concluded that the lineup could not be credited to refill the "B"
- PCCW header. The team also reviewed other refill lineups described in the abnormal operating procedure and determined - although there were a variety of sources that could be used to supply makeup water to
PCCW "B" header - none of the lineups were
seismically qualified and could not be used as a credited refill source to meet the design
requirements for the "B" PCCW train. The team further determined that none of the
engineering evaluations had evaluated the acceptability of "B" train leakage without a qualified refill source. Based on this and the allowed leakage limits for the "B" train isolation valves, the team determined, during certain design basis events with an
assumed single failure on the "A" train, a loss of the PCCW system could occur. Finally, the team concluded that the leakage identified in 2011 and 2012 could have resulted in
the loss of all PCCW within approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the design basis event.
Following identification, NextEra entered the issue into their corrective action program and evaluated the operability of the PCCW system. Based on the as-left condition of the
"B" train isolation valves in September 2012 of 0.0 gpm leakage, NextEra determined
that there was reasonable assurance of operability on the "B" train of PCCW. The team
reviewed the operability evaluation and determined that it was reasonable.
Analysis: The team identified that NextEra did not adequately translate the PCCW design basis into specifications, drawings, procedures, and instructions. Specifically, NextEra modified the isolation valve leakage limits but did not ensure that adequate
abnormal operating procedures existed to maintain the B train PCCW inventory following
a seismic event. The team determined that this performance deficiency was reasonably within NextEra's ability to foresee and correct and should have been prevented. The team determined that the finding was more than minor because it was similar to example
3.k of
- NRC [[]]
IMC 0612, Appendix E, "Examples of Minor Issues," in that design control measures for verifying the adequacy of design were not implemented. NextEra's change to the isolation valve leakage acceptance limits resulted in a condition where
there was now a reasonable doubt on the operability of the PCCW system. Specifically,
the allowed leakage invalidated NextEra's assumption that operators had sufficient time
to provide PCCW inventory make-up from a seismically qualified source following a
seismic event. Additionally, the finding was associated with the Protection Against
External Factors (seismic event) attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the reliability of systems that respond to
initiating events to prevent undesirable consequences (i.e., core damage).
The team evaluated the finding in accordance with
- IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," and screened this external event issue under Mitigating Systems. In accordance with
IMC 0609, Exhibit 2.B, the finding involved the loss or degradation of equipment or function
specifically designed to mitigate a seismic initiating event (the PCCW isolation valves
were designed to mitigate a seismic-induced failure of non-safety related system piping) and triggered the use of Exhibit 4, "External Events Screening Questions." The "yes" response to question 1 of Exhibit 4 directed a DRE because if the valve isolation function
was assumed to be completely failed following a seismic event, it would degrade one or
more trains of a system that supports a risk significant system or function. A Region I
DRE and determined that the finding was of very low safety
significance (Green). The
- CDF of 3.6E-7 with a dominant sequence of seismically induced loss of offsite power with a failure of the "A"
SEPS diesel generator set.
The
SPAR model, SAPHIRE 8, and the
following major assumptions.
The condition existed for an exposure period of 365 days. The combined leakage through the "B" train
- SEPS was available to supply the E5 vital bus (based on its seismic rigidity). No credit for operator recovery (providing makeup to the "B"
- PCCW loop or aligning alternate cooling to the B charging pump). No credit for the non-seismic demineralized water system and normal
PCCW.
In accordance with IMC 0609, Appendix H, "Containment Integrity Significance
Determination Process," the finding was evaluated for risk associated with LERF. Based
on Table 5.1, the issue was determined not to increase the LERF risk since it was not
associated with inter-system loss-of-coolant accidents (LOCAs) or steam generator tube
ruptures.
The team determined that the finding had a cross-cutting aspect in the area of Human
Performance, Resources, because NextEra did not ensure that personnel, equipment,
procedures, and other resources were available and adequate to assure nuclear safety.
Specifically, those necessary for complete, accurate, and up-to-date design documentation, procedures, and work packages in that engineering evaluations and
maintenance procedures associated with PCCW isolation valves did not align with the
design and licensing basis requirements for a leak tight system. [IMC 0310 Aspect
Enforcement
III, "Design Control," requires, states, in part, that measures shall be established to assure that applicable regulatory
requirements and the design basis are correctly translated into specifications, drawings,
procedures, and instructions. Contrary to the above, since April 2011, NextEra's design
control measures did not adequately translate the
- PCCW design basis into specifications, drawings, procedures, and instructions. Specifically, NextEra did not adequately evaluate the impact of isolation valve leakage on B train
PCCW inventory
during design basis events. Because this violation was of very low safety significance and has been entered into NextEra's corrective action program (CR 1865448), it is being
treated as a
(NCV 05000443/2013008-04, Primary Component Cooling Water System
Unavailable Following a Seismic Event)
.2.2.2 NRC Information Notice 2012-11: Age-Related Capacitor Degradation
a. Inspection Scope
The team inspected the review performed by the licensee on
- NRC [[]]
"Age-Related Capacitor Degradation." The IN described failure of safety related
equipment due to the failure of capacitors that had be installed for longer that the
manufactures qualified life. The team reviewed the licensee evaluation (AR 01792133) in order to evaluate the NextEra's response to the operating experience. The team found that NextEra concluded that age-related capacitor failures are addressed in the
Seabrook systems
PM strategy to ensure NExtEra was replacing the power supplies on an adequate PM frequency, was inspecting for
visible degradation, and trending to determine if ripple is increasing.
b. Findings
No findings were identified.
