ML14212A458
| ML14212A458 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 08/05/2014 |
| From: | Glenn Dentel Reactor Projects Branch 3 |
| To: | Dean Curtland NextEra Energy Seabrook |
| DENTEL, GT | |
| References | |
| IR-14-003 | |
| Download: ML14212A458 (62) | |
See also: IR 05000443/2014003
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION I
2100 RENAISSANCE BLVD., SUITE 100
KING OF PRUSSIA, PA 19406-2713
August 5, 2014
Mr. Dean Curtland
Site Vice President
Seabrook Nuclear Power Plant
NextEra Energy Seabrook, LLC
c/o Mr. Michael Ossing
P.O. Box 300
Seabrook, NH 03874
SUBJECT: SEABROOK STATION, UNIT NO. 1 - NRC INTEGRATED INSPECTION
REPORT 05000443/2014003
Dear Mr. Curtland:
On June 30, 2014, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at
Seabrook Station, Unit No. 1. The enclosed inspection report documents the inspection results,
which were discussed on July 10, 2014, with you and other members of your staff.
The inspection examined activities conducted under your license as they relate to safety and
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
NRC inspectors documented two findings of very low safety significance (Green) in this report.
These findings did not involve a violation of NRC requirements. Further, inspectors
documented a licensee-identified violation, which was determined to be of very low safety
significance, in this report. The NRC is treating the finding as a non-cited violation (NCV),
consistent with Section 2.3.2.a of the NRC Enforcement Policy. If you contest the subject or
severity of any NCV 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 Nuclear Regulatory Commission,
ATTN.: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional
Administrator, Region I; the Director, Office of Enforcement, United States Nuclear Regulatory
Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at Seabrook
Station. In addition, if you disagree with the cross-cutting aspect assigned to the findings 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.
D. Curtland 2
In accordance with Title 10 of the Code of Federal Regulations (CFR) 2.390 of the NRCs Rules
of Practice, a copy of this letter, its enclosure, and your response (if any) will be available
electronically for public inspection in the NRCs Public Document Room or from the Publicly
Available Records component of the NRCs Agencywide Documents Access Management
System (ADAMS). ADAMS is accessible from the NRC website at http://www.nrc.gov/reading-
rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Glenn T. Dentel, Chief
Reactor Projects Branch 3
Division of Reactor Projects
Docket No. 50-443
License No: NPF-86
Enclosure: Inspection Report No. 05000443/2014003
w/ Attachment: Supplemental Information
cc w/encl: Distribution via ListServ
Non-Sensitive Publicly Available
SUNSI Review
Sensitive Non-Publicly Available
OFFICE RI/DRP RI/DRP RI/DRP
NAME PCataldo/ MSD per telecon MDraxton/MSD GDentel/ GTD
DATE 07/31/14 07/31/14 07/31 /14
1
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No.: 50-443
License No.: NPF-86
Report No.: 05000443/2014003
Licensee: NextEra Energy Seabrook, LLC
Facility: Seabrook Station, Unit No.1
Location: Seabrook, New Hampshire 03874
Dates: April 1, 2014 through June 30, 2014
Inspectors: P. Cataldo, Senior Resident Inspector
C. Newport, Resident Inspector
T. OHara, Reactor Inspector
B. Dionne, Health Physicist
Approved by: Glenn T. Dentel, Chief
Reactor Projects Branch 3
Division of Reactor Projects
Enclosure
2
TABLE OF CONTENTS
SUMMARY ................................................................................................................................ 3
REPORT DETAILS .................................................................................................................... 5
1. REACTOR SAFETY ........................................................................................................... 5
1R01 Adverse Weather Protection .................................................................................... 5
1R04 Equipment Alignment ............................................................................................... 6
1R05 Fire Protection .......................................................................................................... 7
1R06 Flood Protection Measures ...................................................................................... 8
1R07 Heat Sink Performance ............................................................................................ 9
1R08 In-service Inspection ................................................................................................ 9
1R11 Licensed Operator Requalification Program ...........................................................14
1R12 Maintenance Effectiveness .....................................................................................15
1R13 Maintenance Risk Assessments and Emergent Work Control ................................17
1R15 Operability Determinations and Functionality Assessments ....................................18
1R18 Plant Modifications .................................................................................................18
1R19 Post-Maintenance Testing .......................................................................................19
1R20 Refueling and Other Outage Activities ....................................................................19
1R22 Surveillance Testing ...............................................................................................20
2. RADIATION SAFETY.........................................................................................................21
2RS1 Radiological Hazard Assessment and Exposure Controls ......................................21
2RS2 Occupational ALARA Planning and Controls ...........................................................23
2RS3 In-Plant Airborne Radioactivity Control and Mitigation ............................................24
2RS4 Occupational Dose Assessment .............................................................................25
2RS5 Radiation Monitoring Instrumentation .....................................................................28
4. OTHER ACTIVITIES ..........................................................................................................31
4OA1 Performance Indicator Verification ..........................................................................31
4OA2 Problem Identification and Resolution ....................................................................31
4OA3 Follow-Up of Events and Notices of Enforcement Discretion ..................................33
4OA6 Meetings, Including Exit ...........................................................................................36
4OA7 Licensee-Identified Violation ....................................................................................36
ATTACHMENT: SUPPLEMENTARY INFORMATION...............................................................37
SUPPLEMENTARY INFORMATION....................................................................................... A-1
KEY POINTS OF CONTACT .................................................................................................. A-1
LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED .................................... A-1
LIST OF DOCUMENTS REVIEWED....................................................................................... A-2
LIST OF ACRONYMS ........................................................................................................... A-21
Enclosure
3
SUMMARY
IR 05000443/2014003; 04/01/2014-06/30/2014; Seabrook Station, Unit No. 1; Maintenance
Effectiveness and Follow-up of Events and Notices of Enforcement Discretion.
This report covered a three-month period of inspection by resident inspectors and announced
inspections performed by regional inspectors. Inspectors identified two findings of very low
safety significance (Green). The significance of most findings is indicated by their color (i.e.,
greater than Green, or Green, White, Yellow, Red) and determined using Inspection Manual
Chapter (IMC) 0609, Significance Determination Process, dated June 2, 2011. Cross-cutting
aspects are determined using IMC 0310, Components Within Cross-Cutting Areas, dated
December 19, 2013. All violations of NRC requirements are dispositioned in accordance with
the NRCs Enforcement Policy, dated July 9, 2013. The NRCs program for overseeing the safe
operation of commercial nuclear power reactors is described in NUREG-1649, Reactor
Oversight Process, Revision 5.
Cornerstone: Mitigating Systems
Green. The inspectors identified a finding of very low safety significance (Green) because
NextEra did not perform adequate evaluations of safety-related residual heat removal
(RHR) vaults. Specifically, additional technical evaluation and analysis was not adequately
conducted on the safety-related A and B RHR concrete vaults when it was determined
that they exceeded the quantitative limits specified in NextEra procedures. NextEra entered
the failure to perform adequate technical evaluations on concrete structures exceeding
American Concrete Institute (ACI) Tier II quantitative requirements into the CAP (action
request (AR) 01929460), and planned to perform a formal technical evaluation of the A
and B RHR vault conditions in accordance with their structural monitoring program
procedure and the ACI 349.3R-96 standard.
The performance deficiency was considered to be more than minor because it affected the
protection against external factors attribute of the Mitigating Systems cornerstone and its
objective to ensure the availability, reliability, and capability of systems that respond to
initiating events to prevent undesirable consequences. Specifically, the inspectors
concluded that the reliability of the structures was affected in that they exceeded the
specified Tier II limits without the performance of further technical evaluations. The issue
was evaluated in accordance with IMC 0609, Appendix A, The Significance Determination
Process for Findings At-Power, and determined to be of very low safety significance
(Green) because it did not represent an actual loss of function of at least a single train for
greater than its Technical Specification Allowed Outage Time or two separate safety
systems out-of-service for greater than it Technical Specification Allowed Outage Time.
This finding is related to the cross-cutting area of Human Performance - Procedure
Adherence, because NextEra did not follow processes, procedures, and work instructions.
Specifically, NextEra personnel did not perform an adequate technical evaluation of two
safety-related concrete structures that exceeded the quantitative criteria requiring such an
evaluation [H.8]. (Section 1R12)
Green. The inspectors identified a self-revealing finding of very low safety significance
(Green), because NextEra did not ensure that adequate procedural guidance existed in
ON1046.12, Operation of the Main Generator Breaker to limit the likelihood of events that
upset plant stability. Specifically, Seabrook station experienced an automatic reactor trip
from approximately 15 percent reactor power on April 1, 2014 when two of four reactor
Enclosure
4
coolant pumps (RCPs) tripped on low bus voltage. The cause of the reactor trip was
directly attributable to the main generator breaker inadvertently closing and actuating the
main generator multi-function protective relay. NextEra entered the event into their CAP,
and conducted a root cause evaluation to determine the root and contributing causes,
extent of condition and extent of cause, and to identify corrective actions to prevent
recurrence. NextEra initiated actions to revise ON1046.12 to add controls regarding the
potential risk associated with placing the main generator breaker control in local, conducted
briefings with Maintenance groups involved in the event, and evaluated the adequacy of
other Operations procedures that place equipment in a configuration where protective
features are bypassed or defeated.
The performance deficiency was more than minor because it was associated with the
procedure quality attribute of the Initiating Events cornerstone, and it adversely affected the
cornerstone objective to limit the likelihood of events that upset plant stability and challenge
critical safety functions during shutdown as well as power operations. The finding was
evaluated under IMC 0609, Attachment 4, Phase 1 - Initial Characterization of Findings.
The inspectors determined that the finding is of very low safety significance (Green)
because it did not result in both a reactor trip and the loss of mitigating equipment relied
upon to transition the plant from the onset of the trip to a stable shutdown condition. The
finding has a cross-cutting aspect in the area of Human Performance - Work Management,
because NextEra did not ensure that a process of planning, controlling, and executing work
activities such that nuclear safety is the overriding priority was implemented. Specifically,
ON1046.12, Operation of the Main Generator Breaker did not contain adequate
procedural guidance regarding the impacts of positioning the Main Generator Selector
Switch to local, take mitigating actions, and minimize time spent at increased risk
configurations [H.5]. (Section 4OA3)
Other Findings
A violation of very low safety significance that was identified by NextEra was reviewed by
the inspectors. Corrective actions taken or planned by NextEra have been entered into
NextEras corrective action program. This violation and corrective action tracking number
are listed in Section 4OA7 of this report.
Enclosure
5
REPORT DETAILS
Summary of Plant Status
Seabrook Unit 1 began the assessment period reducing power to begin a planned refueling
outage (OR) 16. However, at 12:26 a.m., on April 1, 2014, an unanticipated reactor trip
occurred while the plant was at approximately 15 percent reactor power. Following the reactor
trip, Unit 1 remained shutdown until April 24, 2014 when the main unit generator was
successfully synchronized with offsite power, and resumed full power operation (100 percent)
on April 28, 2014. Seabrook Unit 1 operated at essentially 100 percent power for the remainder
of the assessment period. Documents reviewed for each section of this inspection report are
listed in the Attachment.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection (71111.01 - 3 samples)
.1 Readiness for Seasonal Extreme Weather Conditions
a. Inspection Scope
The inspectors performed a review of NextEras readiness for the onset of seasonal high
temperatures. The review focused on the service water cooling tower, service water
pump house, switchyard, termination yard, and the control building. The inspectors
reviewed the Updated Final Safety Analysis Report (UFSAR), technical specifications
(TSs), the seasonal readiness memorandum, and the corrective action program to
determine specific temperatures or other seasonal weather that could challenge these
systems, and to ensure NextEra personnel had adequately prepared for these
challenges. The inspectors reviewed station procedures, including NextEras seasonal
weather preparation procedure and applicable operating procedures. The inspectors
performed walkdowns of the selected systems to ensure station personnel identified
issues that could challenge the operability of the systems during hot weather conditions.
b. Findings
No findings were identified.
.2 Summer Readiness of Offsite and Alternate Alternating Current (AC) Power Systems
a. Inspection Scope
The inspectors performed a review of plant features and procedures for the operation
and continued availability of the offsite and alternate AC power system to evaluate
readiness of the systems prior to seasonal high grid loading. The inspectors reviewed
NextEras procedures affecting these areas and the communication protocols between
the transmission system operator and NextEra. This review focused on changes to the
established program and material condition of the offsite and alternate AC power
equipment. The inspectors assessed whether NextEra established and implemented
appropriate procedures and protocols to monitor and maintain availability and reliability
Enclosure
6
of both the offsite AC power system and the onsite alternate AC power system. The
inspectors evaluated the material condition of the associated equipment by interviewing
the responsible system manager, reviewing condition reports (CRs) and open work
orders, and walking down portions of the offsite and AC power systems including the
345 kilovolt (kV) termination yard, the 345 kV switchyard, and the relay room.
b. Findings
No findings were identified.
.3 External Flooding
a. Inspection Scope
During the period of June 16 to June 20, 2014, the inspectors performed an inspection of
the external flood protection measures for Seabrook Station. The inspectors reviewed
the Updated Final Safety Analysis Report (UFSAR), Chapter 2.4.2.2, which depicts the
design flood levels and protection areas containing safety-related equipment to identify
areas that may be affected by external flooding. The inspectors conducted a general
site walkdown of the outside area of the site, fuel storage building, the control building,
and the emergency diesel generator building to ensure that NextEra erected flood
protection measures in accordance with design specifications. The inspectors also
reviewed operating procedures for mitigating external flooding during severe weather to
determine if NextEra planned or established adequate measures to protect against
external flooding events.
b. Findings
No findings were identified.
1R04 Equipment Alignment
.1 Partial System Walkdowns (71111.04Q - 4 samples)
a. Inspection Scope
The inspectors performed partial walkdowns of the following systems:
Containment penetration return to service on April 17, 2014
'A' residual heal removal (RHR) return to service on April 23, 2014
B' emergency feedwater (EFW) pump return to service on June 19, 2014
'B' service water return to service on June 26, 2014
The inspectors selected these systems based on their risk-significance relative to the
reactor safety cornerstones at the time they were inspected. The inspectors reviewed
applicable operating procedures, system diagrams, the UFSAR, TSs, work orders
(WOs), CRs, and the impact of ongoing work activities on redundant trains of equipment
in order to identify conditions that could have impacted system performance of their
intended safety functions. The inspectors also performed field walkdowns of accessible
portions of the systems to verify system components and support equipment were
Enclosure
7
aligned correctly and were operable. The inspectors examined the material condition of
the components and observed operating parameters of equipment to verify that there
were no deficiencies. The inspectors also reviewed whether NextEra staff had properly
identified equipment issues and entered them into the corrective action program (CAP)
for resolution with the appropriate significance characterization.
b. Findings
No findings were identified.
.2 Full System Walkdown (71111.04S - 1 sample)
a. Inspection Scope
On April 17, 2014, the inspectors performed a complete system walkdown of accessible
portions of the service water system to verify the existing equipment lineup was correct.
The inspectors reviewed operating procedures, surveillance tests, drawings, equipment
line-up check-off lists, and the UFSAR to verify the system was aligned to perform its
required safety functions. The inspectors also reviewed electrical power availability,
component lubrication and equipment cooling, hanger and support functionality, and
operability of support systems. The inspectors performed field walkdowns of accessible
portions of the systems to verify system components and support equipment were
aligned correctly and operable. The inspectors examined the material condition of the
components and observed operating parameters of equipment to verify that there were
no deficiencies. Additionally, the inspectors reviewed a sample of related CRs and work
orders to ensure NextEra appropriately evaluated and resolved any deficiencies.
b. Findings
No findings were identified.
1R05 Fire Protection
.1 Resident Inspector Quarterly Walkdowns (71111.05Q - 5 samples)
a. Inspection Scope
The inspectors conducted tours of the areas listed below to assess the material
condition and operational status of fire protection features. The inspectors verified that
NextEra controlled combustible materials and ignition sources in accordance with
administrative procedures. The inspectors verified that fire protection and suppression
equipment was available for use as specified in the area pre-fire plan, and passive fire
barriers were maintained in good material condition. The inspectors also verified that
station personnel implemented compensatory measures for out of service (OOS),
degraded, or inoperable fire protection equipment, as applicable, in accordance with
procedures.
Containment Building C-F-1-Z, C-F-2-Z, & C-F-3-Z on April 7, 2014
Emergency Feedwater Pump House 27'-0" EFP-F-1-A on May 30, 2014
Primary Auxiliary Building 53' & 81' PAB-F-3A-Z, PAB-F-3B-Z, PAB-F-4-Z on June 2,
2014
Enclosure
8
Site Plan/Hydrant Locations PLT-F-1-0 on June 16, 2014
Containment Enclosure Ventilation Area, CE-F-1-A on June 23, 2014
.2 Fire Protection - Fire Brigade Response on April 16, 2014 (71111.05A - 1 sample)
a. Inspection Scope
The inspectors observed and evaluated control room operator and fire brigade response
to fire alarms within containment on April 16, 2014, that involved flames, sparks, and
smoke associated with the terminal box for the A reactor coolant pump. The inspectors
verified that NextEra personnel identified deficiencies and took appropriate corrective
actions as required. The inspectors evaluated specific attributes as follows:
Proper wearing of turnout gear and self-contained breathing apparatus
Sufficient fire-fighting equipment brought to the scene
Effectiveness of command and control
Communications established between fire brigade and control room
Propagation of the fire into other plant areas
Smoke removal operations
Utilization of pre-planned strategies
Procedure adherence regarding fire alarm response, fire brigade dispatch, fire alarm
activation, containment evacuation alarm initiation, and assessment of emergency
plan actions
b. Findings
No findings were identified.
1R06 Flood Protection Measures (71111.06 - 2 samples)
.1 Internal Flooding Review
a. Inspection Scope
The inspectors reviewed the UFSAR, the site flooding analysis, and plant procedures to
assess susceptibilities involving internal flooding. The inspectors also reviewed the CAP
to determine if NextEra identified and corrected flooding problems and whether operator
actions for coping with flooding were adequate. The inspectors also focused on the
control building to verify the adequacy of equipment seals located below the flood line,
floor and water penetration seals, watertight door seals, common drain lines and sumps,
sump pumps, level alarms, control circuits, and temporary or removable flood barriers.
b. Findings
No findings were identified.