.2.2.3 NRC Information Notice 2009-03: Solid State Protection System Card Failure Results in Spurious Safety Injection Actuation and Reactor Trip
a. Inspection Scope
NRC Information Notice 2009-03 informed lic
ensees about recent operating experience regarding an SSPS failure that resulted in a spurious safety injection (Sl) actuation which could not be reset with the normal control room override switches. The team reviewed NextEra's evaluation and follow-up actions for the described failure modes and effects.
Specifically, the team evaluated NextEra's ability to reset a spurious Sl signal if the control room switches were ineffective, and the established
SSPS electronic cards.
b. Findinqs
No findings were identified.
.2.2.4 NRC Information Notice 2010-27: Ventilation System Preventive Maintenance and
Design Issues
a. Inspection Scope
The team reviewed Seabrook's evaluation of IN 2010-27, "Ventilation System Preventive Maintenance and Design Issues" and the associated corrective action report
(CR 1607275) in order to evaluate NextEra's response to the operating experience. The
IN to alert licensees of recently identified ventilation system preventive maintenance and design issues. The team reviewed NextEra's evaluation the potential impact of the identified issues to determine if the issues in the IN were directly applicable
to Seabrook. The team reviewed Seabrook's preventive maintenance program to
ensure concerns associated with the station's prior
PM activities associated with several
ventilation system components to evaluate whether the existing PM program was being adequately implemented. Finally, the team interviewed responsible engineers and walked down specific ventilation system components and controls to assess the
installation configuration, material condition, and potential vulnerability to hazards.
b. Findings
No findings were identified.
4.
- OTHER [[]]
ACTIVITIES
4OA2 Identification and Resolution of Problems (IP 71152)
a. Inspection Scope
The team reviewed a sample of problems that NextEra identified and entered into their
corrective action program. The team reviewed these issues to evaluate whether NextEra had an appropriate threshold for identifying issues and to evaluate the effectiveness of corrective actions. In addition, corrective action documents written on
issues identified during the inspection were reviewed to evaluate adequate problem
identification and incorporation of the problem into the corrective action program. The
corrective action documents that were sampled and reviewed by the team are listed in
the attachment.
b. Findings
No findings were identified.
4OA6 Meetings, including Exit
On April 26, 2013, the team presented the inspection results to Mr. Kevin Walsh, Site Vice President, and other members of the NextEra staff. The team verified that none of the information in this report is proprietary.
Attachment: Supplemental Information
Attachment
ATTACHMENT
- KEY [[]]
- POINTS [[]]
- OF [[]]
CONTACT
Licensee Personnel
K. Walsh Site Vice President J. Connolly Engineering Director
K. Douglas Maintenance Director
M. Collins Design Engineering Manager
M. Ossing Engineering Programs Manager M. O'Keefe Licensing Manager V. Brown Senior Nuclear Analyst
C. Thomas Mechanical Design Engineer
M. Woods Fire Protection Engineer
M. Lee Mechanical Design Engineer E. Mathews System Engineer R. Parry Engineering Supervisor
D. Yates Senior Engineer
J. Mizzau Electrical Design Engineering
V. Patel Electrical Design Engineer
P. Brangiel System Engineer K. Shea System Engineer B. Woodland Electrical Design Engineer
G. Sessler System Engineer
- R. Jamison Electrical Design Engineer W. McCallister Electrical Design Engineer K. Letourneau Electrical Design Engineer
- LIST [[]]
- OF [[]]
- CLOSED [[]]
AND DISCUSSED
Opened and Closed
05000443/2013008-01 NCV Failure to Verify Adequate Fault Protection for Safety Related Equipment from Non-Safety Related
Load Fault (1R21.2.1.3)05000443/2013008-02 NCV Condensate Storage Tank Water Level above Limits of Seismic Qualification (1R21.2.1.12)05000443/2013008-03 FIN Failure to Perform Preventative Maintenance on the
Supplemental Emergency Power System