Enclosure
9
.2 Annual Review of Cables Located in Underground Bunkers/Manholes
a. Inspection Scope
The inspectors conducted an inspection of two samples of underground
bunkers/manholes subject to flooding that contain cables whose failure could affect risk-
significant equipment. The inspectors performed walkdowns of risk-significant areas,
specifically manholes No. 5 and No. 11, which contain safety-related cables routed to
the service water pumphouse from essential switchgear. The inspectors verified water
level in the sump/manhole to ensure that the cables were not submerged. The
inspectors verified that the bunkers/manholes were dewatered in accordance with
station procedures.
b. Findings
No findings were identified.
1R07 Heat Sink Performance (71111.07A - 1 sample)
a. Inspection Scope
The inspectors reviewed the B emergency diesel generator (EDG) jacket water heat
exchanger to determine its readiness and availability to perform its safety functions.
The inspectors reviewed the design basis for the component and verified NextEras
commitments to NRC Generic Letter 89-13. The inspectors observed actual
performance tests for the heat exchangers and/or reviewed the results of previous
inspections of the B EDG jacket water and similar heat exchangers. The inspectors
discussed the results of the most recent inspection with engineering staff and reviewed
pictures of the as-found and as-left conditions. The inspectors verified that NextEra
initiated appropriate corrective actions for identified deficiencies. The inspectors also
verified that the number of tubes plugged within the heat exchanger did not exceed the
maximum amount allowed.
b. Findings
No findings were identified.
1R08 In-service Inspection (71111.08 - 1 sample)
a. Inspection Scope
From April 4 to April 11, 2014 and from April 13 to April 16, 2014, the inspectors
conducted a review of NextEras implementation of in-service inspection (ISI) program
activities for Seabrook Unit 1. These activities monitor the reactor coolant system
pressure boundary, risk significant piping and components and the containment to
identify degradation, complete evaluations and make repairs or replacements as
required. Sample selection was based on the inspection procedure objectives and risk
priority of those pressure retaining components in systems where degradation could
result in a significant increase in risk. The inspectors observed in-process non-
destructive examinations (NDE), reviewed documentation records, and interviewed
inspection personnel to verify that the non-destructive examination activities performed
Enclosure
10
as part of Period 2 of the third 10-year interval of the Seabrook ISI Program, during
refueling outage OR16, were conducted in accordance with the requirements of the
ASME Boiler and Pressure Vessel Code Section XI, 2004 Edition, No Addenda.
Nondestructive Examination (NDE) and Welding Activities
ASME Code-Required Examinations
The inspectors reviewed the equipment calibration sheet and the inspection data sheets
from the ultrasonic testing (UT) examination of the pressurizer nozzle inner radius weld
(RC E 10 A IR). There were no recordable indications from this examination and 100
percent coverage was achieved.
The inspectors reviewed the equipment calibration sheet and inspection data sheets
from the UT examination of the pressurizer nozzle weld (RC E-10 A-NZ). There were no
recordable indications from this examination; however coverage was limited to 74
percent due to the nozzle configuration.
The inspectors reviewed the equipment calibration sheet and the inspection data sheets
from the UT examination of the Steam Generator (SG) A nozzle to pipe weld (RC E
11A 2B NZ). There were no recordable indications from this examination and 100
percent coverage was achieved.
The inspectors reviewed the equipment calibration sheet and inspection data sheets
from the UT examination of the pressurizer nozzle weld (RC E-10 2A-NZ). There were
no recordable indications from this examination; however coverage was limited to 98
percent due to the nozzle configuration.
The inspectors reviewed the equipment calibration sheet and the inspection data sheets
from the UT examination of the SG A pipe to nozzle weld inner radius (RC E 11 2A-IR).
There were no recordable indications from this examination and 100 percent coverage
was achieved.
The inspectors reviewed the equipment calibration sheet and the inspection data sheets
from the UT examination of the SG A nozzle to pipe weld inner radius (RC E-11A 2B
IR). There were no recordable indications from this examination and 98 percent
coverage was achieved.
The inspectors reviewed the equipment calibration sheets and the inspection data
sheets from the UT examination of the pressurizer nozzle weld (RC E-11 2B-NZ). There
were no recordable indications from this examination and 98 percent coverage was
achieved.
The inspectors reviewed the inspection data sheet from the bare metal visual
examination of the reactor vessel closure head (RVCH) and control rod drive mechanism
(CRDM) nozzle penetrations performed in accordance with ASME Code Case 792-1.
The inspectors reviewed certifications of the NDE technicians performing these
examinations and verified that the inspections were performed in accordance with
approved procedures and that the results were reviewed and evaluated by certified
Level III NDE personnel.
Enclosure
11
Other Containment Liner Examinations
The inspectors reviewed additional UT results completed by NextEra to examine the
containment liner. NextEra performed ultrasonic thickness measurements of 51, one-
square foot area sample locations of the containment liner. These measurements were
taken from the inside of the containment at elevation -26 feet and were spaced
approximately every 6 degrees in azimuth around the containment circumference in
accessible areas. The containment liner thicknesses were measured to be nominally
0.375 inch in thickness at all locations, and no exterior or interior surface degradation of
the liner was evident.
The inspectors reviewed the visual testing (VT) procedures and the recorded inspection
results for the inspection of an anomaly on the upper dome portion of the containment
liner. The inspectors determined these inspections were conducted in accordance with
the ASME Code Section XI, Subsection IWE. An engineering evaluation of the
inspection results concluded that no observable change had occurred in the anomaly
and recommended re-inspection during the next refueling outage, OR17.
The inspectors reviewed the records of a water intrusion condition in the fuel transfer
tube vault which had been identified in 2009. One wall of the fuel transfer tube vault is
also part of the containment liner. In April 2011, NextEra staff made an entry into the
vault area and conducted an inspection of the liner condition inside the vault. Some
areas of the containment liner portion of the fuel tube vault had minor, light rust which
was removed. UT measurements of the liner thickness in 2011 verified that all
previously rusted areas on the liner were greater than 0.375 inch. Also, NextEra staff
made repairs to the non-liner portion of the vault wall which had shown signs of leakage.
Because this repair was not on the containment liner portion of the vault, no repair has
been made under the ASME Code, Section IWE in the fuel transfer vault. The
inspectors determined there were no examinations completed during OR16. An
examination was not required because the leak was stopped in 2011 and a repair was
not made to the containment liner.
Review of Originally Rejectable Indications Accepted by Evaluation
The inspectors reviewed a volumetric examination data sheet for an indication identified
during a previous (OR15) inspection of a piping weld in line SI- 0251-07-09. A flaw
evaluation completed by a qualified NDE Level III individual later determined that the
flaw was acceptable per ASME Code, Subsection XI, Table IWB-3514-2.
The inspectors reviewed the NDE report of a weld flaw detected in the J-groove weld
during the ultrasonic (UT) examination of RVCH penetration No. 57. These inspections
are required by 10 CFR 50.55(a) and Code Case N-729-1. NextEra staff performed a
manual eddy current (EC) inspection to determine if the indication was fully contained
within the weld. The EC exam showed that the indication was completely contained by
the weld material and was acceptable per the requirements of the ASME Code without
repair.
Repair/Replacement Activities
The inspectors reviewed WO package 40108170-01 for replacement and testing
activities of pressure relief valve SI-V-60, the D accumulator relief valve. The
inspectors verified that the replacement valve was successfully tested with a leak test
Enclosure
12
pressure of 630.5 psig. The inspectors also reviewed the VT-2 visual examination data
sheet for SI-V-60, and the Form A of the Flanged Joint Torque Traveler for the valve.
The inspectors also reviewed the Field Work Closeout Form and the ASME Form NIS-
2A Repair/Replacement Certification Record. The inspectors determined that the
replacement was completed in accordance with ASME Section XI, Repair/Replacement
requirements.
The inspectors reviewed WO 40220458-03 associated with the shop fabrication of piping
to replace the service water pump P-41 B/D discharge piping with AL6XN piping material
per engineering change EC278717. The inspectors reviewed the ASME Section XI
Repair/Replacement Plan Traveler, the Form A Weld Traveler, and the liquid penetrant
test records to verify that the replacement piping had been fabricated in accordance with
the change document. The inspectors verified that the Repair/Replacement Plan had
been completed correctly in accordance with the ASME Section XI Code.
PWR Vessel Upper Head Penetration (VUHP) Inspection Activities
The inspectors reviewed the NextEra calculations of Effective Degradation Years (EDY)
and Re-inspection Year (RIY) for the Seabrook Unit 1 Reactor Vessel Upper Head
completed prior to OR16. Based on these calculation results, NextEra completed an
inspection of the Unit 1 VUHP J-groove welds during the OR16 refueling outage.
These calculations were performed in accordance with the requirements of 10 CFR
50.55a(g)(6)(ii)(D) and the ASME Boiler and Pressure Vessel Code Case N 729 1,
Alternative Examination Requirements for PWR Reactor Vessel Upper Heads, to
ensure the structural integrity of the reactor vessel head pressure boundary.
The inspectors reviewed visual inspection reports (Data Sheets) of the remote bare
metal visual examination of the exterior surface of Unit 1 VUHP to confirm appropriate
inspection coverage was achieved and to verify that no boric acid leakage or wastage
had occurred.
The inspectors observed the ultrasonic examination of several RVCH penetrations and
reviewed the data from the inspection of penetration No. 57 which identified a recordable
indication in the J-groove weld. The inspectors reviewed the Eddy Current (ET)
inspection data used to verify the location of the indication. This inspection verified that
the indication was enclosed within the J-groove weld and could be accepted for
continued use without repair. There were no other recordable indications detected in the
other reactor vessel upper head nozzle penetrations.
Boric Acid Corrosion Control (BACC) Inspection Activities
The inspectors reviewed the boric acid corrosion control program, which is conducted
in accordance with Seabrook Unit 1 Station procedures and NextEra procedures,
discussed the program with the boric acid program owner, and sampled photographic
inspection records of boric acid leaks found on safety significant piping and components
inside the Seabrook Unit 1 containment. The walkdowns were conducted by NextEra
personnel and were directly observed by the NRC Resident Inspectors on April 1, 2014,
during the initial containment entry for the OR16 outage. The inspectors reviewed a
sample of leaks observed and reported, the identification and documentation of non-
conforming conditions identified in the corrective action program and reviewed a sample
of boric acid evaluations completed by engineering to repair or monitor the conditions.
Enclosure
13
The inspectors verified that potential deficiencies identified during the walkdowns were
entered into NextEras corrective action program and reviewed a sample of engineering
evaluations of the conditions reported to verify that the corrective actions were
consistent with the requirements of the NextEra procedures and 10 CFR 50, Appendix
B, Criterion XVI.
Steam Generator (SG) Tube Inspection Activities
The inspectors observed a portion of SG eddy current testing (ECT) examinations, data
evaluation, and documentation review of the final list of pluggable tubes from this
inspection.
The inspectors remotely observed a sample of the Unit 1 SG eddy current tube
examinations which were based upon Seabrook's operating experience and the
assessment of past degradation mechanisms.
The scope consisted of the following examinations:
100 percent bobbin coil probe examination of SG 'B'
20 percent bobbin coil probe examination of SG 'A', SG 'C', and SG 'D', including
Outside Diameter Stress Corrosion Cracking (ODSCC) susceptible tubes, and tubes
susceptible to tube-to-tube wear
100 percent +Point examination of dings/dents > 5 volts on HL (including U-Bend)
in SG 'C'
20 percent +Point examination of dings/dents > 5 volts on HL (including U-bend)
Resolution of all bobbin coil probe "I" codes
Visual inspection of all plugs (mechanical and welded)
Visual inspection of channel heads in all SG's
The inspectors reviewed a sample of the indications identified in the SGs during the
OR15 eddy current inspections to verify that they were consistent with the potential
degradation mechanisms that may be observed during the current OR16 inspections
as documented in SG-SGMP-13-21, Revision 3, Steam Generator Degradation
Assessment for the Seabrook OR16 Refueling Outage. The inspectors also reviewed
the Condition Monitoring and Operational Assessments from the prior outage, OR15.
The inspectors verified that the SG eddy current tube examinations were performed
in accordance with Unit 1 Technical Specifications and the Unit 1 Steam Generator
Program by reviewing the SG tube eddy current test results to verify that no in-situ
pressure testing was required, no tubes required stabilization, and no increased primary-
to-secondary leakage had occurred over the operating cycle.
The inspector also verified that the tubes that exhibited degradation and did not meet
acceptance criteria were plugged (3 tubes, due to anti-vibration bar wear) or sleeved (0
tubes) using the alternate repair criteria per Generic Letter 95-05, Voltage-Based Repair
Criteria for Westinghouse Steam Generator Tubes Affected by ODSCC during the OR16
inspection.
Enclosure
14
The inspectors verified that SG tubes that exhibited degradation examination screening
criteria was in accordance with the Electric Power Research Institute (EPRI) Steam
Generator Guidelines and flaw sizing was in accordance with EPRI examination
technique specification sheet. During OR16, NextEra plugged three tubes in C SG due
to anti-vibration bar wear. The total tubes plugged in all four SG's, at the completion of
OR16, was 185.
NextEra staff did not conduct foreign object search and retrieval (FOSAR) on the
secondary side of the SGs during OR16, because the secondary side of the SG's were
not opened and inspected during this outage.
The inspectors reviewed the Westinghouse eddy current testing procedures and verified
that NextEra completed the steam generator inspections in accordance with the
requirements of NEI 97-06, Pressurized Water Reactor Steam Generator Examination
Guidelines, Revision 7. The inspectors also reviewed the Westinghouse procedure
qualification certifications and a sample of data analysts' personnel certifications. The
inspectors reviewed a sample of eddy current qualification records for Primary and
Secondary resolution analysts, Independent Quality Data Analysts, and Utility Level III
Quality Data Analysts.
Identification and Resolution of Problems
The inspectors reviewed a sample of Seabrook Station Unit 1 condition reports, which
identified NDE indications, deficiencies and other non-conforming conditions since the
previous OR15 outage and during the current OR16 outage. The inspectors verified that
nonconforming conditions were properly identified, characterized, evaluated and entered
into the corrective action program.
b. Findings
No findings were identified.
1R11 Licensed Operator Requalification Program (71111.11 - 2 samples)
.1 Quarterly Review of Licensed Operator Requalification Testing and Training
a. Inspection Scope
The inspectors observed licensed operator simulator training on June 25, 2014, which
included a loss of coolant accident coincident with a loss of off-site power. The
inspectors evaluated operator performance during the simulated event and verified
completion of risk significant operator actions, including the use of abnormal and
emergency operating procedures. The inspectors assessed the clarity and effectiveness
of communications, implementation of actions in response to alarms and degrading plant
conditions, and the oversight and direction provided by the control room supervisor. The
inspectors verified the accuracy and timeliness of the emergency classification made by
the shift manager and the TS action statements entered by the shift technical
advisor. Additionally, the inspectors assessed the ability of the crew and training staff to
identify and document crew performance problems.
Enclosure
15
b. Findings
No findings were identified.
.2 Quarterly Review of Licensed Operator Performance in the Main Control Room
a. Inspection Scope
The inspectors observed licensed operator performance in the main control room during
plant shutdown in preparation for a planned refueling outage (OR16), and observed
operations staff event response to a reactor trip and transition to Mode 3 on April 1,
2014. Additionally, inspectors observed containment fire response on April 16, 2014;
initial reactor coolant system (RCS) draindown and heatup activities on April 4 and 21,
2014; approach to criticality and transition to Mode 2 on April 23, 2014; control room
turnover on June 10, 2014; and RCS boration and turbine control system reboot on
June 12, 2014. The inspectors observed applicable test performance to verify that
procedure use, crew communications, and coordination of activities between work
groups similarly met established expectations and standards.
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness (71111.12 - 2 samples)
a. Inspection Scope
The inspectors reviewed the samples listed below to assess the effectiveness of
maintenance activities on structure, system, and component (SSC) performance and
reliability. The inspectors reviewed system health reports, CAP documents,
maintenance WOs, and maintenance rule (MR) basis documents to ensure that
NextEra was identifying and properly evaluating performance problems within the scope
of the MR. For each sample selected, the inspectors verified that the SSC was properly
scoped into the MR in accordance with 10 CFR 50.65 and verified that the (a)(2)
performance criteria established by NextEra staff was reasonable. As applicable, for
SSCs classified as (a)(1), the inspectors assessed the adequacy of goals and corrective
actions to return these SSCs to (a)(2). Additionally, the inspectors ensured that NextEra
staff was identifying and addressing common cause failures that occurred within and
across MR system boundaries.
4000-RM-10 Snubber failure on April 11, 2014
Structures monitoring program - RHR vaults on May 23, 2014
b. Findings
Introduction: The inspectors identified a finding of very low safety significance (Green)
because NextEra did not perform adequate evaluations of safety-related RHR vaults.
Specifically, additional technical evaluation and analysis was not adequately conducted
on the safety-related A and B RHR concrete vaults when it was determined that they
exceeded the quantitative limits specified in NextEra procedures.
Enclosure
16
Description: NextEras Engineering Department Standard 36180, Structural Monitoring
Program provides guidance for the conduct of the structural condition monitoring
program to meet the requirements of 10CFR 50.65, the Maintenance Rule. The
procedure provides a systematic approach for evaluation of plant structures to provide
reasonable assurance that those structures are capable of fulfilling their intended safety
function. This is accomplished, in part, by periodic reviews of the condition of plant
structures via systematic walkdowns. Additionally, NextEras structural monitoring
program commits to the requirements specified in the American Concrete Institute
standard ACI 349.3R-96, Evaluation of Existing Nuclear Safety Related Concrete
Structures for the evaluation of conditions identified in safety-related concrete
structures. This commitment was formally implemented by procedure revision on
July 29, 2013.