(1R21.2.1.17)05000443/2013008-04 NCV Primary Component Cooling Water System Unavailable Following a Seismic Event
(1R21.2.2.1)
Attachment
- LIST [[]]
OF DOCUMENTS REVIEWED
Calculations
- 07-014, Seabrook Nuclear Power Plant Tank Vortexing Analysis, Revision B
4.3.07.27F, PCCW Instrument Setpoints, Revision 7
4.3.07.50F, PCCW System Hydraulic Transient Loads, Revision 0
4.3.07.52F,
SW System Steady State Analysis, Revision 10 4.3.7-1F, Primary Component Cooling Water (PCCW) Heat Loads and Flow Rates for Various
Plant Operating Modes, Revision 6
4.3.8.11F, Service Water and Cooling Tower Pumps - NPSH Available, Revision 2
4.3.8.40F, Service Water System Pressure Transient Evaluation, dated 1/9/86 6.01.61.03,
NPSH Available, Revision 2
737-37, Condensate System Control Setpoints, Revision 6
737-50, EFW Pump Startup, Revision 0
737-51, Inventory in CST, Revision 2
737-60,
- EFW Pump Recirculation Pressure Drop, Revision 0 760-11, Fuel Oil Storage System Capacity, Revision 0 760-13,
EDG Fuel Oil Transfer Pump NPSH, Revision 0
9763-5-SP-00-4-F, PCCW Backup Air Supply, Revision 0
9763-3-ED-00-01-F, Calculation of Short Circuit Currents, Revision 8
9763-3-ED-00-02-F, Voltage Regulation, Revision 13
9763-3-ED-00-14-F, Batteries, Chargers, and Motor Feeders, Revision 15 9763-3-ED-00-23-F, Medium Voltage Protective Relay Coordination and Miscellaneous Relay Settings, Revision 59763-3-ED-00-27-F, Unit Substation Load Study, Revision 10 9763-3-ED-00-28-F, Motor Control Circuit Protection, Revision 7
9763-3-ED-00-31-F, 480V Coordination, Revision 3
9763-3-ED-00-34-F, Uninterruptible Power Supplies Load, Class 1E, Revision 8 9763-3-ED-00-43-F,
ED-00-44-F, 125 VDC Breaker Coordination, Revision 2
9763-3-ED-00-46-F, Failure of Non-Class 1E Loads on Class 1E Buses, Revision 3
9763-3-ED-00-66-F, Control Circuit Voltage Drop, Revision 4
9763-3-ED-00-68-F, 120 VAC Breaker Coordination, Revision 2
9763-3-ED-00-83, Diesel Generator Loading, Revision
CN-TA-03-189, LONF/LOAC Analysis for Seabrook 7.4% Uprate, Revision 0
C-S-1-20801, EFW System Flow Study, Revision 1
C-S-1-20818,
EFW Pump Operation, Revision 0
C-S-1-20819, CST Required Volume for Hot Standby and Plant Cooldown at Uprate Power Level, Revision 0 C-S-1-23704, Allowable Leakage from Safety Related Air Supplies, Revision 3
C-S-1-25107, DG Control Air Usage / Air Compressor Capacity, Revision 3
Attachment
C-S-1-28009, Primary Component Cooling Water System Heat Loads and Flow Rates for Various Plant Operating Modes after SPU, Revision 0 C-S-1-28058, RHR Cooldown Cases, Revision 0
C-S-1-28092, CST Volume Calculation, Revision 0
C-S-1-57017,
- PCCW [[]]
HX Outlet Temperature Uncertainties, Revision 3
C-S-1-80903, FW-V-156 and FW-V-163 Differential Pressure Analysis, Revision 3
C-S-1-80903, Motor-Operated Valve Differential Pressure Calculations, Revision 1 C-S-1-80904,
IST Acceptance Criteria, Revision 9 C-S-1-83704, Hydraulic Modeling of PCCW Flow Distribution, Revision 3
C-S-1-84213, Appendix R Timing Calculations for Reactor Coolant Inventory Control, Revision 1 C-S-1-86901, Containment Pressure Following a
PCCW Heat
Exchanger, Revision 2 C-S-1-E-0161,
CW60, Design of Supports for Sluice-Gate Guide-Rail & Estimate of Materials, Revision 2
C-X-1-27801,
- SEPS Diesel Generators Minimum Fuel Requirements, Revision 0 C-X-1-50004, Air Accumulators Sizing (1-
PDM Design Calculation-Floating Cover, Revision 0
- MSVCS -FAG-11, Containment Enclosure Ventilation Area Post-LOCA Temperature Transient, Revision 0 MT-21, Missile Evaluation for CST, Revision 0 SBC-227,
SBC-565, Diesel Generator Fuel Oil Tank Vortexing Evaluation, Revision 0
Completed Surveillance and Modification Acceptance Testing
- 1-CBA-DP-133-M32-000, Train 'A' Switchgear
CC-P-2295-CAL-1, 1-CC-TV-2271-2 Air Accumulator Pressure Calibration, performed 6/4/12
1-SWA-DP-192-M32-000, Train 'B' Control Room SA Fire Damper Inspection, performed 4/6/09
1-SWA-DP-L1-000, Train 'B'
SWA-FN-40-A-E360-0722-000, Train 'A' Switchgear Supply Fan 1E Breaker Current Injection Testing, performed 6/13/05 1-SWA-FN-40-B-E360-0722-000, Train 'B' Switchgear Supply Fan 1E Breaker Current Injection Testing, performed 12/18/03 40158940-01, CC Train B Temperature Control Valve Inspection, performed 12/6/12
93EDSI000108 WO 93W001418, Westinghouse Inverter Special Test, performed 5/13/93
0333446,
- FW [[-V-163, dated 3/22/04 EX1808.013, Control Room Emergency Makeup Air and Filtration Subsystem 18-Month Surveillance, performed 7/28/11 and 8/16/11 EX1808.014, Containment Enclosure Emergency Exhaust Filter System 18-Month Surveillance, performed 9/17/12 IS1616.411, CC-T-2271]]
- PCCW [[]]
- PCCW [[]]
LP-B Supply Header Temperature Indication and Alarm Calibration, performed 3/5/12
Attachment
- PCCW [[]]
- PCCW [[]]
PCP-UV/UF Input Relay Time Response Test, performed 10/13/12
PM, performed 4/11/08 LS0556.08, 7.5KVA Westinghouse Inverter Routine PM, performed 4/25/11
LS0563.154, 1-CC-P-11-D Trip Checks, performed 9/10/10
TDPU Timing Relay (1-CC-P-11B), performed 4/24/11
LS0564.33, 480 Volt Static Motor Testing and Dynamic Motor Monitoring, performed 8/26/11 LX0556.04, "B" Battery Service Test, performed 5/08/10
- PCCW Train B Quarterly Operability, 18 Month Position Indication, and Comprehensive Pump Testing, performed 12/5/12 and 3/5/13
OX1416.01, Monthly Service Water Valve Verification, performed 2/11/13
OX1416.04, Service Water Quarterly Pump and Discharge Valve Test and Comprehensive Pump Test, performed 3/28/12 and 12/28/12 OX1426.01, Diesel Generator 1A Monthly Operability Surveillance, performed 2/11/13 OX1426.22, Emergency Diesel Generator 1A 24 Hour Load Test and Hot Restart Surveillance, performed 11/14/11 OX1426.26, Diesel Generator 1A Semiannual Fast Start Operability, performed 10/13/12
PR 13-135, E-42A EDG - Jacket Water Preliminary Report, dated 10/2/12
Seabrook Station U1
UCC Project No. 01-05290.20, Inspections and Cleaning Performed on Service Water Intake Structure, Circulating Water Intake Structure, Offshore Intakes and Discharge Structures Inspection Report, dated 10/10/12 V1195754, 1-CC-P-11D Pump Outboard Bearing Oil Analysis Report, dated 12/12/12
V1195755, 1-CC-P-11D Pump Inboard Bearing Oil Analysis Report, dated 12/12/12
V1195756, 1-CC-P-11D Motor Outboard Bearing Oil Analysis Report, dated 12/13/12
V1195761, 1-CC-P-11D Motor Inboard Bearing Oil Analysis Report, dated 12/12/12 V1200654, 1-SW-P-41DL Motor Lower Bearing Oil Analysis Report, dated 1/16/13 V1200655, 1-SW-P-41DU Motor Upper Bearing Oil Analysis Report, dated 1/16/13
V1202722, DG-1A TK-102A Rocker Arm Drain Tank Oil Analysis Report, dated 2/1/13
Attachment
V1202723,
DG-V354A LO Pump Discharge Header Oil Analysis Report, dated 3/16/13
V1210139, DG-V264A Main Bearings, LO Supply Header Oil Analysis Report, dated 3/16/13
V5008965, Diesel Fuel Oil Shipment Certificate of Analysis Report, dated 11/2/12
V5010138, Diesel Fuel Oil Shipment Certificate of Analysis Report, dated 2/15/13
Corrective Action Program Conditon Reports
- 0011036
0036556
0038184
0097802 0192624 0196869
0197779
200520
201099 0202081 0207103
207350
208861
210324
212979 0214109 0214560
214845
215124
219030 0219231 0220114
21390
0391457
04-00538
044937 05-00299 0567148
0570384
0573970
0574260 0578644 0587067
06-06070
07-01183
07-04812 07-15715 1607275
1609878
21344 1623270 1623371
24211
1644079
1644866 1667470 1671623
1684800
1693821
1694951
1719149 1741534 1749483
1755119
1769909
1792133 1797453 1804421
1808124
1808127
1808696
1813412 1813674 1819834
20627
1836696
1846854 1858210 1858321
1858566*
1858909
1859248 1859370 1859681
1859956
1859964* 1860344* 1860439*
1860505*
1860547*
1860549* 1860767* 1860925*
1860926*
1861161*
1861169*
1861712* 1861817* 1861821*
1861832*
1861866*
1861868* 1861902* 1862223*
1862680
1862733*
1862758*
1862895* 1862969 1863099
1863107*
1863477*
1863954 1864067* 1864162*
1864175*
1864211
1864431* 1864567 1864889*
1865060*
1865114* 1865153* 1865285
1865338*
1865448*
1865454* 1865455* 1865459*
1865462*
1865478*
1865485*
1865493* 1865495* 1865499*
1865544*
1865637
1865647 1866080 1866097
1866276
1866283
1866633*
1866676 1866810* 1866929
1867070*
1867518
1867521 1867947 *
1867956*
1867996*
1868055*
1868137* 1868156* 1868175*
1868196*
1868199* 1868355* 1868359*
1868402*
1868416*
1868447* 1868458* 1868510*
1868583*
1868587*
1868818*
1868959* 1869111* 1869150*
1869296*
1869364*
1869375* 1869376* 1869380*
1869381*
1870937*
1875462*
1877891* 40108199 40163481
96-1442
- AR [[]]
PMCR 66367
- NRC identified during this inspection.