NextEras structural monitoring program procedure states that measurable
discontinuities exceeding specified ACI Tier II quantitative limits shall be considered
unacceptable and in need of further technical evaluation and that further evaluation
should consider the use of other inspection, testing or analytical tools to obtain condition
and functional information of the structures in question.
While performing a plant walkdown on May 23, 2014, the inspectors identified several
instances of concrete conditions in the A and B RHR vaults that exceeded the
quantitative Tier II criteria specified in NextEras procedure. These conditions included:
spalling greater than 20 millimeter (mm) in depth, passive cracks greater than 1 mm,
and staining of an undefined source on concrete surfaces. The load bearing walls of
the RHR vaults are approximately 30 inches thick and are reinforced with number
8 vertical reinforcing bars spaced at 12 inches each face and number 9 horizontal
reinforcing bars spaced at 9 inches each face.
The inspectors reviewed NextEras routine structures monitoring program walkdown
evaluations conducted on the A RHR vault on October 10, 2013 and the B RHR vault
on March 25, 2014. The inspectors noted that NextEra personnel had identified the
conditions, but had not conducted adequate technical evaluation of those conditions.
Specifically, NextEras analysis of the conditions in the RHR vaults relied largely on
engineering judgment and did not use any other inspection, testing or analytical tools to
obtain additional condition and functional information of the two vaults.
The inspectors consulted with regional specialists, reviewed a subsequent licensee
evaluation of the RHR vaults, and concluded that the RHR vaults remain operable and
have reasonable assurance of structural integrity. This conclusion is based on the
significant reduction in crack width below the concrete cover, and the absence of
degradation or distress in the concrete surrounding the cracks. The observed cracks
do not show any indications that they are alkali-silica reaction (ASR) related/induced
cracks Rather, the cracking is more likely due to shrinkage and/or settlement-induced
stress relief. Additional engineering review and analysis is warranted to more clearly
identify and understand the cause(s) of the observed cracking in the RHR vaults.
NextEra entered the failure to perform adequate technical evaluations on concrete
structures exceeding Tier II quantitative requirements into the Corrective Action Program
(AR 01929460), and planned to perform a formal technical evaluation of the A and B
RHR vault conditions in accordance with their structural monitoring program procedure
and the ACI 349.3R-96 standard.
Enclosure
17
Analysis: The inspectors determined that NextEras failure to perform an adequate
technical evaluation of conditions identified in the A and B RHR vaults that exceeded
the Tier II quantitative criteria in the structures monitoring program procedure was a
performance deficiency within NextEras ability to foresee and correct. The performance
deficiency was considered to be more than minor because it affected the protection
against external factors attribute of the Mitigating Systems cornerstone and its objective
to ensure the availability, reliability, and capability of systems that respond to initiating
events to prevent undesirable consequences. Specifically, the inspectors concluded that
the reliability of the structures was affected in that they exceeded the specified Tier II
limits without the performance of further technical evaluations. The issue was evaluated
in accordance with IMC 0609, Appendix A, The Significance Determination Process
for Findings At-Power, and determined to be of very low safety significance (Green)
because it did not represent an actual loss of function of at least a single train for greater
than its TS Allowed Outage Time or two separate safety systems out-of-service for
greater than its TS Allowed Outage Time. This finding is related to the cross-cutting
area of Human Performance - Procedure Adherence, because NextEra did not follow
processes, procedures, and work instructions. Specifically, NextEra personnel did not
perform an adequate technical evaluation of two safety-related concrete structures that
exceeded the quantitative criteria requiring such an evaluation [H.8].
Enforcement: Enforcement action does not apply because the performance deficiency
did not involve a violation of a regulatory requirement. (FIN 05000443/2014003-01,
Inadequate Technical Evaluation of Safety-Related Structures)
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 6 samples)
a. Inspection Scope
The inspectors reviewed station evaluation and management of plant risk for the
maintenance and emergent work activities listed below to verify that NextEra performed
the appropriate risk assessments prior to removing equipment for work. The inspectors
selected these activities based on potential risk significance relative to the reactor safety
cornerstones. As applicable for each activity, the inspectors verified that NextEra
personnel performed risk assessments as required by 10 CFR 50.65(a)(4) and that the
assessments were accurate and complete. When NextEra performed emergent work,
the inspectors verified that operations personnel promptly assessed and managed plant
risk. The inspectors reviewed the scope of maintenance work and discussed the results
of the assessment with the stations probabilistic risk analyst to verify plant conditions
were consistent with the risk assessment. The inspectors also reviewed the TS
requirements and inspected portions of redundant safety systems, when applicable, to
verify risk analysis assumptions were valid and applicable requirements were met.
Lowering of reactor water level for reactor vessel head removal on April 3, 2014
Unit Auxiliary Transformers (UATs) and RHR OOS for maintenance on April 8, 2014
Mid-loop operations for steam generator nozzle dam removal on April 13, 2014
Risk assessment for 1-ED-X-2-A/B OOS for main generator breaker repairs while
entering Mode 4 from Mode 5, Mode 3 from Mode 4, and Mode 2 from Mode 3 on
April 22, 2014
'B' containment instrument air compressor corrective maintenance on June 2, 2014
Battery charger return to service on June 19, 2014
Enclosure
18
b. Findings
No findings were identified.
1R15 Operability Determinations and Functionality Assessments (71111.15 - 4 samples)
a. Inspection Scope
The inspectors reviewed operability determinations for the following degraded or non-
conforming conditions:
A EFW past operability due to oil leak identified on March 19, 2014
Vital inverter EDE-1B increased DC input amps on May 25, 2014
RCP undervoltage (UV) time delay relays out of tolerance on June 9, 2014
Seat leakage on 1-MS-V-393 on June 19, 2014
The inspectors selected these issues based on the risk significance of the associated
components and systems. The inspectors evaluated the technical adequacy of the
operability determinations to assess whether TS operability was properly justified and
the subject component or system remained available such that no unrecognized
increase in risk occurred. The inspectors compared the operability and design criteria
in the appropriate sections of the TSs and UFSAR to NextEras evaluations to determine
whether the components or systems were operable. Where compensatory measures
were required to maintain operability, the inspectors determined whether the measures
in place would function as intended and were properly controlled by NextEra. The
inspectors determined, where appropriate, compliance with bounding limitations
associated with the evaluations.
b. Findings
No findings were identified.
1R18 Plant Modifications (71111.18 - 1 sample)
a. Inspection Scope
The inspectors reviewed the temporary modifications listed below to determine whether
the modifications affected the safety functions of systems that are important to safety.
The inspectors reviewed 10 CFR 50.59 documentation and post-modification testing
results, and conducted field walkdowns of the modifications to verify that the temporary
modifications did not degrade the design bases, licensing bases, and performance
capability of the affected systems. In addition, the inspectors reviewed modification
documents associated with the upgrade, including associated engineering changes,
correspondence with the vendor, industry operating experience, environmental and
seismic qualifications, as well as the 10 CFR 50.59 documentation and post-modification
testing results, as applicable.
Engineering Change 250048, main feed pump turbine digital upgrade
Enclosure
19
b. Findings
No findings were identified.
1R19 Post-Maintenance Testing (71111.19 - 6 samples)
a. Inspection Scope
The inspectors reviewed the post-maintenance tests for the maintenance activities
listed below to verify that procedures and test activities ensured system operability
and functional capability. The inspectors reviewed the test procedure to verify that the
procedure adequately tested the safety functions that may have been affected by the
maintenance activity, that the acceptance criteria in the procedure was consistent with
the information in the applicable licensing basis and/or design basis documents, and
that the procedure had been properly reviewed and approved. The inspectors also
witnessed the test or reviewed test data to verify that the test results adequately
demonstrated restoration of the affected safety functions.
CBS-V2 following electrical maintenance on valve motor, on April 1, 2014
RHR pump 8B comprehensive test following maintenance on April 14, 2014
Retest for strainer 11 bypass leak on April 15, 2014
BC-1A energize and parallel following maintenance on May 22, 2014
CS-FCV-111A actuator diaphragm replacement on June 11, 2014
B service water train comprehensive test on June 25, 2014
b. Findings
No findings were identified.
1R20 Refueling and Other Outage Activities (71111.20 - 1 sample)
a. Inspection Scope
The inspectors reviewed the stations work schedule and outage risk plan for the
maintenance and refueling outage (OR16), which was conducted April 1 to 22, 2014.
The inspectors reviewed NextEras development and implementation of outage plans
and schedules to verify that risk, industry experience, previous site-specific problems,
and defense-in-depth were considered. During the outage, the inspectors observed
and evaluated the following outage activities:
Shutdown and cooldown operations, and transition to Mode 5 and entry into RHR
operations
Configuration management, including maintenance of defense-in-depth,
commensurate with the outage plan for the key safety functions and compliance with
the applicable technical specifications when taking equipment out of service
Implementation of clearance activities and confirmation that tags were properly hung
and that equipment was appropriately configured to safely support the associated
work or testing
Installation and configuration of reactor coolant pressure, level, and temperature
instruments to provide accurate indication and instrument error accounting
Enclosure
20
Status and configuration of electrical systems and switchyard activities to ensure that
technical specifications were met
Monitoring of decay heat removal (RHR) operations
Impact of outage work on the ability of the operators to operate the spent fuel pool
cooling system
Reactor water inventory controls, including flow paths, configurations, alternative
means for inventory additions, and controls to prevent inventory loss
Activities that could affect reactivity , including fuel handling and receipt inspections
Fatigue management
Tracking of startup prerequisites, including mode transition reviews, walkdown of the
primary containment to verify that debris had not been left which could block the
emergency core cooling system suction strainers, and startup and ascension to full
power operation
Reactor start-up and plant heat-up activities
Identification and resolution of problems related to refueling outage activities
Consistent with TI 2515/188: Inspection of Near-Term Task Force Recommendation
2.3 - Seismic Walkdowns, reviewed deferred inspections due to inaccessibility from
seismic walkdowns conducted in 2012. These activities were performed on: 4160-
volt safety bus No. 6, located in the B essential switchgear room, and 480-volt
buses E63 and E64, located in containment.
b. Findings
No findings were identified.
1R22 Surveillance Testing (71111.22 - 9 samples)
a. Inspection Scope
The inspectors observed performance of surveillance tests and/or reviewed test data of
selected risk-significant SSCs to assess whether test results satisfied TSs, the UFSAR,
and NextEra procedure requirements. The inspectors verified that test acceptance
criteria were clear, tests demonstrated operational readiness and were consistent with
design documentation, test instrumentation had current calibrations and the range and
accuracy for the application, tests were performed as written, and applicable test
prerequisites were satisfied. Upon test completion, the inspectors considered whether
the test results supported that equipment was capable of performing the required safety
functions. The inspectors reviewed the following surveillance tests:
Reactor vent paths cold shutdown and 18 month surveillance on April 1, 2014
CBS-V-17 and CBS-V-18 local leakage rate tests on April 6, 2014 (containment
isolation valve)
Containment enclosure emergency exhaust filter system 18 month surveillance on
April 9, 2014 (containment isolation valve)
RC-V-88 and RC-V-89 local leakage rate tests on April 9, 2014
Phase B CBS and CVI actuation 18 month surveillance on April 10, 2014
EFW comprehensive flow test on April 23, 2014
Containment personnel hatch door seal leakage test on May 21, 2014 (inservice
testing)
Primary system sample on June 24, 2014
RCS steady state leak rate calculation on June 28, 2014 (RCS leakage detection)
Enclosure
21
b. Findings
No findings were identified.
2. RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2RS1 Radiological Hazard Assessment and Exposure Controls (71124.01 - 1 sample)
a. Inspection Scope
During April 14 to 17, 2014, the inspectors reviewed NextEra performance in assessing
the radiological hazards and exposure control for OR16. The inspectors used the
requirements in 10 CFR Part 20 and guidance in Regulatory Guide (RG) 8.38 Control
of Access to High and Very High Radiation Areas for Nuclear Plants, TSs, and NextEra
procedures required by TSs as criteria for determining compliance.
Radiological Hazard Assessment
The inspectors reviewed the last two radiological surveys from the primary channel
heads in the four steam generators and from the reactor cavity. The inspectors
evaluated whether the thoroughness and frequency of the surveys were appropriate
for the given radiological hazard.
The inspectors selected the following radiologically-significant work activities:
Steam Generator Eddy Current Testing and Tube Plugging
Reactor Cavity Work during OR16, Includes Reactor Head Lift/Set
Replace Reactor Head O-ring
The inspectors evaluated whether continuous air monitors (CAMs) were located in areas
that were representative of actual work areas. The inspectors evaluated the NextEra
program for monitoring levels of loose surface contamination in areas of the plant.
The inspectors reviewed several radiation work permits (RWP) used to access locked
high radiation areas (LHRA) and evaluated if the specified work control instructions and
control barriers were consistent with TS requirements for LHRA.
The inspectors observed the Access Control Point location where NextEra monitors
material leaving the radiological control area and inspected the methods used for
control, survey, and release of these materials. The inspectors observed the
performance of personnel surveying and releasing material for unrestricted use and
evaluated whether the work was performed in accordance with plant procedures.
The inspectors assessed whether the radiation monitoring instrumentation used for
equipment release and personnel contamination surveys had appropriate sensitivity
for the contamination present.
The inspectors evaluated the adequacy of radiological controls, required surveys,
radiation protection job coverage, and contamination controls. The inspectors evaluated
Enclosure
22
NextEras use of electronic personal dosimeters (EPDs) in high noise areas that were
also high radiation areas (HRAs).
The inspectors assessed whether radiation monitoring devices were placed on the
individuals body consistent with NextEra procedures. The inspectors assessed whether
the dosimeter was placed in the location of highest expected dose or that NextEra
properly implemented an NRC-approved method of determining effective dose
equivalent. The inspectors reviewed the application of dosimetry to effectively monitor
exposure to personnel in high-radiation work areas with significant dose rate gradients.
Instructions to Workers
The inspectors reviewed several RWPs for work within airborne radioactivity areas with
the potential for individual worker internal exposures.
For these RWPs, the inspectors evaluated airborne radioactive controls and monitoring,
including potential for significant airborne levels. The inspectors assessed applicable
containment barrier integrity and the operation of temporary high-efficiency particulate
air ventilation systems.
Radiological Hazards Control and Work Coverage
The inspectors discussed with first-line health physics supervisors the controls in place
for special areas that have the potential to become very high radiation areas (VHRAs)
during certain plant operations. The inspectors assessed whether these plant
operations require communication beforehand with the health physics group, so as to
allow corresponding timely actions to properly control and monitor the radiation hazards.
Radiation Worker Performance
The inspectors observed the performance of radiation workers with respect to stated
radiation protection (RP) work requirements. The inspectors assessed whether workers
were aware of the radiological conditions in their workplace and the RWP controls/limits
in place, and whether their behavior reflected the level of radiological hazards present.
RP Technician Proficiency
The inspectors observed the performance of the RP technicians with respect to
controlling radiation work. The inspectors evaluated whether technicians were aware of
the radiological conditions in their workplace and the RWP controls/limits, and whether
their behavior was consistent with their training and qualifications with respect to the
radiological hazards and work activities.
The inspectors reviewed two radiological problem reports since the last inspection that
attributed the cause of the event to RP technician error. The inspectors evaluated
whether there was an observable pattern traceable to a similar cause. The inspectors
assessed whether this perspective matched the corrective action approach taken by
NextEra to resolve the reported problems.
Enclosure
23
b. Findings
No findings were identified.
2RS2 Occupational ALARA Planning and Controls (71124.02 - 1 sample)
a. Inspection Scope
During April 14 to 17, 2014, the inspectors assessed performance with respect to
maintaining occupational individual and collective radiation exposures as low as is
reasonably achievable (ALARA) for OR16. The inspectors used the requirements in
10 CFR Part 20, RG 8.8 - Information Relevant to Ensuring that Occupational Radiation
Exposures at Nuclear Power Plants will be As Low As Is Reasonably Achievable, RG
8.10 - Operating Philosophy for Maintaining Occupational Radiation Exposure As Low as
Is Reasonably Achievable, TSs, and NextEra procedures required by TSs as criteria for
determining compliance.
Radiological Work Planning
The inspectors selected the following work activities that had the highest exposure
significance.
Pre-Job ALARA Review 14-02 OR16 Steam Generator Eddy Current Testing and
Tube Plugging
Pre-Job ALARA Review 14-03 OR16 In Service Inspection
Pre-Job ALARA Review 14-07 OR16 Fuel Handling Project
Pre-Job ALARA Review 14-09 OR16 RCP Seal Replacement
Pre-Job ALARA Review 14-10 OR16 Scaffolding
Pre-Job ALARA Review 14-13 Replace Reactor Ventillation Ducting Under Vessel
with New Design
The inspectors reviewed the ALARA work activity evaluations, exposure estimates,
and exposure reduction requirements. The inspectors determined whether NextEra
reasonably grouped the radiological work into work activities, based on historical
precedence and industry standards. The inspectors compared the results achieved
(actual dose) with the intended dose for these work activities. The inspectors compared
the person-hour estimates provided by maintenance planning and other groups to the
RP group actual person-hours for the work activity, and evaluated the accuracy of these
time estimates. The inspectors assessed the reasons for any inconsistencies between
intended and actual work activity doses.
Verification of Dose Estimates and Exposure Tracking Systems
The inspectors reviewed the assumptions and basis for the collective dose estimates for
routine operations and the refueling outage. The inspectors reviewed applicable
procedures to determine the methodology for estimating exposures from specific work
activities and for department and station collective dose goals.
The inspectors evaluated whether the licensee had established measures to track, trend,
and to reduce occupational doses for ongoing work activities. The inspectors assessed
Enclosure
24
whether dose threshold criteria were established to prompt additional ALARA planning
and controls.
The inspectors evaluated the licensees method of adjusting exposure estimates, or
re-planning work, when unexpected changes in scope or emergent work were
encountered. The inspectors assessed whether adjustments to exposure estimates
were based on sound RP and ALARA principles or if they were just adjusted to account
for failures to plan/control the work.
Radiation Worker Performance
The inspectors observed radiation worker and RP technician performance during work
activities being performed in radiation areas, airborne radioactivity areas, and LHRAs.
The inspectors evaluated whether workers demonstrated the ALARA philosophy in
practice and whether there were any procedure or RWP compliance issues.
b. Findings
No findings were identified.
2RS3 In-Plant Airborne Radioactivity Control and Mitigation (71124.03)
a. Inspection Scope
During April 14 to 17, 2014, the inspectors verified whether in-plant airborne
concentrations were being controlled consistent with ALARA principles and the use
of respiratory protection devices on-site did not pose an undue risk to the wearer.
The inspectors used the requirements in 10 CFR Part 20, the guidance in RG 8.15
Acceptable Programs for Respiratory Protection, RG 8.25 Air Sampling in the
Workplace, NUREG-0041 Manual of Respiratory Protection Against Airborne
Radioactive Material, TSs, and NextEra procedures required by TSs as criteria for
determining compliance.
Inspection Planning
The inspectors reviewed the Updated Final Safety Analysis Report to identify areas of
the plant designed as potential airborne radiation areas and any associated ventilation
systems or airborne monitoring instrumentation. This review included instruments used
to identify changing airborne radiological conditions. The inspectors reviewed the
respiratory protection program and a description of the types of respiratory protection
devices used. The inspectors reviewed the procedures for maintenance, inspection, and
use of respiratory protection equipment including self-contained breathing apparatus, as
well as, procedures for air quality maintenance. The inspectors reviewed reported
performance indicators to identify any related to unintended dose resulting from intakes
of radioactive material.
Engineering Controls
The inspectors reviewed the licensees use of permanent and temporary ventilation to
determine whether the licensee uses ventilation systems as part of its engineering
controls to control airborne radioactivity. The inspectors reviewed procedural guidance
Enclosure
25
for use of installed plant systems to reduce dose and assessed whether the systems are
used during high-risk activities.
The inspectors selected two temporary ventilation system used to support work in
contaminated areas. The inspectors assessed whether the use of these systems was
consistent with NextEra procedural guidance and ALARA. The inspectors assessed
whether the licensee had established threshold criteria for evaluating levels of airborne
beta-emitting and alpha-emitting radionuclides.
Use of Respiratory Protection Devices
The inspectors selected two work activities where respiratory protection devices were
used to limit the intake of radioactive materials, and assessed whether the licensee
performed an engineering evaluation concluding that the use of respirators is not
required based on ALARA. The inspectors also evaluated whether the licensee had
established means (such as routine bioassay) to determine if the level of protection
provided by the respiratory protection devices during use was at least as good as that
assumed in the licensees work controls and dose assessment.
The inspectors assessed whether respiratory protection devices used to limit the intake
of radioactive materials was certified by the National Institute for Occupational Safety
and Health/Mine Safety and Health Administration (NIOSH/MSHA) or have been
approved by the NRC. The inspectors evaluated whether the devices were used
consistent with their NIOSH/MSHA certification or NRC approval.
Problem Identification and Resolution
The inspectors evaluated whether problems associated with the control and mitigation
of in-plant airborne radioactivity were being identified by the licensee at an appropriate
threshold and were properly addressed for resolution in the licensee corrective action
program. The inspectors assessed whether the corrective actions were appropriate for
a selected sample of problems involving airborne radioactivity and were appropriately
documented by the licensee.
b. Findings
No findings were identified.
2RS4 Occupational Dose Assessment (71124.04)
a. Inspection Scope
During April 14 to 17, 2014, the inspectors verified that occupational dose is
appropriately monitored, assessed and reported by NextEra. The inspectors used the
requirements in 10 CFR Part 20, the guidance in RG 8.13 - Instructions Concerning
Prenatal Radiation Exposures, RG 8.36 - Radiation Dose to Embryo Fetus, RG 8.40 -
Methods for Measuring Effective Dose Equivalent from External Exposure, TSs, and the
licensees procedures required by TSs as criteria for determining compliance.
Enclosure
26
Inspection Planning
The inspectors reviewed the results of NextEra RP program audits related to internal
and external dosimetry. The inspectors reviewed the most recent National Voluntary
Laboratory Accreditation Program (NVLAP) report on the principal dosimetry used to
establish dose of legal record.
A review was conducted of NextEra procedures associated with dosimetry operations,
including issuance/use of external dosimetry, and assessments of external and internal
dose for radiological incidents. The inspectors evaluated whether NextEra had
established procedural requirements for determining when external dosimetry and
internal dose assessments are required.
External Dosimetry
The inspectors evaluated whether the NextEra dosimetry vendor is NVLAP accredited
and if the approved irradiation test categories for each type of personnel dosimeter used
are consistent with the types and energies of the radiation present.
The inspectors evaluated the onsite storage of dosimeters before issuance, during use,
and before processing/reading. The inspectors also reviewed the guidance provided to
radiation workers with respect to use of dosimeters.
The inspectors assessed the use of EPDs to determine if NextEra uses a correction
factor to address the response of the EPD as compared to the dosimeter of legal record
for situations when the EPD is used to assign dose.
The inspectors reviewed three dosimetry occurrence reports or corrective action
program documents for adverse trends related to EPDs. The inspectors assessed
whether NextEra had identified any adverse trends and implemented appropriate
corrective actions.
Internal Dosimetry
Routine Bioassay (In Vivo)
The inspectors reviewed procedures used to assess the dose from internally deposited
radionuclides using whole body count (WBC) equipment. The inspectors evaluated
whether the procedures addressed methods for differentiating between internal and
external contamination, the release of contaminated individuals, determining the route of
intake and the assignment of dose.
The inspectors reviewed the whole body count process to determine if the frequency of
measurements was consistent with the biological half-life of the radionuclides available
for intake.
The inspectors reviewed NextEra evaluation for use of its portal radiation monitors as a
passive monitoring system. The inspectors assessed if instrument minimum detectable
activities were adequate to determine the potential for internally deposited radionuclides
sufficient to prompt an investigation.
Enclosure
27
Special Bioassay (In Vitro)
There were no internal dose assessments obtained using In Vitro results for the
inspectors to review, i.e., no urinalysis or fecal sample results.
The inspectors reviewed the vendor laboratory quality assurance program and assessed
whether the laboratory participated in an industry recognized cross-check program
including whether out-of-tolerance results were reviewed, evaluated and resolved
appropriately.
Internal Dose Assessment - Airborne Monitoring
NextEras had not performed any internal dose assessments using airborne/derived air
concentration monitoring during the period reviewed.
Special Dosimetric Situations
Declared Pregnant Workers
The inspectors assessed whether NextEra informs workers of the risks of radiation
exposure to the embryo/fetus, the regulatory aspects of declaring a pregnancy, and the
specific process to be used for voluntarily declaring a pregnancy.
The inspectors reviewed the records of one individual who had declared pregnancy
during the current assessment period and evaluated whether the NextEras radiological
monitoring program (internal and external) for declared pregnant workers is technically
adequate to assess the dose to the embryo/fetus. The inspectors reviewed exposure
results and monitoring controls that were implemented for declared pregnant workers
during the inspection period.
Dosimeter Placement and Assessment of Effective Dose Equivalent for External
Exposures
The inspectors reviewed the NextEra methodology for monitoring external dose in non-
uniform radiation fields where large dose gradients exist. The inspectors evaluated the
NextEra criteria for determining when alternate monitoring, such as use of multi-badging,
is to be implemented. The inspectors reviewed selected dose assessments performed
using multi-badging to evaluate whether the assessment was performed consistent with
procedures and dosimetric standards.
Shallow Dose Equivalent
The inspectors reviewed one dose assessments for shallow dose equivalent for
adequacy. The inspectors evaluated the NextEra method for calculating shallow dose
equivalent from distributed skin contamination.
Neutron Dose Assessment
The inspectors evaluated the NextEras neutron dosimetry program, including dosimeter
types and radiation survey instrumentation.
Enclosure
28
The inspectors reviewed neutron exposure occurrences and assessed whether
(a) dosimetry and instrumentation was appropriate for the expected neutron spectra,
(b) there was sufficient sensitivity for low dose and/or dose rate measurement, and
(c) neutron dosimetry and neutron detection instruments were properly calibrated. The
inspectors also assessed whether interference by gamma radiation had been accounted
for in the calibration.
Assigning Dose of Record
For the special dosimetric situations reviewed in this section, the inspectors assessed
how NextEra assigns dose of record for total effective dose equivalent, shallow dose
equivalent, and lens dose equivalent. This included an assessment of external and
internal monitoring results, supplementary information on individual exposures, and
radiation surveys when dose assignment was based on these techniques.
Problem Identification and Resolution
The inspectors assessed whether problems associated with occupational dose
assessment were being identified by NextEra at an appropriate threshold and are
properly addressed for resolution in the licensee corrective action program. The
inspectors assessed the appropriateness of the corrective actions for a selected sample
of problems documented by the licensee involving occupational dose assessment.
b. Findings
No findings were identified.
2RS5 Radiation Monitoring Instrumentation (71124.05)
a. Inspection Scope
During May 19 to 23, 2014, the inspectors reviewed the accuracy and operability of
radiation monitoring instruments that are used to protect occupational workers and to
protect the public from nuclear power plant operations. The inspectors used the
requirements in 10 CFR Part 20, 10 CFR Part 50 Appendix A - Criterion 60 Control of
Release of Radioactivity to the Environment and Criterion 64 Monitoring Radioactive
Releases, 10 CFR 50 Appendix I Numerical Guides for Design Objectives and Limiting
Conditions for Operation to meet the Criterion As Low as is Reasonably Achievable for
Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents, 40 CFR
Part 190 Environmental Radiation Protection Standards for Nuclear Power Operations,
NUREG 0737 Clarification of Three Mile Island Corrective Action Requirements,
TSs/Offsite Dose Calculation Manual (ODCM), applicable industry standards, and
NextEras procedures required by TSs as criteria for determining compliance.
Inspection Planning
The inspector reviewed the Seabrook Station UFSAR to identify radiation instruments
associated with monitoring area radiation, airborne radioactivity, process streams,
effluents, materials/articles, and workers. Additionally, the inspectors reviewed the
associated TS requirements for post-accident monitoring instrumentation. The
inspectors reviewed a listing of in-service survey instrumentation including: air
Enclosure
29
samplers, small article monitors, radiation monitoring instruments, personnel
contamination monitors, portal monitors, and whole-body counters. The inspectors
assessed whether an adequate number and type of instruments were available to
support operations. The inspectors reviewed NextEra and third-party evaluation reports
of the radiation monitoring program since the last inspection. The inspectors reviewed
procedures that govern instrument source checks and calibrations, including instruments
used for monitoring transient high radiological conditions and instruments used for
underwater radiation surveys.
Walkdowns and Observations
The inspectors walked down four effluent radiation monitoring systems, including
RM6454, Storm Drain Liquid Effluent Monitor; RM6509, Liquid Waste Test Tank
Discharge Effluent Monitor; RM6528-1, Plant Vent Wide Range Gas Monitor Effluent
Monitor; and RM6526-1, Containment Air Particulate Monitor. The inspectors assessed
whether the effluent/process monitor configurations align with what is described in the
The inspectors selected five portable survey instruments in use or available for issuance
and assessed calibration and source check stickers for currency, as well as, instrument
material condition and operability.
The inspectors observed NextEra staff performance in conducting source checks for the
following types of portable survey instruments: MGPI Telepole, Eberline AMP-100 Area
Radiation Monitor, Ludlum 44-9 GM Frisker, Eberline RM-14 Pancake Frisker and
Eberline RO-2A Ion Chamber. The inspectors assessed whether high-range
instruments are source checked on all appropriate scales.
The inspectors walked down five area radiation monitors (ARMs) and four CAMs to
determine their adequacy and operability. The inspectors compared the ARM response
(via local readout or remote control room indications) with actual area radiological
conditions for consistency.
The inspectors selected the Argos 4AB personnel alpha/beta contamination monitor, the
GEM-5 gamma portal monitors, the CHRONOS-4 large article monitor, and the SAM-12
small article monitor located at the health physics (HP) Control Point, and evaluated
whether the periodic source checks were performed in accordance with the
manufacturers recommendations and NextEra procedures.
Calibration and Testing Program
Process and Effluent Monitors
The inspectors selected three process monitor instruments: RM6532-1, Primary
Auxiliary Building (PAB) Air particulate; RM6526-2, Containment Radiogas; and
RM6482-1/2, Main Steam Line Monitors and evaluated whether channel calibration
and functional tests were performed consistent with Seabrook Stations TSs/ODCM.
The inspectors assessed whether; (a) NextEra calibrated its monitors with National
Institute of Standards and Technology (NIST) traceable sources; (b) the primary
calibrations adequately represented the plant radionuclide mix; (c) when secondary
Enclosure
30
calibration sources were used, the sources were verified by comparison with the primary
calibration source; and (d) NextEra channel calibrations encompassed the instruments
alarm set-point range.
Laboratory Instrumentation
The inspectors assessed laboratory analytical instruments used for radiological analyses
to determine whether daily performance checks and calibration data indicate that the
frequency of the calibrations is adequate and there were no indications of degraded
performance. The inspectors assessed whether appropriate corrective actions were
implemented in response to indications of degraded performance.
Whole Body Counter (WBC)
The inspectors reviewed calibration records for the WBC and the methods and sources
used to perform functional checks on the WBC before daily use and assessed whether
calibration and check sources were appropriate and align with the plants radionuclide
mix and that appropriate calibration phantoms were used.
Portal Monitors, Personnel Contamination Monitors, and SAMs
The inspectors selected one of each type of these instruments and verified that the
alarm set-point values are reasonable to ensure that licensed material is not released
from the site. The inspectors reviewed calibration documentation for each instrument
selected and reviewed the calibration methods to determine consistency with the
manufacturers recommendations.
Portable Survey Instruments, ARMs, Electronic Dosimetry, and Air Samplers/CAMs
The inspectors reviewed calibration documentation for the following portable instruments
in use: Eberline AMS-4 Continuous Air Monitor, Eberline RO-2A Ion Chamber, Fluke
Biomedical 451B Ion Chamber, RADECO HD-29A Air Sampler, MGP AMP 200 Area
Monitor Probe and DMC 2000 Electronic Personal Dosimeter (EPD). For portable
survey instruments and ARMs, the inspectors reviewed detector measurement geometry
and calibration methods and reviewed the use of its instrument calibrator as applicable.
Instrument Calibrator
The inspectors reviewed the current radiation output values for the licensees portable
survey and ARM instrument calibrator units. The inspectors assessed that the licensee
had verified calibrator output over the range of the exposure rates/dose rates using an
ion chamber/electrometer. The inspectors verified that the measuring devices had been
calibrated by a facility using NIST traceable methods and were properly applied by the
licensee in performing radiation survey instrument calibrations.
Calibration and Check Sources
The inspectors reviewed the NextEras waste stream characterization per 10 CFR Part
61, Licensing Requirements for Land Disposal of Radioactive Waste, to assess
whether calibration sources used were representative of the types and energies of
radiation encountered in the plant.
Enclosure
31
Problem Identification and Resolution
The inspectors evaluated whether problems associated with radiation monitoring
instrumentation were being identified by the licensee at an appropriate threshold and
were properly addressed for resolution in the licensee corrective action program. The
inspectors assessed the appropriateness of the corrective actions for a selected sample
of problems documented by the licensee that involve radiation monitoring
instrumentation.
b. Inspection Findings
No findings were identified.
4. OTHER ACTIVITIES
4OA1 Performance Indicator Verification (71151)
Reactor Coolant System (RCS) Specific Activity and RCS Leak Rate (2 samples)
a. Inspection Scope
The inspectors reviewed NextEras submittal for the RCS specific activity and RCS
leak rate performance indicators for the period of April 1, 2013 to March 31, 2014. To
determine the accuracy of the performance indicator data reported during those periods,
the inspectors used definitions and guidance contained in NEI Document 99-02,
Regulatory Assessment Performance Indicator Guideline, Revision 7. The inspectors
also reviewed RCS sample analysis and logs of daily measurements of RCS leakage
and activity, and compared that information to the data reported by the performance
indicator.
b. Findings
No findings were identified.
4OA2 Problem Identification and Resolution (71152 - 1 sample)
.1 Routine Review of Problem Identification and Resolution Activities
a. Inspection Scope
As required by Inspection Procedure 71152, Problem Identification and Resolution, the
inspectors routinely reviewed issues during baseline inspection activities and plant
status reviews to verify that NextEra entered issues into the CAP at an appropriate
threshold, gave adequate attention to timely corrective actions, and identified and
addressed adverse trends. In order to assist with the identification of repetitive
equipment failures and specific human performance issues for follow-up, the inspectors
performed a daily screening of items entered into the CAP and periodically attended
condition report screening meetings.
Enclosure
32
b. Findings
No findings were identified.
.2 Semi-Annual Trend Review
a. Inspection Scope
The inspectors performed a semi-annual review of site issues, as required by Inspection
Procedure 71152, Problem Identification and Resolution, to identify trends that might
indicate the existence of more significant safety issues. In this review, the inspectors
included repetitive or closely-related issues that may have been documented by NextEra
outside of the CAP, such as trend reports, performance indicators, major equipment
problem lists, system health reports, MR assessments, and maintenance or CAP
backlogs. The inspectors also reviewed NextEras CAP database for the first and
second quarters of 2014 to assess CRs written in various subject areas (equipment
problems, human performance issues, etc.), as well as individual issues identified during
the NRCs daily CR review (Section 4OA2.1). The inspectors reviewed NextEras
quarterly trend report for the first quarter of 2014, conducted under PI-AA-207-1000,
Station Self-Evaluation and Trending, Revision 1, to verify that NextEra personnel were
appropriately evaluating and trending adverse conditions in accordance with applicable
procedures.
b. Findings and Observations
No findings were identified.
In general, the inspectors did not identify significant issues with the content and
utilization of the station trending report, including any identified adverse trends, potential
adverse trends, or management awareness areas. As a result, the inspectors evaluated
a sample of departments that are required to provide input into the quarterly trend
reports, which included operations, maintenance and engineering departments. This
review included a sample of issues and events that occurred over the course of the past
two quarters to objectively determine whether issues were appropriately considered or
ruled as emerging or adverse trends, and in some cases, verified the appropriate
disposition of resolved trends. The inspectors verified that these issues were addressed
within the scope of the corrective action program, or through department review and
documentation in the quarterly trend report for overall assessment.