Attachment
Drawings and Wiring Diagrams
- 310042 Sht. 1,
VDC 1-Line Diagram, Revision 3
310102 Sht. A7Aa, SEPS 4kV Incoming Feeder Diagram, Revision 2
310102 Sht. A7Ab, SEPS 4kV Close Circuit Schematic, Revision 1
310102 Sht. A7Ac, SEPS 4kV Trip Circuit Schematic, Revision 0
310102 Sht. A7Ad,
SEPS 4kv Auxiliary Contacts, Revision 0 310105 Sht. D26a/b, UPS I-1B 3-Line Diagram, Revision 6
310105 Sht. E02/10a, UPS I-1B Ground Detection Circuit, Revision 8
310105 Sht. E02a,
PP-1B Schedule, Revision 12
310231 Sht. 45a, Motor Load List
SEPS Cable Schematic, Revision 0
310970 Sht.
SEPS Control and Switchgear Schematics, Revision 7
CBA-B20302, Control Building Air Handling Emergency Switchgear Area, Revision 4 1-CC-B20204, Primary Component Cooling Loop 'A' Detail, Revision 4
1-CC-B20205, Primary Component Cooling Loop 'A' Detail, Revision 25
1-NHY-310844, Sht. C2Ra, Electrical Schematic, FW-V-163, Revision 10
1-NHY-504162, Logic Diagram, FW-V-163, Revision 12
1-NHY-506235,
NHY-507044, EFW Pump P-37B Control Loop Diagram, Revision 14 9763-F-101327, Condensate Storage Tank Concrete Plan, Revision 9
Foreign Print number 52320, CST Bottom and Anchor Bolt Layout, Revision J
Foreign Print number 52320, PDM Tank Delta Seal, Revision 2
1-CO-B20426, Condensate System PID, Revision 30
1-FW-B20688,
PID, Revision 20 1-CC-B20206, Primary Component Cooling Loop 'A' Detail, Revision 16
1-CC-B20207, Primary Component Cooling Loop 'A' Detail, Revision 12
1-CC-B20211, Primary Component Cooling Loop 'B' Detail, Revision 21
1-DAH-B20624, Diesel Generator Building Air Handling, Revision 7
1-MAH-B20495, Miscellaneous Air Handling
RHR Vaults Detail, Revision 12
1-NHY-503511, EAH - Containment Enclosure Cooler Fan Logic Diagram, Revision 7
1-NHY-503514, EAH - Containment Enclosure Return Air Dampers Logic Diagram, Revision 5
1-NHY-503515,
RH-B20662, Residual Heat Removal System, Train 'A' Detail, Revision 17
1-RH-B20663, Residual Heat Removal System, Train 'B' Cross Tie Detail, Revision 21
1-SW-B20795, Service Water System Nuclear Detail, Revision 40
1-SWA-B20372, Air Handling System for SW Pumphouse Cooling Tower, Revision 7
5618, Outline Drawing Vertical Residual Heat Exchanger, Revision 2 5620, Vertical Residual Heat Exchanger Assembly and Details, Revision 3 CBV-W5-10-0003, 1" Screwed End Check Valve, ANSI Class 150, Al-Bronze Construction, Revision B
Attachment
1-NHY-310103, Sht. 5N, 480V Unit Substation Bus 1-E52 125V
AC Aux Buses, Revision 6 1-NHY-310103, Sht. AC2, 480V Bus 1-E52 Incoming Line Three Line Diagram, Revision 4
1-NHY-310102, Sht. A63a, 4160V Feed to 480V Xfmr 1-EDE-X-5B Three Line Diagram, Revision 6 1-NHY-310013, 480V Unit Substation Buses E-51 & E-52 One Line Diagram, Revision 22
1-NHY-310008, 4160V Switchgear Bus 1-E6 One Line Diagram, Revision 18 1-NHY-310952, Sht. AF5a, Containment Encl Cooler Fan 1-FN-5A Three Line Diagram, Revision 4 1-NHY-310952, Sht. AF5b, Containment Encl Cooler Close Circuit Fan 1-FN-5A Schematic
Diagram, Revision 10 1-NHY-310952, Sht.