For example, the inspectors noted that consistent with the issue identified in Section
4OA3.3, regarding the surveillance test failure, NextEra had identified that a critical
warm-up period identified in manufacturer documents was not translated to applicable
surveillance procedures. While this was a dated issue from several years ago, this
aspect of design interface weaknesses was identified by electrical maintenance following
the implementation of a design change for protective relays. However, while not
identified specifically as a trend, the issue of appropriately ensuring design interface
documents are appropriately revised or updated is being addressed through the
corrective action program for the issue referenced later in this report.
Enclosure
33
Another example identified by the inspectors involved the use of the CAP (versus the
trend report) to address trends in the use of the RCS Leak Rate Program. Operations
personnel coordinated the use of the CAP and the Training Department to ensure the
differences seen in the calculation of identified RCS leak rate were well understood by
the control room operators, but required some potential training due to some
weaknesses identified in this area.
4OA3 Follow-Up of Events and Notices of Enforcement Discretion (71153 - 3 samples)
.1 (Closed) Licensee Event Report (LER) 05000443/2013-001: Failure to Enter Technical
Specification Following Discovery of SW Leak
As documented in NRC Inspection Report, 50-443/2013-005, the inspectors
dispositioned a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V,
Instructions, Procedures, and Drawings, and an associated violation of technical
specification (TS) 3.7.4, because NextEra did not follow the requirements of station
procedure EN-AA-203-1001, Operability Determinations/ Functionality Assessments.
Specifically, NextEra did not properly evaluate and document an adequate basis for
operability, when relevant information was available that would have challenged the
reasonable expectation of operability threshold for a service water (SW) through-wall
leak that degraded incrementally from weepage on August 7, 2013, to a significantly
larger leak on August 28, 2013. NextEra completed a temporary non-code repair of the
flaw with the installation of a weldolet on September 1, 2013, following NRC review and
approval of a relief request. Additionally, during refueling outage OR16, in April 2014,
NextEra removed the temporary weldolet and installed a new piping segment to
complete the restoration of the section of degraded piping that had developed a leak in
August 2013. Extent of condition inspections and liner repairs were also completed for
similar piping configurations to the degraded piping segment that were inaccessible
during power operation. Moreover, the subject LER was submitted by NextEra when
they had concluded, on October 30, 2013, following a review of the actions taken for the
SW leak in question, that the plant had operated in a condition prohibited by technical
specifications for 24 days, from August 8, 2013 to September 1, 2013. The inspector
reviewed several extent of condition inspections, evaluations and repairs, the actual
degraded pipe restoration activity (See Section 1R19 for the final acceptance leak test),
and verified the adequacy of NextEras additional corrective actions for the performance
issues that contributed to the identified TS violation. These actions included: (1)
mentoring of individuals on documentation requirements for prompt operability
determinations, (2) procedure revisions to ensure additional barriers are in place for
future engineering evaluations regarding field information, assumptions and supervisory
reviews, (3) corrective action program requirements were reinforced, and (4) a case
study of this service water leak issue was identified to reinforce the relative issues to
station personnel.
As a result of this inspection, no additional findings or violations of NRC requirements
were identified. This LER is closed.
Enclosure
34
.2 (Closed) Licensee Event Report (LER) 05000443/2014-001: Reactor Trip Due to Delay
in Bus Transfer Resulting in Reactor Coolant Pump (RCP) Loop Low Flow
a. Inspection Scope
On April 1, 2014, at 00:26 while operating at approximately 15 percent power following
turbine shutdown and removal of the main generator from service, Seabrook station
experienced an automatic reactor trip on reactor coolant two loop loss of flow. The loss
of flow was the result of the unexpected closure of the main generator breaker (MGB) B
phase which caused the 345kV Bus 6 to de-energize and isolate the MGB. All buses
transferred to the reserve auxiliary transformers as designed; however, a slight delay in
the automatic transfer for bus 1 resulted in two RCPs tripping. The RCPs tripping
resulted in an automatic reactor trip due to reactor coolant loop low flow. The
emergency feedwater system actuated on low steam generator level, and all plant
equipment functioned as expected. NextEra personnel completed a root cause
evaluation to determine the cause of the reactor trip. Corrective actions include: revising
procedures to add controls regarding the potential risk, ensure the use of guarded
equipment controls, and minimizing the time spent with the main generator breaker in
local.
The inspectors reviewed LER 2014-001-00 and associated corrective actions and
identified a performance deficiency that was characterized as more than minor and is
documented below. This LER is closed.
b. Findings
Introduction. The inspectors identified a Green self-revealing finding, because NextEra
did not ensure that adequate procedural guidance existed in ON1046.12, Operation of
the Main Generator Breaker to limit the likelihood of events that upset plant stability.
Specifically, Seabrook station experienced an automatic reactor trip from approximately
15 percent reactor power on April 1, 2014 when two of four reactor coolant pumps
(RCPs) tripped on low bus voltage. The cause of the reactor trip was directly attributable
to the main generator breaker inadvertently closing and actuating the main generator
multi-function protective relay.
Description. At midnight on April 1, 2014, with reactor power at approximately
15 percent, the station turbine generator was shutdown and the main generator breaker
was opened in preparation for the start of a scheduled refueling outage. The Main
Generator Breaker Selector Switch was subsequently aligned to local in preparation for
scheduled main turbine overspeed testing. At 12:26 a.m., the main generator breaker
B phase unexpectedly closed, actuating the main generator multi-function protective
relay. As a result of the protective relay actuation, 13.8kV buses 1, 2, 3, 4, 5, and 6
automatically transferred from the UATs to the Reserve Auxiliary Transformers (RATs).
However, buses 1 and 5 experienced a delayed transfer from the UATs to the RATs
based on residual bus voltage, and the two RCPs powered from buses 1 and 5 tripped
on low bus voltage as designed. The delayed transfer of buses 1 and 5 was due to a
combination of the delay introduced by the B pole remaining closed and initiating an out
of synchronization condition, and the more restrictive dead band reset of the newer style
synchronization check relays that were installed on buses 1 and 5. The subsequent loss
of reactor coolant flow caused an automatic reactor protection system (RPS) actuation
and reactor trip on loop low flow.
Enclosure
35
NextEra determined that the cause of the B main generator breaker pole unexpected
closure was from the combination of an air leak from a stuck open pressure reducer
valve in the main generator B pole control cabinet, and inadvertent contact with the B
pole local closure push button inside the B pole control cabinet during an investigation
of the air leak by an operator. NextEra determined that the root cause for the event was
inadequate procedural guidance contained in ON1046.12, Operation of the Main
Generator Breaker to communicate the impacts of positioning the Main Generator
Selector Switch to local, take mitigating actions, and minimize time spent at increased
risk configurations. Specifically, ON1046.12 contained no cautions, notes or
prerequisites to ensure that licensee personnel were aware of the potential risk
associated with the Main Generator Selector Switch being placed in local. Specifically,
with the switch in local, the main generator breaker pole disagreement protective
features that would have auto-opened the pole upon inadvertent closure were disabled.
As a result of the lack of procedural information, licensee personnel did not implement
controls in accordance with OP-AA-102-1003, Guarded Equipment, or take measures
to minimize the time spent in the risk significant configuration. NextEra noted that
procedures used for controlling similar evolutions contain additional cautions and notes
that identify the increased risk configuration.
NextEra entered the event into their CAP (AR 01953543), and conducted a root cause
evaluation to determine the root and contributing causes, extent of condition and extent
of cause, and to identify corrective actions to prevent recurrence. NextEra initiated
actions to revise ON1046.12 to add controls to communicate the potential risk
associated with placing the main generator breaker control in local, conducted briefings
with Maintenance groups involved in the event, and evaluated the adequacy of other
Operations procedures that place equipment in a configuration where protective features
are bypassed or defeated. The inspectors reviewed the root cause evaluation and
associated documentation, and determined that NextEra adequately identified the root
and contributing causes and implemented appropriate corrective actions to prevent
recurrence.
Analysis. The inspectors determined that the inadequate procedural guidance contained
in ON1046.12, Operation of the Main Generator Breaker was a performance deficiency
that was within NextEras ability to foresee and correct. The performance deficiency was
more than minor because it was associated with the procedure quality attribute of the
Initiating Events cornerstone, and it adversely affected the cornerstone objective to limit
the likelihood of events that upset plant stability and challenge critical safety functions
during shutdown as well as power operations. The finding was evaluated under IMC
0609, Attachment 4, Phase 1 - Initial Characterization of Findings. The inspectors
determined that the finding is of very low safety significance (Green) because it did not
result both in a reactor trip and the loss of mitigating equipment relied upon to transition
the plant from the onset of the trip to a stable shutdown condition. The finding has a
cross-cutting aspect in the area of Human Performance - Work Management, because
NextEra did not ensure that a process of planning, controlling, and executing work
activities such that nuclear safety is the overriding priority was implemented (H.5).
Specifically, ON1046.12, Operation of the Main Generator Breaker did not contain
adequate procedural guidance regarding the impacts of positioning the Main Generator
Selector Switch to local, take mitigating actions, and minimize time spent at increased
risk configurations [H.5].
Enclosure
36
Enforcement. Enforcement action does not apply because the performance deficiency
did not involve a violation of a regulatory requirement. (FIN 05000443/2014003-02,
Unexpected Main Generator Breaker Pole Closure Results in Reactor Trip)
.3 (Closed) Licensee Event Report (LER) 05000443/2014-002: Reactor Coolant Pump
Undervoltage Time Delay Relay Exceeds Acceptance
On April 6, 2014, during a refueling outage, routine RPS surveillance testing identified
that three of four reactor coolant pump (RCP) undervoltage (UV) reactor trip channels
exceeded the TS channel response time acceptance criteria of 1.5 seconds for the RCP
UV reactor trip function. NextEra determined that since this condition involved multiple
similar components, there is evidence indicating that this condition may have arisen over
time and three channels of RCP UV were concurrently inoperable. This resulted in the
plant operating in a condition prohibited by TSs for approximately seventeen months.
NextEra personnel initiated a root cause evaluation to determine the cause of the
violation and determine appropriate corrective actions, replaced one relay, and adjusted
the remaining relays to acceptable response times.
The inspectors reviewed LER 2014-002-00 and associated corrective actions and
dispositioned the issue as a licensee identified violation of regulatory requirements. The
enforcement aspects of this issue are discussed in Section 4OA7. This LER is closed.
4OA6 Meetings, Including Exit
On July 10, 2014, the inspectors presented the inspection results to Mr. Dean Curtland,
Site vice President, and other members of the Seabrook Station staff. The inspectors
verified that no proprietary information was retained by the inspectors or documented in
this report.
4OA7 Licensee-Identified Violation
The following violation of very low safety significance (Green) was identified by NextEra
and is a violation of NRC requirements which meets the criteria of the NRC Enforcement
Policy for being dispositioned as an NCV.
Technical Specification (TS) Surveillance Requirement 4.3.1.2 requires verification of
the response time of each reactor trip function every 18 months. During the 18
month surveillance testing of the RCP UV channels conducted on April 6, 2014,
three of the four RCP UV relays exceeded their allowable maximum response time,
resulting in their associated UV reactor trip channels exceeding the limit of 1.5
seconds. NextEra determined that the three channels were inoperable. TS 3.3.1,
Reactor Trip System Instrumentation, requires four channels of RCP UV
instrumentation to be operable in Mode 1. With three RCP UV channels inoperable
in Mode 1, the plant is required to initiate a shutdown within one hour in accordance
with TS 3.0.3. NextEra determined that this condition existed from the time the
relays were last calibrated in OR15 (September 20, 2012) until the plant entered
OR16 (April 1, 2014). Contrary to TS 3.0.3, Seabrook station operated in Mode 1
with three of four RCP UV channels inoperable for approximately 17 months without
taking the required TS actions. NextEra entered this issue into the CAP as AR
01964167 and performed a detailed analysis of the impact of the increased channel
Enclosure
37
response time. NextEra, in consultation with Westinghouse, determined that the
safety function of the RCP UV trip channel (prevention of departure from nucleate
boiling) was maintained throughout the period of inoperability. NextEra planned to
develop a maintenance procedure to allow for on-line re-calibration of the RCP
UV relays. The inspectors determined that the finding was of very low safety
significance (Green) in accordance with IMC 0609, Appendix A, Determining the
Significance of Reactor Inspection Findings for At-Power Situations because the
deficiency did not affect a single RPS trip signal to initiate a reactor scram and the
function of other redundant trips or diverse methods of reactor shutdown, did not
involve control manipulations that unintentionally added positive reactivity, and did
not result in a mismanagement of reactivity by operators.