NHY-301107, Shts. AR4a, AR4b, AR4f, and AR4g, Service Water Pump 1-P-41D Three Line Diagram, Revisions 4, 12, 11 and 9 1-NHY-310844, Shts. C3Ta and C3c, Emerg Feedwater Recirc Valve V-347, Revisions 6 and 7
1-CC-B20211, Primary Component Cooling Loop "B" Detail P&ID, Revision 21 1-DG-B20459, Diesel Generator Fuel Oil System Train "A" Detail, Revision 16 1-DG-B20460, Diesel Generator Starting Air System Train "A" Detail, Revision 25
1-DG-B20461, Diesel Generator Cooling Water System Train "A" Detail, Revision 22
1-NHY-310895 Sht. A79b, PCCW Loop B Pump 1-P-11D Close Schematic, Revision 11
1-NHY-BD-2033, Main Steam Feedwater Pipe Enclosure - East, Revision 5
1-SW-B20794, Service Water System Nuclear Detail, Revision 35 1-SW-B20795, Service Water System Nuclear Detail, Revision 40 1-SW-B20796, Service Water System Nuclear Detail, Revision 5
9763-F-202382, Emergency Feed Pump Building Sleeves Piping, Revision 14
2332, EF-26-EFSTI-1005 Penetration Seal Design, Revision 0
Engineering Evaluations
- 219030-04,
- DGA Tripped on High Lube Oil Temperature during Monthly Surveillance Apparent Cause Evaluation, dated 5/6/10
- SW [[]]
- DG Supply/Return Line Inspections Prompt Operability Determination, dated 10/19/12 C-S-1-38016, Supplemental Emergency Power System Voltage Regulation, Revision 0
- EC -145220, Service Water Pump Key Material Substitution, Revision 1 EC-145282, Reconciliation of the Impacts of
- DG [[]]
- DG [[]]
EC-145296, Service Water Pump Shaft Material Substitution, Revision 1
EC-145348, Alternate Termination Method for SW-P-41 Motors, Revision 0
- ASCO [[]]
SOV Part Substitution, Revision 0 EC-272116, Review of Air Relay and Signal Diaphragm Elastomers Used in PCCW Temperature Control Valve Positioners, Revision 1
Attachment
- PCCW [[]]
TCV's Rendered Inoperable Due to Inadequate Capacity of High Pressure Backup Gas Supply, Revision C ECA-98/806118, Will Equipment Damage Result from the Slamming Shut of PCCW Pump
Discharge Check Valve, Revision A
EE 90-50, Internal Flooding Potential through Plant Drain and Sump Systems, dated 11/30/90
EE-04-024, Operator Action Response Times Assumed in the UFSAR, Revision 4
EE-93-22, Diesel Generator Starting Air System Operability Requirements, Revision 1 EC-145290, - Reconciliation of Vortex Issue for the CST, Revision 0 EE-SS-EV-98006, Attachment 1, CC-P-11B/D Minimum Performance, Revision 9
File No. 173-5-1M, Masoneilan Butterfly Valve Mechanical Equipment Qualification, Revision 2
PCCW Heat Removal Capacity, Revision 0
SBC-128, Technical Specifications - Setpoints and Allowable Values, Revision 15
SEA-11-0051, PCCW Venting Capability Enhancement, dated 5/10/10
Licensing and Design Basis Documentation
DBD-CC-01, Primary Component Cooling Water System Design Basis Document, Revision 4
DBD-DG-01, Emergency Diesel Generator - Mechanical Design Basis Document, Revision 4 DBD-EAH-01, Containment Enclosure Cooling and Exhaust Filter Systems Design Basis
Document, Revision
- DBD Emergency Feedwater System Design Basis Document, Revision 6 DBD-SW-01, Service Water System Design Basis Document, Revision 6]]
DCN # 21 and 23
- USNRC , Requests for Additional Information Related to Instrumentation and Control Design, dated 5/10/86 Technical Requirement Manual
SEPS, Revision 99
Technical Specifications 3/4.8.1 and Basis,
UFSAR Change Request 05-015, 03DCR012 DCN 05 (Power Up-rate), Revision 0
Updated Final Safety Analysis Report, Revision 14
Miscellaneous Documents
- 02259913, Diesel Fuel Oil Shipment Certificate of Analysis, dated 11/2/12
40115435-02, SW-1811-02 VT-2 Visual Examination Report, performed 10/15/11
Attachment
9763-006-128-1, Specification Motor Data - Service Water Pumps, Revision 3
PCCW Bearing Temperature Trend Data, dated 2/3/13 to 3/7/13
- DG Cooling Water Heat Exchanger E-42-A Fouling Factors, dated 10/24/96 to 11/13/12 Diesel Generator System Walkdown Report, completed 7/12/12 and 10/12/12
EDG Roving Logs, dated 3/9/13 to 3/12/13
Fiche 52790,
- SSPS [[Maintenance Field Service Report, dated October 2012 Ingersoll-Dresser Certified Performance Pump Curve for Service Water Pump S/N 0300-004, dated 12/18/00 Ingersoll-Rand Characteristic Pump Curve for Component Cooling Water Pump No. 117528, dated 4/23/80 L0134J, Align Alternate (Firewater) Cooling to]]
CCP Lube Oil Cooler Job Performance Measure, Revision 3 L0135J, Align Alternate (Demin Water) Cooling to CCP Lube Oil Cooler Job Performance
Measure, Revision 3 L0138J,
- MOV Grouping, Selection, and Exclusion Criteria for Differential Pressure Testing, dated 3/07/92 Letter
NYN-97058, Seabrook to NRC, Response to Generic Letter 96-01, dated 5/22/97
Maintenance Rule Basis Document, EDE-04, dated 3/28/13
Maintenance Rule Basis Document,
- NRC Information Notice 2009-03, Solid State Protection System Card Failure Results in Spurious Safety Injection Actuation and Reactor Trip, dated 3/11/09
- NRC Safety Evaluation Report (SER), Amendment No. 97 to Seabrook Operating License Regarding Change to Emergency Power Systems (ML042240471), dated 9/21/04 Operator Aid #97-001,
- PCCW Temperature Control Manual Operation, dated 10/11/12 Primary Component Cooling Water System Walkdown Report, dated 7/18/12 & 11/5/12 Primary Operator Rounds, dated 3/25/13 to 3/29/13
RPT# 09-022, Seabrook Nuclear Assurance Report, dated 5/11/09
Attachment
- SBK [[]]
- LOIT [[]]
SRO Instant Candidate NSO Critical and Infrequent Task Guide, Revision 0 Seabrook Plant Trip Data: Loss of Feedwater Pump, dated 10/6/11
Service Water System Walkdown Report, dated 6/7/12 & 2/6/13
CST High Level Alarm Discrepancy, dated 4/15/13
Spec. Number 9763-006-246-6, Spec. for Safety Class 3 Field Fabricated Tanks, Revision 10
SW-P41D IST Pump Data Log, dated 6/21/06 to 12/28/12
Procedures
- CX0901.22, Diesel Generator Fuel Oil Tank Surveillance, Revision 20
PRA Risk Ranking and Performance Criteria Based on SSPSS-2009, Revision 1
ER-AA-201-2001, System and Program Health Reporting, Revision 6 ES1804.055, Inservice Testing Pump and Valve Program, Revision 6
PM, Revision 1
VDC K-Line Breakers, Revision 16
PM, Revision 12
MS0599.20, Fire Barrier Wrap Installation and Repair, Revision 2
NADC, Design Control Manual, Revision 59
OX1400.02, Remote Safe Shutdown System 18 Month Operability Check, Revision 9
PEG-208, Service Water System Performance Monitoring, Revision 5 PEG-94, Service Water Inspection and Repair Trending Plant Engineering Guide, Revision 6
SITR, Inservice Test Reference, Revision 23
Procedures (Operating)
- Alarm
- ID [[]]
RHR Vault-2 Stairwell Temp Hi, Revision 2 B5811, Alarm Response, Elect Tunnel Walkway Temp Hi, Revision 1
B6893, Alarm Response, Containment Enclosure Emer Exh Flow Hi-Hi, Revision 2
B7627, Alarm Response, Containment EnclosureF-9 Inlet High, Revision 3
D4525, Alarm Response, Containment Enclosure Sup Fan 5A Disch Press Lo, Revision 3 D4532, Alarm Response, Containment Enclosure Sup Fan 5B Disch Press Lo, Revision 2 D5502,
- SW [[]]
- DP [[]]
HI, Revision 7
D7416, Alarm Response, Containment Enclosure Cooler FN5A SS in Local, Revision 2
D7418, Alarm Response, Containment Enclosure Return FN31A SS in Local, Revision 1
D7783, Alarm Response, Containment Enclosure Emer Filter 9 Temp High, Revision 2
D7784, Alarm Response, Containment Encl Emer Filter 69 Temp High, Revision 2 D7785, Alarm Response, Containment Spray Pump A Room Temp High, Revision 2 D7786, Alarm Response, Safety Injection Pump A Room Temp High, Revision 2
D7789, Alarm Response, Centrifugal Charging Pump A Room Temp High, Revision 2
Attachment
D7794, Alarm Response, Containment Encl Return Fan Suction Pres Hi/Lo, Revision 1
DG-CP-75A UA-9558, Panel DG-CP-75 UA-9558 Local Alarm Response, Revision 54 E-0, Reactor Trip or Safety Injection, Revision 49
E-1, Loss of Reactor or Secondary Coolant, Revision 41
SI Required, Revision 31
- RHR [[]]
- RHR [[]]
HX B PCCW Flow Low, Revision 3
F4430, Alarm Response, Charging Pump Area Exh Vent Flow Lo, Revision 2
F4517, Alarm Response, Charging Pump Area Exh Both Fans Running, Revision 2
F4535, Alarm Response, Containment Encl Supply Fans None Running, Revision 2 F6894, Alarm Response, Containment Encl Emer Exh Flow High, Revision 0 F7526, Alarm Response, Containment Encl Emer Exh Flow Low, Revision 1
F7765, Alarm Response, Containment Encl/Outside Atmosph D/P Low, Revision 3
F7768, Alarm Response, Containment Encl Cool FN5A Auto Start, Revision 3
F7769, Alarm Response, Containment Encl Cool FN5B Trip, Revision 3
F7770, Alarm Response, Containment Encl Cool FN5A Trip, Revision 3
F7799, Alarm Response, Containment Encl Sup Fans Both Running, Revision 3
F7915,
- SW [[]]