ATTACHMENT: SUPPLEMENTARY INFORMATION
Enclosure
A-1
SUPPLEMENTARY INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
D. Curtland, Site Vice President
T. Vehec, Site Director
R. Dodds, Plant General Manager
J. Berg, Chemistry Supervisor
K. Boehl, Radiation Protection Supervisor
V. Brown, Senior Licensing Engineer
R. Campion, Oversight Supervisor
A. Chesno, Performance Improvement and Licensing
M. Collins, Engineering Director
J. Connolly, Engineering Director
K. Douglas, Maintenance Director
D. Egonis, Repairs and Welding Services
M. Feeney, Instrumentation and Controls Dept. Head
D. Flahardy, Radiation Protection Manager
S. Hamel, Engineering Support
Z. Kuljis, WesDyne International, EC Level III
R. Leider, Engineer
B. McAllister, Nuclear Engineer
M. Ossing, Licensing Manager
D. Perkins, Radiation Protection Analyst
D. Robinson, Chemistry Manager
D, Snyder, Boric Acid Corrosion Control Program Manager
T. Vassallo, Engineering
T. Waechter, Project Manager
B. Westovic, WesDyne International, Waltz Mill Facility
K. Whitney, ISI Program Manager
LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED
Opened/Closed
05000443/2014003-01 FIN Inadequate Technical Evaluation of Safety-
Related Structures (Section 1R12)05000443/2014003-02 FIN Unexpected Main Generator Breaker Pole Closure
Results in Reactor Trip (Section 4OA3.2)
Opened
None
Closed
05000443/2013-001-00 LER Failure to Enter Technical Specification Following
Discovery of SW Leak (Section 4OA3.1)
Attachment
A-2
05000443/2014-001-00 LER Automatic Reactor Trip Due to Delay in Automatic
Bus Transfer Resulting in RCP Loop Low Flow
(Section 4OA3.2)
05000443/2014-002-00 LER Reactor Coolant Pump Undervoltage Time Delay
Relay Exceeds Acceptance (Section 4OA3.3)
LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection
Procedures
ON1090.13, Response to Natural Phenomena Affecting Plant Operations, Revision 1
ON1246.03, GSU Trouble, Revision 5
OP-AA-102-1002, Seasonal Readiness, Revision 3
OS1200.03, Severe Weather Conditions, Revision 2
Condition Reports
01932588 01962633 01964824 01964840 01967006 01968990
01969460 01969599 01970614
Maintenance Orders/Work Orders
40268452 40244709
Miscellaneous
ISO New England Operating Procedure No. 4, Action During a Capacity Deficiency,
Revision 12
Master/Local Control Center Procedure No. 1, Nuclear Plant Transmission Operations,
Revision 13
Master/Local Control Center Procedure No. 2, Abnormal Conditions Alert, Revision 17.1
Seasonal Readiness Memo to Mano Nazar dated 5/24/14
UFSAR Chapter 2
UFSAR Section 8
Section 1R04: Equipment Alignment
Procedures
OS1001.01, Reactor Coolant System Fill and Vent, Revision 25
OS1013.03, Residual Heat Removal Train A Startup and Operation, Revision 27
OS1016.01, Service Water System Fill and Vent, Revision 18
OS1036.01, Aligning the Emergency Feedwater System for Automatic Initiation, Revision 17
OS1056.03, Containment Penetrations, Revision 10
OS1090.05, Component Configuration Control, Revision 57
Condition Reports
00207129 00208974 00208987 01843745 01918488 01955956
01956317 01960210
Attachment
A-3
Miscellaneous
Seabrook Station Updated Final Safety Analysis Report, Chapter 9, Revision 12
Section 1R05: Fire Protection
Condition Reports
01803431 01967187 01969430 1958635 1958639
Miscellaneous
Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, CE-F-1-A
Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, C-F-1-Z, C-F-2-Z, & C-F-3-Z
Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, EFP-F-1-A
Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, PAB-F-3A-Z, PAB-F-3B-Z, &
PAB-F-4-Z
Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, PLT-F-1-0
Procedures
OS1200.00, Response to Fire or Fire Alarm Actuation, Revision 21
FP-AA-104, Fire Protection Program, Revision 0
Section 1R06: Flood Protection Measures
Condition Reports
01969726
Miscellaneous
Seabrook Station Moderate Energy Line Break Study
Seabrook Station UFSAR, Revision 15, Section 3 & Section 9
Maintenance Orders/Work Orders
40209732-06 40209732-07 40277675-01 40241371-01
Drawings
9763-F-310248
9763-F-310249
Section 1R07: Heat Sink Procedures
Procedures
ES1850.017, SW Heat Exchanger Program, Revision 1
PEG-268, Heat Exchanger and NRC GL 89-13 Program, Revision 0
Condition Reports
01809068 01957744 01958294 01959163
Maintenance Orders/Work Orders
40262651
Miscellaneous
Calculation C-S-1-25115, DG Heat Exchanger (DG-E-42A/B) Performance After Tube Plugging,
Revision 0
Attachment
A-4
Section 1R08: In-service Inspection
Procedures
Engineering Department Instructions, EDI N0. 30560, Boric Acid Evaluations, Revision 1
NextEra Energy Program Description ER-AP-121, Steam Generator Integrity, Revision 0
Seabrook Station, Engineering Procedure, Reactor Vessel Head Penetration Ultrasonic
Examination Analysis, ES13-01-08, Revision 00
Seabrook Station, Engineering Procedure, Procedure for Ultrasonic Examination of Reactor
Vessel Head Penetrations, ES13-01-07, Revision 00
Seabrook Station, Engineering Procedure, RPVH Nozzle Bottom OD Surface Eddy Current
Inspection, ES-01-01, Revision 00
Seabrook Station, Engineering Procedure, Perform VE of RPV Top Head At Penetration 57,
Unit 1, 4/11/14
Seabrook Station, Engineering Procedure, Eddy Current Inspection of Preservice and Inservice
Heat Exchanger Tubing, ES01-100, Revision 07
Seabrook Station, Procedure, Steam Generator Eddy Current Data Analysis Guidelines Manual,
Revision 7
Seabrook Station Administrative Procedure, Boric Acid Corrosion Control Program MA 10.3,
Revision 12
Seabrook Station Engineering Procedure, Manual Ultrasonic Procedure for the Examination of
Non-PDI Nozzle Inner Corner Regions, ES10-01-38, Revision 01
Seabrook Station Engineering Procedure, Manual Ultrasonic Procedure for the Examination of
Non-RPV Nozzle-to-Shell Welds in Vessels >2, ES10-01-39, Revision 01
Seabrook Station Reference Manual, Steam Generator Maintenance Reference, SGRE,
Revision 20
Westinghouse Procedure Number MRS-2.4.2 GEN 35, Revision 15, 8/11/11; Eddy Current
Inspection of Preservice and Inservice Heat Exchange Tubing
Westinghouse SG-SGMP-12-15, Revision 1, Seabrook OR15, Condition Monitoring
Assessment and Final Operational Assessment, January 2013
Westinghouse SG-SGMP-13-21, Revision 3, Steam Generator Degradation Assessment for
Seabrook OR16 Refueling Outage, March 2014
Program Documents
EPRI Report 1000975, November 2001; Boric Acid Corrosion Guidebook, Revision 1; Managing
Boric Acid Corrosion Issues at PWR Power Stations
NEI 03-08, January 2010; Guidelines for the Management of Materials Issues, Revision 2
NEI 97-06, Pressurized Water Reactor Steam Generator Examination Guidelines,
Requirements 1013706, Revision 7
Seabrook Station Reference Manual, Inservice Inspection, SIIR Revision 16, 12/23/13
WCAP-15988, Revision 1, February 2005; Generic Guidance for an Effective Boric Acid
Inspection Program for Pressurized Water Reactors
Westinghouse Non-Proprietary Class 3, SG-SGMP-09-22, Revision 2; Seabrook OR13
Condition Monitoring Assessment and Final Operational Assessment, March 2010
Westinghouse Owners Group Letter WOG 05-91, dated March 15, 2005; Subject: Transmittal
of the Final Non-Proprietary Version of WCAP-15988-NP, Revision 1 Entitled Generic
Guidance for an Effective Boric Acid Inspection Program for Pressurized Water
Reactors, February 2005, PA-MSC-0096
Westinghouse Seabrook OR14 Condition Monitoring and Operational Assessment, SG-SGMP-
11-14, Revision 0; April 2011
Attachment
A-5
Westinghouse Seabrook OR16 Condition Monitoring Assessment and Final Operational
Assessment, SG-SGMP-14-10, Revision 0, April 2014
Westinghouse Steam Generator Degradation Assessment for Seabrook OR15 Refueling
Outage, SG-SGMP-12-8, Revision 0, September 2012
Westinghouse Steam Generator Degradation Assessment for Seabrook OR15 Refueling
Outage, SG-SGMP-12-8, Revision 1, September 2012
Program Health Reports
Boric Acid Corrosion Control Program, 7/1/2013 - 9/30/2013
Boric Acid Corrosion Control Program, 10/1/2013 - 12/31/2013
Eddy Current Examination Technique Sheets
Eddy Current Examination Technique Sheet, ETSS #96004.1, Revision 13, April 2010
Eddy Current Examination Technique Sheet, ETSS #96004.3, Revision 13, April 2010
Eddy Current Examination Technique Sheet, ETSS #96005.2, Revision 9, July 2006
Eddy Current Examination Technique Sheet, ETSS #24013.1, Revision 2, August 2006
Eddy Current Examination Technique Sheet, ETSS #10013.1, Revision 1, May 2012
Eddy Current Examination Technique Sheet, ETSS #128411, Revision 3, February 2011
Eddy Current Examination Technique Sheet, ETSS #128413, Revision 3, February 2011
Eddy Current Examination Technique Sheet, ETSS #27091.2, Revision 1, September 2012
Eddy Current Examination Technique Sheet, ETSS #21409.1, Revision 7, May 2010
Eddy Current Examination Technique Sheet, ETSS #21998.1, Revision 4, August 2006
Eddy Current Examination Technique Sheet, ETSS #128424, Revision 3, February 2011
Eddy Current Examination Technique Sheet, Appendix 1, ETSS #124425, Revision 3,
February 2011
Eddy Current Examination Technique Sheet, Appendix 1, ETSS #128431, Revision 2,
February 2011
Eddy Current Examination Technique Sheet, Appendix 1, ETSS #128432, Revision 2,
February 2011
Eddy Current Examination Technique Sheet, ETSS #21410.1, Revision 6, October 2006
Eddy Current Examination Technique Sheet, ETSS #22401.1, Revision 4, August 2006
Eddy Current Examination Technique Sheet, ETSS #96511.1, Revision 16, August 2006
Eddy Current Examination Technique Sheet, ETSS #20510.1, Revision 7, October 2006
Eddy Current Examination Technique Sheet, ETSS #21511.1, Revision 8, July 2006
Eddy Current Examination Technique Sheet, ETSS #279101.1, Revision 1, May 2012
Eddy Current Examination Technique Sheet, ETSS #27901.3, Revision 1, May 2012
Eddy Current Examination Technique Sheet, ETSS #27902.1, Revision 1, May 2012
Eddy Current Examination Technique Sheet, ETSS #27902.2, Revision 1, May 2012
Eddy Current Examination Technique Sheet, ETSS #27902.3, Revision 1, May 2012
Eddy Current Examination Technique Sheet, ETSS #10908.4, Revision 1, May 2012
Eddy Current Examination Technique Sheet, ETSS #20407.1, Revision 7, July 2006
MRS-TRC-2163, Seabrook Appendix H & I Techniques Spring 2014 Inspection, April 2014
Steam Generator Eddy Current Inspection Parameters
ACTS# NAH-01-114, Revision 0, Jan. 15, 2014
ACTS# NAH-02-114, Revision 0, Jan. 16, 2014
ACTS# NAH-03-114, Revision 0, Jan. 16, 2014
ACTS# NAH-04-114, Revision 0, Jan. 16, 2014
ACTS# NAH-05-114, Revision 0, Jan. 16, 2014
Attachment
A-6
ACTS# NAH-06-114, Revision 0, Jan. 16, 2014
ACTS# NAH-07-114, Revision 0, Jan. 15, 2014
ACTS# NAH-08-114, Revision 0, Jan. 15, 2014
ACTS# NAH-09-114, Revision 0, Jan. 15, 2014
ACTS# NAH-10-114, Revision 0, Jan. 16, 2014
ACTS# NAH-11-114, Revision 0, Jan. 15, 2014
ANTS# NAH-A-114, Revision 0, Jan. 9, 2014
ANTS# NAH-B-114, Revision 0, Jan. 9, 2014
ANTS# NAH-C-114, Revision 0, Jan. 9, 2014
ANTS# NAH-D-114, Revision 0, Jan. 9, 2014
ANTS# NAH-D-114, Revision 0, Jan. 9, 2014
Containment References
Calculation C-S-1-10096, Containment Liner Wall Thickness Requirements Guideline
OR16 Containment Liner Dome Anomaly Examination Plan Revision 00; 4/2/14
Revision 1, 1/21/03
Seabrook Station Engineering Procedure ES1807.032, Revision 01; Inservice Inspection
Procedure Primary Containment Section XI IWE Program
United Engineers and Constructors, Inc.; Containment Design Specification For Public Service
Company of New Hampshire Seabrook Station Unit Nos. 1 & 2
Condition Reports (*NRC identified)
00208123 00392489 01792134 01799982 01800069 01800665
01804656 01805955 01806325 01807338 01808286 01809420
01809558 01809566 01812473 01812701 01812870 01814971
01818743 01833942 01837215 01849156 01850224 01856860
01874676 01875085 01875160 01875925 01884233 01884233
01885999 01888956 01892958 01903871 01915704 01924003
01931227 01935241 01938925 01940747 01941757 01945332
01954160 01956084 01958085 01960191* 01960193*
Maintenance Orders/Work Orders
0138224401 0138224402 4020903201 4010817001
Miscellaneous
Code Class 1, 2, and 3 Piping, 6/15/90
NEXTERA Energy, Nuclear Fleet, Process Description, PI-AA-204, Revision 24; Condition
Identification and Screening Process, 1/30/14
NEXTERA Energy, Nuclear Fleet, Process Description, PI-AA-205, Revision 25; Condition
Evaluation and Corrective Action, 1/30/14
NRC Letter Dated 3/28/2007, Subject: Seabrook Station Unit No. 1 - Issuance of Amendment
Re: Technical Specification Task Force (TSTF)-449, Steam Generator Tube Integrity,
Nuclear Regulatory Commission (NRC) Generic Letter 90-05; Guidance for Performing
Temporary Non-Code Repair of ASME
Seabrook Station Reference Manual, SIIR, Revision 16 (contains Alloy 600 examinations),
Sections 8.2, 8.3, and 8.4
Steam Generator Management Program: Pressurized Water Reactor Steam Generator
Examination Guidelines, Revision 7, October 2007
Attachment
A-7
Steam Generator Management Program: Steam Generator Integrity Assessment Guidelines,
Revision 3, November 2009
Section XI Repair/Replacement Samples
W0108170 01, Relief Valve Setpoint Pressure Test of valve 1-SV-V-60, completed on 10/26/12
W0220458 03, Replace P-41B/D Discharge Piping With AL6XN per EC278717, completed
on 11/14/13
NDE Examination Reports & Data Sheets
40209032 01, OR16, Perform VT-2 RPV Top Head Under Insulation
40209032 01, OR16, Perform VT-2 OF RPV Top Head Under Insulation, 4/7/14
GV & VT-3 Examination Data Sheet 000446, Mechanical Penetration PEN-X62, 8/16/01
GV & VT-3 Examination Data Sheet 05-13-148, Mechanical Penetration PEN-X62, 5/12/05
GV & VT-3 Examination Data Sheet 06-13-007, Mechanical Penetration PEN-X62, 10/1/06
GV Examination, 12-13-003,90-180 EL 119 to 192, 10/3/12, 100 percent Accept, 12-13-003
GV Examination Data Sheet 11-13-169, Mechanical Penetration PEN-X62, 5/14/11
GV Examination Data Sheet 12-13-008, Mechanical Penetration PEN-X62, 10/15/12
IWE - VT-3 Examination Data Sheet 09-06-043, Mechanical Penetration PEN-X62, 12/8/10
IWE - VT-3 Examination Data Sheet 09-06-043, Mechanical Penetration PEN-X62, 4/16/08
IWE-VT-1 Examination,90-180 EL 119 to 192, Indication IWE-12-159-176, accept by
engineering evaluation (11 pages) 10/3/12
Liquid Penetrant Examination Data Sheet, SW piping fabrication, Joint F1421, 10/10/13
Liquid Penetrant Examination Data Sheet, SW piping fabrication, Joint F1422, 10/10/13
Liquid Penetrant Examination Data Sheet, SW piping fabrication, Joint F1429, 10/9/13
Liquid Penetrant Examination Data Sheet, SW piping fabrication, Joint F1511, 10/9/13
Liquid Penetrant Examination Data Sheet, SW piping fabrication, Joint F1512, 10/9/13
Liquid Penetrant Examination Data Sheet, SW piping fabrication, Joint F1413, 10/24/13
Liquid Penetrant Examination Data Sheet, SW piping fabrication, Joint F1514, 10/24/13
Liquid Penetrant Examination Data Sheet, SW piping fabrication, Joint F1515, 10/24/13
Liquid Penetrant Examination Data Sheet, SW piping fabrication, Joint F1516, 9/13/13
NextEra, Seabrook Unit 1, OR16 License Renewal Containment Liner UT Exam, 5/7/14
SG A, OR16 Listing of eddy current indications detected and reported, 4/14/14
SG B, OR16 Listing of eddy current indications detected and reported, 4/14/14
SG C, OR16 Listing of eddy current indications detected and reported, 4/14/14
SG D, OR16 Listing of eddy current indications detected and reported, 4/14/14
Ultrasonic Report Data Sheet: NAH-R16-CPO2-57-4150-01Y, Head Penetration #57 (7 Pages)
Ultrasonic Thickness Examination Data Sheet 08-06-170, Mechanical Penetration PEN-X62,
10/8/09
Ultrasonic Thickness Examination Data Sheet 11-01-033, Mechanical Penetration PEN-X62,
5/1/11
UT Data Sheet, 09-06-031, IWE-VT-1 Examination,90-180 EL 119 to 192, 78 percent,
Indication IWE-12-159-176 accept by engineering evaluation, 10/29/09
UT Data Sheet 12-01-143, SI 0251-07, Pipe-to-Pipe weld, 10/8/12, Steps 4.11, 4.12, sign off of
final NDE for SW piping fabrication
UT Data Sheet: 14-01-020, component: RC-E-10 A-IR, Nozzle Inner Radius, NRI, 4/15/14
UT Data Sheet: 14-01-021, component: RC-E-10 A-NZ, nozzle-to-vessel weld, NRI, 4/15/14
UT Data Sheet: 14-01-022, component: RC-E-11A 2A-IR, Nozzle Inner Radius, NRI, 4/15/14
UT Data Sheet: 14-01-023, component: RC-E-11A 2A-NZ, nozzle-to-vessel weld, NRI, 4/15/14
UT Data Sheet: 14-01-024, component: RC-E-11A 2B-IR, Nozzle Inner Radius, NRI, 4/15/14
Attachment
A-8
UT Data Sheet: 14-01-025, component: RC-E-11 2B-NZ, nozzle-to-vessel weld, NRI, 4/15/14
Wes Dyne, RPVH CRDM Nozzle 57 OD Manual Eddy Current (ET) Examination Report
April 2014 OR16 Outage - Final Report, 4/16/14 (15 pages)
Eddy Current Data Acquisition Technicians and Data Analysts Certifications
10232/P4536 10271/E3848 10275/G8167 10281/P2305
10288/S1703 10298/D8884 10305/K9235 10312/S7963
10325/S1253 10339/H1487 10344/P8908 10399/Y1950
18835/P2465 19166/M2691 19257/T8398 20494/W2506
20613/B0613 23225/B4838 25553/H3260 28737/E4963
40256/M8414 45189/J5189 51753/K1753 A2945
A3502 B9540 C0042 C2886
C3274 C9055 C9162 D2172
D4576 D4816 D8021 F1726
F4008 G0071 G3127 H1748
H6377 I2393 J0009 J0145
J1515 J8109 K0073 K2676
K8138 L1107 L1342 L6066
L7003 M3442 P0017 P0116
R0609 R1509 R3919 R7770
S4256 S5339 S5760 S9385
T0042 T3673 V6207 W1758
W2545 Y0624 Z0059
Weld Traveler
40220458-03, 9/25/13, 6 pages, SW Piping Fabrication
Engineering Evaluations, Analyses, Calculations & Standards
ASME IWE VT-1 Examination Indication Containment Dome, AR1809517, 10/4/12,
Revised 10/24/12
ASME IWE VT-1 Examination Indication Containment Dome, AR1809517, 4/3/14,
Revised 4/12/14
Calculation C-S-1-24004, Revision 07, RC System, 9/13/2012; Seabrook Reactor Vessel Head
Effective Degradation Years (EDY) & Re-Inspection Years
Containment Liner IWE Examination Indications AR1646065, 5/15/13
Evaluation #14-013, REVISION 0, 4/14/14; Seabrook Station Reactor Vessel Head Effective
Degradation Years (EDY) & Re-Inspection Year for Cycle 16
Boric Acid Corrosion Control (BACC) Leak Screening
WR94094749, AR01956281; 1-CS-V-143 (Charging to Regen HX Isolation); no
evaluation, 4/12/14
WR94095011, AR01957408; 1-SI-V-25 (Isolation Level Column Accumulator Tank 9B); no
evaluation, 4/12/14
WR94095012, AR01957412; 1-SI-V-223 (Drain Valve-Level Column Accumulator Tank 9D); no
evaluation, 4/12/14
WR94095014, AR01957417; 1-SI-V-57 (Isolation for LT 956 for Accumulator Tank 9D); no
evaluation, 4/12/14
WR94095016, AR01957419; 1-SI-V-55 (Isolation for LT 957 for accumulator Tank 9D); no
evaluation, 4/12/14
Attachment
A-9
WR94095018, AR01957422; 1-SI-V-52 (Isolation sample connection for accumulator Tank 9D);
no evaluation, 4/12/14
Boric Acid Corrosion Control (BACC) Leak Evaluation
AR01792134, B3 Leak from fitting on 1-CBS-P-9-A, 8/8/12
AR01884233, Active Boric Acid Leak, valve 1-CS-TCV-381-B, 6/21/13
AR01915704, Inactive Leak at 1-CS-V-250 Has Become an Active Leak, 10/28/13
AR01948324, 1-CS-FCV-121 Active Boric Acid Packing Leak, 3/14/14
AR01955327, Active Boric Acid Leak at 1-SI-V-139, 4/6/14
Drawings
United Engineers drawing 9763-F-805147, Revision 7, Seabrook Unit 1; Containment Structure
Piping Zones 57E & F Line NOS. 1214, 1216, 1225 & 1226, 12/11/90
Section 1R12: Maintenance Effectiveness
Procedures
EX1805.01. Visual Examination and Functional Testing Program for Snubbers, Revision 11
MS0515.08, Paul-Monroe 2300, 2400, and 2500 Hydraulic Snubber Maintenance, Revision 4
PEG-40, Scoping Changes and Program Interfaces, Revision 5
PEG-45, Maintenance Rule Program Monitoring Activities, Revision 17
Condition Reports
00140964 01282530 01686357 01688626 01725046 01804460
01811427 01929460 01932657 01957128 01959695 01959700
01978310
Maintenance Orders/Work Orders
40288714 40287787
Miscellaneous
ACI 349.3R-02, Evaluation of Existing Nuclear Safety-Related Concrete Structures
EE-10-010, Maintenance Rule - PRA Basis Document PRA Risk Ranking and performance
criteria based on SSPSS-2009, Revision 1
Engineering Department Standard 36180, Structural Monitoring Program, Revision 4
Loss of Fluid Evaluation of PMH 2000 Series Hydraulic Snubbers
Maintenance Support Evaluation 07MSE181, Remove Doors P801 and P802
NEI-99-02, Revision 7
Snubbers System Health Report
Technical Specification Section 3.7.7
Section 1R13: Maintenance Risk Assessments and Emergent Work Control
Procedures
ODI.101, Guarded Equipment Recommendations for Refueling Outages, Revision 13
OP-AA-102-1003, Guarded Equipment, Revision 4
PRA-301, MR (a)(4) Process for On-Line Maintenance Group Instruction, Revision 0
WM-AA-100-1000, Work Activity Risk Management, Revision 0
Condition Reports
01954498
Attachment
A-10
Miscellaneous
PRA-301, MR (a)(4) Process for On-Line Maintenance Group Instruction, Revision 0
Risk Assessment 1-ED-X-2-A/B OOS for Main Generator Breaker Repairs While Entering
Mode 4 from 5, Mode 3 from Mode 4, or Mode 2 from Mode 3
WM-AA-100-1000, Work Activity Risk Management, Revision 0
Section 1R15: Operability Determinations and Functionality Assessments
Procedures
EN-AA-203-1001, Operability Determinations/Functionality Assessments, Revision 16
LX0563.02, Reactor Coolant Pump Undervoltage Channel Calibration and Relay PM,
Revision 11
Condition Reports
01846345 01889301 01896874 01967934 01969615 01973026
01973578 01974039 1949876 1965480
Maintenance Orders/Work Orders
40288839 40245357
Miscellaneous
Adverse Condition Monitoring and Contingency Plan for Seat Leakage on 1-MS-V-393
Breaker on Higher Bus Voltages, Revision 5
Calculation SBC-128, Technical Specifications - Setpoints and Allowable Values, Revision 15
DBD-ESF-01, Engineered Safety Features Response Times Design Basis Document,
Revision 2
EC-280600, Change TAP Settings of Westinghouse Inverters to prevent Tripping AC Input
Maintenance Rule Functional Failure Evaluation, under AR 1949876
UFSAR Section 6.8, Emergency Feedwater System, Revision 15
FP-22849, Terry Turbine Instruction Manual, Revision 1
Engineering Evaluation EE-11-031, FPL - Seabrook Mitigating System Performance Indicator
Basis Document, Revision 00
Section 1R18: Plant Modifications
Procedures
ON1435.05, Feed Pump Turbine Overspeed Trip Test, Revision 5
PR-AA-1008-F01, Owners Acceptance Review Checklist, Revision 0
Condition Reports
01957267 01961655
Miscellaneous
DBD-ED-04, 120 VAC Vital & Non-Vital Instrument Power Design Basis Document, Revision 2
DBD-FW-01, Feedwater System Design Basis Document, Revision 3
Engineering Change 250048, Revision 1
Feedwater System Health Report
Standing Order 14-002, Main Feed Pump Master/Slave Controller Auto/Manual Transfer
Operation
UFCR 12-006, Steam Generator Feed Pump Turbine Digital Upgrade Project, Revision 1
UFSAR Section 7.1.2.5
Attachment
A-11
Drawings
1-NHY-503100 1-NHY-503226 1-NHY-503581
1-NHY-503590 1-NHY-503593 1-NHY-504138
1-NHY-504168 1-NHY-506484 1-NHY-509053
Section 1R19: Post-Maintenance Testing
Procedures
OX1413.08, RH-P-8B Comprehensive Pump Test, Revision 8
OX1416.04, Service Water Quarterly Pump and Discharge Valve Test and Comprehensive
Test, Revision 19
OX1456.81, Operability Testing of IST Valves, Revision 19
OS1048.13, Vital Bus 11A Operation, Revision 10
OS1008.01, Chemical and Volume Control System Makeup Operations, Revision 33
ES1850.012, Air Operated Valve Program Procedure, Revision 4
MN0520.16, Copes-Vulcan Model D-100/1000 Operator Maintenance, Revision 2
MS0517.12, Application and Repair of Protective Coating(s), Revision 12
MS0517.43, Piping Installation and Maintenance, Revision 2
ES1807.025, Inservice Inspection (ISI) Visual Examination Procedure, Revision 5
Condition Reports
01799084 01970614 1956739 1953484 1973225
Maintenance Orders/Work Orders
40202481 40265447 40303546 40238124 40238125 40238501
40238143 40263623 40260904
Miscellaneous
IST Pump Data Sheet, RH-P-8B, 1-RH-OT-0007
IST Pump Data Sheet, SW-P41B, 1-SW-OT-008
IST Pump Data Sheet, SW-P41B, 1-SW-OT-038
IST Pump Data Sheet, SW-P41D, 1-SW-OT-008
IST Pump Data Sheet, SW-P41D, 1-SW-OT-038
Section 1R20: Refueling and Other Outage Activities
Procedures
AD-AA-101-1004, Work Hour Controls, Revision 14
ODI.82A, Mode Change Notice, Revision 19
ODI.82E, Mode Change Checklist, Revision 19
ON1090.04, Containment Entry, Revision 27
OS1000.01, Heatup From Cold Shutdown to Hot Standby, Revision 39
OS1000.02, Plant Startup From Hot Standby to Minimum Load, Revision 28
OS1000.03, Plant Shutdown From Minimum Load to Hot Standby, Revision 26
OS1000.04, Plant Cooldown From Hot Standby to Cold Shutdown, Revision 44
OS1000.05, Power Increase, Revision 25
OS1000.06, Power Decrease, Revision 18
OS1000.07, Approach to Criticality, Revision 13
OS1000.08, Post Trip Review, Revision 29
OS1000.09, Refueling Operation, Revision 28
OS1000.12, Operation With RCS at Reduced Inventory/Midloop Conditions, Revision 12
Attachment
A-12
OS1000.13, Operation With the Reactor Defueled, Revision 03
OS1000.14, Reactor Coolant System Evacuation and Fill, Revision 18
OS1001.01, Reactor Coolant System Fill and Vent, Revision 25
OS1001.02, Draining the Reactor Coolant System for Vessel Head Removal, Revision 17
OS1001.11, Reactor Coolant System Shutdown Level Instrumentation, Revision 9
OS1013.03, Residual Heat Removal Train A Startup and Operation, Revision 27
OS1015.10, Refueling Canal and Cavity Drain, Revision 17
OS1015.18, Setting Containment Integrity for Mode IV Entry, Revision 10
OS1016.11, Contingency Ocean Pump Restoration for SW Work Activities With Ocean Service
Water Pumps Not in Service, Revision 5
OX1406.12, 18 Month Containment and Containment Spray Recirculation Sump Surveillance,
Revision 11
OX1436.13, Turbine Driven Emergency Feedwater Pump Post Cold Shutdown or Post
Maintenance Surveillance and Comprehensive Pump Test, Revision 26
OS1246.01, Loss of Offsite Power Plant Shutdown, Revision 22
PEG-10, System Walkdowns, Revision 21
RS1737, Post Refueling Low Power Physics Testing, Revision 7
Condition Reports
1960185 1960175 1960177 1803194 1799753
Miscellaneous
OS1000.04, Form A, RCS and PZR Cooldown Log, Revision 45, from 4/1 to 4/2/14
PRAE-14-001, OR16 Outage Schedule Shutdown Risk Review, Revision 0
SY-AA-100-1011-F01, Fatigue Assessment Form, Revision 4
Work Order 40203281, Shop Testing of Reactor Coolant Pump Cartridge
Work Order 40202561, Containment and Containment Spray Recirculation Sump Surveillance
Section 1R22: Surveillance Testing
Procedures
CX0901.02, Determination of Dose Equivalent I-131, Revision 12
CS0910.01, Primary Systems Sampling at SS-CP-166A, Revision 18
EX1803.003, Reactor Containment Type B and C Leakage Rate Tests, Revision 12
EX1806.001, RPS And ESFAS Response Time Summation Procedure, Revision 07
EX1808.014, Containment Enclosure Emergency Exhaust Filter System 18 Month Surveillance,
Revision 9
OS1023.66, Containment Enclosure Ventilation System Operation, Revision 17
OX1401.02, RCS Steady State Leak Rate Calculation, Revision 8
OX1401.09, Reactor Vent Paths Cold Shutdown And 18 Month Surveillance, Revision 11
OX1456.94, Train B Phase B, CBS, & CVI Actuation 18 Month Surveillance, Revision 05
OX1456.97, Train A Phase B, CBS, & CVI Actuation 18 Month Surveillance, Revision 02
OX1436.13, Turbine-Driven Emergency Feedwater Pump Post Cold Shutdown or Post
Maintenance Surveillance and Comprehensive Test, Revision 26
OX1460.01, Form A, Weekly Personnel Hatch Air Lock Door Seal testing, Revision 11
Condition Reports (*NRC identified)
01673025 01673033 01954574* 01954972 01955131 01955370
01956011 01958150 01958882 1960453
Attachment
A-13
Maintenance Orders/Work Orders
40201086 40202286 40203229 40204623 40204624 40204884
40204886 40307194 40204750 40258847
Drawings
1-NHY-310900 SH
Section 2RS1: Radiological Hazard Assessment and Exposure Controls
Procedures
HD0958.03, Personnel Survey and Decontamination Techniques, Revision 24
HD0958.04, Posting of Radiologically Controlled Areas, Revision 33
HD0958.19, Evaluation of Dosimetry Abnormalities, Revision 37
HN0958.25, High Radiation Area Control, Revision 37
HN0958.30, Inventory and Control of LHRA or VHRA Keys and Locksets, Revision 26
HN0960.10, Radiological Requirements for Entry Beneath the Reactor Vessel, Revision 26
RP-AA-101, Personnel Monitoring Program, Revision 0
RP-AA-101-1001, Personnel Monitoring Device Issue, Revision 0
RP-AA-101-2004, Method for Monitoring and Assigning Effective Dose Equivalent for High Dose
Gradient Work, Revision 3
RP-AA-102-1002, Dosimetry Data Process for Sentinel, Revision 3
RP-AA-103-1002, High Rad Controls, Revision 1
Audits, Self-Assessments, and Surveillances
Assessment and ALARA Planning and Control, February 3, 2014Seabrook Station Radiation
Protection Department Self Evaluation and Trend Analysis Report
for 4th Quarter 2013, January 31, 2014
Quick Hit Assessment Report 1928716, NRC 71124.01 and .02 Radiological Hazard
Condition Reports
01836289 01855852 01955642 01957485
Miscellaneous
EPRI Standard Radiation Monitoring Program Data for RO 16, April 9, 2014
RWP 14-0004, OR16 Reactor Cavity Decontamination, April 1, 2014
Seabrook 2014 Air Sample Log, March 29 - April 15, 2014
Seabrook HRA/LHRA Briefing Acknowledgement Form, April 15, 2014
Seabrook LHRA In-Service Key Box Log (Containment Alternate Control Point), April 15, 2014
Seabrook LHRA In-Service Key Box Log (RP Access Control Point), April 15, 2014
Seabrook LHRA/VHRA Key Issue Log (Containment Alternate Control Point), April 15, 2014
Seabrook LHRA/VHRA Key Issue Log (RP Access Control Point), April 15, 2014
Seabrook Log of VHRA and LHRA Access Points, April 15, 2014
Seabrook RWP 14-0031, Rx Cavity Work during OR16, Includes Rx Head Lift/Set, April 1, 2014
Seabrook RWP 14-0032, Replace Reactor Head O-ring include Preparations, Decontamination
and QC Inspections, April 11, 2014
Seabrook RWP 14-0038, OR16 Steam Generator Eddy Current Testing and Tube Plugging,
April 16, 2014
Seabrook RWP 14-0052, Replace Lower Plenum, Replace Rx Ventilation Ducting Under-
Vessel, March 21, 2014
Seabrook Survey M-20140409-14, CTB S/G A Initial Survey, April 9, 2014
Attachment
A-14
Seabrook Survey M-20140409-16, CTB S/G B Initial Survey, April 9, 2014
Seabrook Survey M-20140409-18, CTB S/G D Initial Survey, April 9, 2014
Seabrook Survey M-20140409-19, CTB S/G C Initial Survey, April 9, 2014
Seabrook Survey M-20140415-8, CTB-00 Cavity Reactor Cavity with Head Raised 18 Above
Flange, April 15, 2014
Seabrook Survey M-20140416-1, CTB-00 Cavity Reactor Cavity with Head on Flange After
Cavity Decon, April 16, 2016
Seabrook Updated Final Safety Analysis Report
Section 2RS2: Occupational ALARA Planning and Controls
Procedures
RP-AA-104 ALARA Program, Revision 2
RP-AA-104-1000, ALARA Implementing Procedure, Revision 3
Audits, Self-Assessments, and Surveillances
Seabrook Nuclear Oversight Report SBK-14-001, Radiation Protection and Radwaste
Programs, February 24, 2014
SFA 1928716, Quick Hit Assessment Report NRC 71124.01 and .02 Radiological Hazard
Assessment and ALARA Planning and Control, February 3, 2014
Condition Reports
01953467
Miscellaneous
EPRI Standard Radiation Monitoring Program Results through OR15, September 25, 2012
Job In-Progress ALARA Review AR No. 14-JIP-01, Reactor Vessel Dissassembly &
Reassembly 25 percent Review, April 12, 2014
Job In-Progress ALARA Review AR No. 14-JIP-02, Reactor Vessel Dissassembly &
Reassembly 75 percent Review, April 17, 2014
Job In-Progress ALARA Review AR No. 14-JIP-03, OR16 Steam Generator Eddy Current
Testing and Tube Plugging - 25 percent Review, April 15, 2014
Job In-Progress ALARA Review AR No. 14-JIP-08, OR16, Valve Maintenance- 50 percent
Review,
April 14, 2014
Job In-Progress ALARA Review AR No. 14-JIP-09, OR16 Fuel Handling Project - 50 percent
Review,
April 13, 2014
Job In-Progress ALARA Review AR No. 14-JIP-16, Replace Lower Plenum. Replace Rx
Ventilation Ducting Under-Vessel, April 14, 2014
Pre-Job ALARA Review Package 14-02, OR 16 Steam Generator Eddy Current Testing and
Tube Plugging, Febuary 25, 2014
Pre-Job ALARA Review Package 14-03, OR16 In Service Inspection, Febuary 25, 2014
Pre-Job ALARA Review Package 14-07, OR16 Fuel Handling Project, Febuary 25, 2014
Pre-Job ALARA Review Package 14-09, OR16 RCP Seal Replacement, Febuary 25, 2014
Pre-Job ALARA Review Package 14-10, OR16 Scaffolding, Febuary 25, 2014
Pre-Job ALARA Review Package 14-13, Replace Rx Ventillation Ducting Under Vessel with
New Design, Febuary 25, 2014
Seabrook Temporary Sheilding Log for OR16, March 2014
Attachment
A-15
Section 2RS3: In-Plant Airborne Radioactivity Control and Mitigation
Procedures
HD0955.01, Analysis of Smears & Air Samples, Revision 28
HD0955.53, Use of AMS-4, Revision 2
HD0955.71, Setup of ASI Breathing Air System, Revision 0
HD0958.01, Air Sampling, Revision 12
HD0965.01, Respiratory Protection Quality Assurance and Maintenance, Revision 19
HD0965.07, Air Supplied Respiratory Protection Equipment and Distribution System,
Revision18
HD0965.12, Respiratory Equipment Issue and Use, Revision 38
OS1023.68, Containment Air Purge Operation, Revision 19
Audits, Self-Assessments, and Surveillances
SFA 01815118, Perform Quick Hit Self-Assessment - Respiratory Protection Program,
April 5, 2013