PMP A Trip Alarm Response Procedure, Revision 5
F7956,
- SW [[]]
- PMP [[]]
- DISCH [[]]
FR-C.1, Response to Inadequate Core Cooling, Revision 26
FR-I.2, Response to Low Pressurizer Level, Revision 21 MM-UA-54, E-3 480V Bus 51/52/53 Volts Lo, Revision 5 MM-UA-55, E-4 4160V Bus 6 Volts Lo, Revision 6
ON1029.01, Water Treatment System Operation, Revision 20
OS1002.02, Operation of Letdown, Charging and Seal Injection, Revision 37
OS1012.04, Primary Component Cooling Water Loop B Operation, Revision 22 OS1013.03, Residual Heat Removal Train 'A' Startup and Operation, Revision 26 OS1013.06, Residual Heat Removal Train 'B' Shutdown, Revision 11
OS1016.04, Service Water Train B Operation, Revision 16
OS1023.66, Containment Enclosure Ventilation System Operation, Revision 17
OS1036.04, Emergency Feed Water Pump B Operation, Revision 2
OS1046.07, Vital 480V Operation, Revision 18 OS1090.01, Manual Operation of Remote Operated Valves, Revision 14
OS1216.01, Degraded Ultimate Heat Sink, Revision 22
OS1246.01, Loss of Offsite Power - Plant Shutdown, Revision 20
OS1412.09, PCCW Monthly Flow Check, Revision 8
Attachment
- PCCW Train B Quarterly Operability, 18 Month Position Indication, and Comprehensive Pump Testing, Revision 20
OX1413.06, RHR/RC Suction Valve 18 Month Interl
ock Verification Surveillance, Revision 2 OX1416.04, Service Water Quarterly Pump and Discharge Valve Test and Comprehensive
Pump Test, Revision
- 16 OX 1426.22, Emergency Diesel Generator 1A 24 Hour Load Test and Hot Restart Surveillance, Revision 14 OX1456.81, Operability Testing of
OX1461.05, SEPS Annual Availability Surveillance, Revision 5 Tech Spec Logs, Mode 1, Revision 138
System Health Reports
- System Health Report, Containment Air Handling, 1
st Quarter 2013 System Health Report, Control Building Air Handling, 1
st Quarter 2013 System Health Report, Diesel Generator Building Air Handling, 1
st Quarter 2013 System Health Report, Diesel Generator, 2
nd and 3 rd Quarter 2012 System Health Report, ED/EDE 4.16/13.8KV, 3
rd Quarter 2012 System Health Report, Emergency Feedwater System, 4
th Quarter 2012 System Health Report, Enclosure Air Handling, 1
st Quarter 2013 System Health Report, Primary Component Cooling Water, 3
rd and 4 th Quarter 2012 System Health Report, Residual Heat Removal System, 3
rd Quarter 2012 System Health Report, SEPS, 4
th Quarter 2012 System Health Report, Service
Water Air Handling System, 1
st Quarter 2013 System Health Report, Service Water, 3
rd Quarter 2012 System Health Report, SSPS, 3
rd Quarter 2012, 4
th Quarter 2012, and 1
st Quarter 2013 System Health Report, Switchyard, 3
rd Quarter 2012
Vendor Manuals
- 9763-006-173-5, Specification for Nuclear Control Valves, Revision 5 9763-248-64, Specification for Gates for Transition Structures, Revision
IM-SNH-66, Instruction Manual for Cooling Units, Revision A
Joseph Oat Corp Heat Exchanger Data Sheet (RHR), Revision 3
Attachment
NPRDS No. C470-1, Emergency Diesel Generat
or System Operation & Maintenance Manual, Revision A
CC-TV-2171/2271-1&2 Butterfly Valve Instructions, Revision 1
S390-3, Installing, Operating and Maintenance Instructions for 42 Inch Service Water Supply
Sluice Gates, Revision 1
Work Orders
- 91W001967
00306530
00406376 00425458 00535979
00627029
00717526
00809308 00811594 01172379
01180029
01186574
01186576
01186682 01186892 01192254
01196392
01197141
200214 01202156 01202157
202159 01202856 01203957
203958
204036
204037 01204043 01204045
204793
204797
207831
208824 01382185 01382193
01384552
01384583
40036698 40036699 40039191
40039192 40043813 40047604
40047605
40050111
40059793 40060109 40074171
40074437
40075419
40075420
40081950 40081951 40085033
40086682
40089233
40090937 40090975 40091311
40091545 40092749 40103603
40103638
40106185
40106852 40107063 40107064
40109115
40109291
40109638
40109660 40115435 40117552
40117553
40121892
40123861 40125816 40158940
40160188 40163184 40163185
40163186
40165907
40165909 40176797 40188955
40189204
40193672
210889
215453 94006130
- LIST [[]]
OF ACRONYMS
A Amps
FP Fire Protection
GL Generic Letter
Attachment
- IN [[[NRC] Information Notice]]
IST In-service Testing
kV Kilovolts
NCV Non-cited Violation
NextEra NextEra Energy Seabrook,
- LCC [[]]
UFSAR Updated Final Safety Analysis Report Vac Volts, Alternating Current
Vdc Volts, Direct Current