SFA 01868045, 2012 Respiratory Protection Program, September 17, 2013
Condition Reports
01880090 01953797 01956903 01957319 01957775
Miscellaneous
Air Sample Result 14-0338, Containment B Steam Generator General Area Platform,
April 15, 2014
Air Sample Result 14-179, Containment El 26 Decon Loop one, April 3, 2014
Air Sample Result 14-206, Containment Cavity during Cavity Flood, April 5, 2014
Air Sample Result 14-266, Containment D Steam Generator Hot Leg Inside Manway,
April 9, 2014
HEPA Portable Ventilation Unit Inventory, April 14, 2014
HPSTID 14-002, Air Systems Breather Box Testing with V4F1 R Delta Suit, March 19, 2014
Respirator Issue Log, April 10-17, 2014
Section 2RS4: Occupational Dose Assessment
Procedures
HD0958.03, Personnel Surveys and Decontamination Techniques, Revision 24
HD0992.02, Issuance and Control of Personnel Monitoring Device, Revision 38
HN0958.39, Multi-Badge Control and Exposure Tracking, Revision 35
HN0958.42, Determination and Control of Dose to an Embryo Fetus, Revision 25
HN0961.29, Internal Dosimetry Assessment, Revision 27
RP 5.3, Expected or Declared Pregnant Worker Exposure Control, Revision 4
RP-AA-101-2004, Method for Monitoring and Assigning Effective Dose Equivalent (EDE) for
High Dose Gradient Work, Revision 2
Audits, Self-Assessments, and Surveillances
SFA 01892653 Self Assessment Quick Hit on Electronic Dosimeter Setpoints,
September 19, 2013
Condition Reports
01896167 01903346 01951842 01952711 01954641 01956098
01956989 01958425
Attachment
A-16
Miscellaneous
HPD0958.03 Form A - Personnel Contamination Report, Westinghouse Worker Right Hand
April 10, 2014
HPD0958.03 Form A - Personnel Contamination Report, Master Lee Worker Left and Right
Forearm, April 11, 2014
HPD0958.03 Form A - Personnel Contamination Report, Operations Worker Right Shoe,
April 11, 2014
HPD0958.03 Form A - Personnel Contamination Report, Bartlett RP Technician Stomach,
April 13, 2014
HPD0958.03 Form A - Personnel Contamination Report, Master Lee Worker Chin,
April 11, 2014
RP 5.1A - Increased Radiation Exposure Request and Authorization, Day Zimmerman Worker
April 1, 2014
RP 5.1A - Increased Radiation Exposure Request and Authorization, Westinghouse Worker,
April 8, 2014
RP 5.3 Expected or Declared Pregnant Worker Exposure Control, Attachment A - Declaration of
Pregnancy, August 28, 2012
RP 5.3 Expected or Declared Pregnant Worker Exposure Control, Attachment B - Statement of
Expected Pregnancy, August 28, 2012
RP 5.3 Expected or Declared Pregnant Worker Exposure Control, Attachment C - Radiological
Controls for Declared Pregnant Workers, August 28, 2012
Section 2RS5: Radiation Monitoring Instrumentation
Procedures
HD0955.03, Use Tennelec APC 175, Revision 9
HD0955.05, Operation Portable Rad Mont Instruments, Revision 21
HD0955.19, Use of Sheppard Model 81 Beam Calibrator, Revision 12
HD0955.30, Use Mini Buck Calibrator, Revision 6
HD0955.39, Use Sheppard Model 89 Box Calibrator, Revision 1
HD0955.42, Operation SAM and Chronos Monitors, Revision 7
HD0955.50, Far West REM-500 Operation, Revision 5
HD0955.53, Use of AMS-4, Revision 4
HD0955.54, Operation TSA Model SPM 906 Portal Monitor, Revision 1
HD0955.62, Use Argos 4AB, Revision 3
HD0955.63, Use Sirius 2 Hand and Foot Counter, Revision 1
HD0955.64, Use MGP DRM 1 and 2 Area Rad Monitors, Revision 5
HD0955.69, Use of GEM 5 Portal Monitor, Revision 2
HD0957.01, Calibration Environmental Air Samplers, Revision 8
HD0958.01, Air Sampling, Revision 14
HD0958.38, Evaluation of 10CFR61 Isotopic Mix, Revision 29
HD0961.31, Canberra Whole Body Counting System Operation, Revision 10
HD0961.32, Canberra WBC Calibration, Revision 1
HD0961.34, Canberra FASTSCAN WBC Operation, Revision 8
HD0963.03, Calibration Eberline E140N Ratemeter, Revision 5
HD0963.08, Calibration Air Sampling Equipment, Revision 15
HD0963.20, Calibration DCA Area Rad Monitor, Revision 2
HD0963.24, Calibration Johnson Extender Teletector, Revision 4
HD0963.27, Calibration Eberline RO2 Bicron RSO5 Ion Chamber, Revision 8
Attachment
A-17
HD0963.28, Calibration MGP DMC 2000 E Dosimeters, Revision 16
HD0963.30, Calibration RO 7 Ion Chamber, Revision 6
HD0963.31, Calibration of Eberline RM-14, Revision 7
HD0963.34, Calibration PNR 4 Neutron Rem Counter, Revision 6
HD0963.44, Calibration of the Bicron MicroRem Meter, Revision 2
HD0963.45, Calibration AMS-4 CAM, Revision 1
HD0963.46, Calibration TSA Model SPM-906 Portal Mont, Revision 3
HD0963.48, Calibration of the MGP AMP-100 AMP-200, Revision 1
HD0963.51, Calibration Argos 4AB, Revision 5
HD0963.52, DRM 2 Area Rad Mont Calibration, Revision 1
HD0963.53, Calibration Sirius 2 Hand Foot Counter, Revision 5
HD0963.54, Calibration Fluke 451B Ion Chamber, Revision 0
HD0963.55, Calibration Fluke 451P Ion Chamber, Revision 1
HD0963.57, Calibration MGP Telepole, Revision 2
HD0963.58, Calibration SAM 12, Revision 2
HD0963.60, Calibration Canberra Chronos 4, Revision 2
HD0963.61, Calibration and maintenance of RADECO HD-29A and AVS-28A Air Sampler
Pump, Revision 1
HD0963.62, Calibration of Canberra GEM-5, Revision 2
HD0963.64, Calibration Ludlum 177, Revision 0
HD0963.65, Calibration Model 9-7 Ion Chamber, Revision 0
HD0963.56, Calibration Canberra S5-APC-GM, Revision 1
HN0955.08, Operation RDMS Continuous Air Monitor, Revision 9
HX0955.32, RDMS Setpoint Determination Rad Monitors, Revision 29
IN1660.601, Dual Channel Calibration ARM 6508, 6517, 6536, 6563, Revision 6
IN1660.604, Single Channel Calibration ARM 6518, 6529, 6540, Revision 6
IN1660.611, RD 10B RD 12 Calibration ARM 6534, 6537, 6550, Revision 6
IN1660.622, Non Safety Related Area Rad Monitors Calibration, Revision 7
IN1660.714, RM 6522, 6531 PAB WPB CAM Calibration, Revision 3
IN1660.731, RM6495 Plant Vent Mid/Hi Range Rad Monitor Calibration, Revision 3
IN1660.812, RM-R-6502 Carbon Delay Beds Inlet Radiation Monitor Calibration, Revision 6
IN1660.813, RM-R-6503 WG Compressors Inlet Radiation Monitor Calibration, Revision 5
IN1660.817 RM 6510, 6511, 6512, 6513 SG Blowdown Calibration, Revision 8
IN1660.990, RM 6486, 6487, 6488, 6489 Portable Continuous Atmosphere Radiation Monitor
Calibration, Revision 6
IN1660.992, RM-R-6454 Storm Drain Effluent Monitor Calibration, Revision 5
IS1660.120, RM-F-6497 Plant Vent Stack Accident Sample Flow Control Calibration, Revision 7
IX1660.110, Plant Vent Stack Flow Transmitter Calibration, Revision 7
IX1660.612, RM-R-6535 A B Manipulator Crane ARM Calibration, Revision 9
IX1660.639, RM 6576A, 6576B Containment Hi Range Rad Mont Calibration Revision 10
IX1660.662, RM-R-6535-A Fuel Manipulator Crane Train A ARM Operation Test, Revision 8
IX1660.663, RM-R-6535 Fuel Manipulator Crane Train B ARM Operation Test, Revision 8
IX1660.689, RM-R-6576-A Containment Hi Range Rad Monitor Operation Test, Revision 9
IX1660.690, RM-R-6576-B Containment Hi Range Rad Mont Operation Test, Revision 8
IX1660.710, RM-R-6506 6507 Control Room Air Intake A B Rad Mont Calibration, Revision 9
IX1660.718, RM 6526 Containment Rad Mont Calibration, Revision 12
IX1660.719, RM-R-6548 Containment Rad Monitor Calibration, Revision 8
Attachment
A-18
IX1660.720, RM-R-6527 COP Trains A B Rad Mont Calibration, Revision 8
IX1660.724, RM-6562 Fuel Storage Bldg Airborne Rad Mont Calibration, Revision 6
IX1660.730, RM-R-6528 Plant Vent Wide Range Gas Monitor Calibration, Revision 9
IX1660.760, RM-R-6506-A Control Room East Air Intake Operation Test, Revision 7
IX1660.761, RM-R-6506-B Control Room West Air Intake Operation Test, Revision 7
IX1660.762, RM-R-6507-A Control Room West Air Intake Operation Test, Revision 7
IX1660.763, RM-R-6507-B Control Room West Air Intake Operation Test, Revision 7
IX1660.768, RM-R-6526 Containment Atmosphere Operation Test, Revision 9
IX1660.769, RM-R-6548 Containment Atmosphere Backup Operation Test, Revision 6
IX1660.770, RM-R-6527-A Containment On-Line Purge Operation Test, Revision 8
IX1660.771, RM-R 6527-B Containment On-Line Purge Operation Test, Revision 9
IX1660.774, RM-R-6562 Fuel Storage Building Ventilation Exhaust Operation Test, Revision 6
IX1660.780, RM-R-6528 Plant Vent Wide Range Gas Monitor Operation Test, Revision 7
IX1660.801, RM-R-6481 6482 Main Steam Line Rad Mont Calibration, Revision 7
IX1660.814, RM-R-6504 Waste Gas Compressor Rad Monitor Calibration, Revision 7
IX1660.814, RM-6503 WG Compressors Discharge Radiation Monitor Calibration, Revision 7
IX1660.815, RM-R-6505 Condenser Air Ejector Discharge Rad Mont Calibration, Revision 2
IX1660.816, RM-R-6509 WLTT Discharge Rad Mont Calibration, Revision 8
IX1660.816, RM-6509 Waste Liquid Test Tanks Discharge Radiation Monitor Calibration,
Revision 7
IX1660.823, RM-R-6515 6516 Loop A B PCCW Rad Mont Calibration, Revision 6
IX1660.824, RM-R-6519 SGBD Flash Tank Discharge Rad Mont Calibration, Revision 9
IX1660.826, RM-R-6521 Turbine Bldg Sump Rad Mont Calibration, Revision 6
IX1660.864, RM-R-6504 WG Compressor Discharge Operation Test, Revision 8
IX1660.872, RM-R-6516 Loop A PCCW Operation Test, Revision 8
IX1660.873, RM-R-6515 Loop B PCCW Operation Test, Revision 8
IX1660.874, RM-R-6519 SB Flash Tank Discharge Operation Test, Revision 7
IX1660.876, RM-R-6521 Turbine Building Sump Pump Discharge Operation Test, Revision 6
Chem SOPs
CS0920.07, Tritium Analysis Liquid Scintillator, Revision 14
CX0917.01, Liquid Effluent Release Setpoints, Revision 20
HD0963.38, Calibration of Ludlum 220 Portable Scaler Ratemeter, Revision 5
HD0963.47, Tennelec Series 5 XLB Smear Counter Calibration, Revision 1
IX1660730, RM-R-6528 Plant Vent Wide Range Gas Monitor Calibration, Revision 9
JS0999.200, Operation of Countroom Analysis System, Revision 3
JS0999.300, Calibration of Gamma Spectroscopy Detectors Using the Count Analysis System,
Revision 2
Audits, Self-Assessments, and Surveillances
Nuclear Oversight Report SBK 14-001, Audit of Radiation Protection and Radwaste, Feb 24,
2014
Quick Hit 1930254 Instruments QHSA-2, December 31, 2013
Quick Hit 1961974, NRC 71124.05 Radiation Monitoring Instrumentation, May 7, 2014
Condition Reports
01767528 01833664 01847095 01856452 01879999 01939604
01941338 01947933 01948312 01961385 01967570 01969397
Attachment
A-19
Miscellaneous
2013 2nd Quarter RM System Health Report, June 30, 2013
2013 3rd Quarter RM System Health Report, September 30, 2013
2013 4th Quarter RM System Health Report, December 31, 2013
2014 1st Quarter RM System Health Report, March 31, 2014
Canberra Argos-4AB SN 0503-107 Calibration Record Printout, January 30, 2014
Canberra Sirius-2B SN 0607-021 Calibration Record Printout, March 20, 2014
Canberra Sirius-2B SN 0607-021 Calibration Record Printout, September 4, 2012
EC 281092, Wide Range Gas Monitor: Replacement of Thomas Model 727CM39 Air Sample
Pump with Metal Bellows Model MB-602, May 2014
HD0963.27 Form A RO-2A Ion Chamber Calibration Record SN 3713, October 29, 2013
HD0963.27 Form A RO-2A Ion Chamber Calibration Record SN 3713, December 1, 2010
HD0963.45 Form A AMS-4 Flow Calibration Record SN 1181, January 25, 2013
HD0963.45 Form A AMS-4 Flow Calibration Record SN 1181, January 30, 2013
HD0963.45 Form A AMS-4 Flow Calibration Record SN 991, January 17, 2014
HD0963.45 Form A AMS-4 Flow Calibration Record SN 991, January 29, 2013
HD0963.45 Form B - AMS-4 Calibration Data Serial No. 1181, December 30, 2013
HD0963.45 Form B AMS-4 Calibration Data SN 1181, January 30, 2013
HD0963.48 Form B AMP-200 Calibration Sheet SN 7702-001, March 7, 2014
HD0963.48 Form B AMP-200 Calibration Sheet SN 7702-001, July 26, 2012
HD0963.54 Form A Fluke Biomedical 451B Ion Chamber Calibration Record SN 0038,
January 17, 2014
HD0963.54 Form A Fluke Biomedical 451B Ion Chamber Calibration Record SN 0038,
August 10, 2012
HD0963.60 Form A Chronos-4 Calibration Record SN 1009-058, March 20, 2014
HD0963.60 Form A Chronos-4 Calibration Record SN 1009-058, September 10, 2012
HD0963.61 Form A HD-29A Air Sampler Calibration Record SN 2253, March 10, 2014
HD0963.61 Form A HD-29A Air Sampler Calibration Record SN 2253, March 14, 2013
Health Physics Instruments - Product Application Literature PAL- REM-500 Calibration
Recommendation, August 2010
Health Physics Instruments- Operations and Repair Manual: Model REM-500 Neutron Survey
Meter Revision A, September 1998
HPSTD-96-004, Basis for Performing Operability Checks for Use of the Far West Technology
REM-500 Neutron Meter, February 10, 1996
HPSTID 05-007, Primary Calibration of RD-64-52 ATL Effluent Radiation Monitor RM-6473,
6509, and 6521, September 26, 2011
HPSTID 11-002 Additional Information on Containment Atmosphere Radiation Monitor Setpoint
Basis, March 9, 2011
HPSTID 13-003, Calibration of WBC Systems - 2013, March 19, 2014
HPSTID 13-008, Verification/Calibration on the Shepherd Model 81-12 (Serial No. 7015) Cs-137
Irradiator, October 28, 2013
HPSTID 14-005, 2013 10 CFR Part 61 Data, December 2013
HPSTID 14-007, Calibration of WBC Systems - 2014, March 19, 2014
SB Offsite Dose Calculation Manual, Revision 35
SB Updated Final Safety Analysis Report, Revision 8A
Seabrook Station Maintenance Rule (a)(1) Improvement Plan for Radiation Monitoring RM-03B
Subsystem Airborne Monitoring Instruments, Revision 3
Seabrook Station Radiation Protection Manual, Revision 64
Attachment
A-20
Maintenance Orders/Work Orders
40086692 40092985 40104805 40107205 40111868 40114187
40122692 40122999 40162465 40194413 40205154 40215550
40218071 40235509 40235518 40235521 40235526 40238054
40263637
Section 4OA1: Performance Indicator Verification
Procedures
CS0910.01, Primary Systems Sampling at SS-CP-166A, Revision 18
CX0901.02, Determination of Dose Equivalent I-131, Revision 12
NAP-206, NRC Performance Indicators, Revision 6
OX1401.02, RCS Steady State Leak Rate Calculation, Revision 8
Miscellaneous
LIC-13026, Documentation Supporting the Seabrook Station NRC 2nd Quarter 2013
Submittal
LIC-13037, Documentation Supporting the Seabrook Station NRC 3rd Quarter 2013
Performance Indicator Submittal
LIC-14004, Documentation Supporting the Seabrook Station NRC 4th Quarter 2013
Performance Indicator Submittal
LIC-14018, Documentation Supporting the Seabrook Station NRC 1st Quarter 2014
Performance Indicator Submittal
Technical Specification, Section 3.4.8
Section 4OA2: Problem Identification and Resolution
Condition Reports
1966262 1963207 1949610 1974681 1974682 1894548
1965182 1965183 1952682
Miscellaneous
Seabrook Station Self-Evaluation and Trending Analysis Report for 1st Quarter 2014
Section 4OA3: Follow-up of Events and Notices of Enforcement Discretion
Procedures
EN-AA-203-1001, Operability Determinations/Functionality Assessments, Revision 16
LX0563.02, Reactor Coolant Pump Undervoltage Channel Calibration and Relay PM,
Revision 11
ON1046.12, Operation of the Main Generator Breaker, Revision 21
OS1000.08, Post Trip Review, Revision 19
PI-AA-103-1001-F01, Human Performance Review Worksheet, Review 2
Condition Reports
01953543 01955993 01956944 01956945 01969615 01930049
01900249 01904703
Miscellaneous
Calculation SBC-128, Technical Specifications - Setpoints and Allowable Values, Revision 15
DBD-ESF-01, Engineered Safety Features Response Times Design Basis Document,
Revision 2
LER 2014-001-000
Attachment
A-21
Reactor Trip Due to RCP Loop Low Flow Low Maintenance Rule Functional Failure Evaluation
Work Order 40260904-03, Repair of Degraded Pipe Wall on SW-1802-4 per EC280429,
completed on April 15, 2014.
LIST OF ACRONYMS
AC alternating current
ACI American Concrete Institute
ADAMS Agencywide Document Access and Management System
ALARA as low as reasonably achievable
AR action request
ARM area radiation monitor
ASME American Society of Mechanical Engineers
BACC Boric Acid Corrosion Control Program
CAM continuous air monitor
CAP corrective action program
CFR Code of Federal Regulations
CR condition report
CRDM control rod drive mechanism
EC eddy current
EDG emergency diesel generator
EDY effective degradation years
EPD electronic personal dosimeter
EPRI Electric Power Research institute
IMC Inspection Manual Chapter
kV kilovolt
LER licensee event report
LHRA locked high radiation area
MR maintenance rule
MSHA Mine Safety and Health Administration
NCV non-cited violation
NDE nondestructive examination
NEI Nuclear Energy Institute
NIOSH National Institute for Occupational Safety and Health
NIST National Institute of Standards and Technology
NRC Nuclear Regulatory Commission
NVLAP National Voluntary Laboratory Accreditation Program
ODCM offsite dose calculation manual
ODSCC Outside Diameter Stress Corrosion Cracking
OOS out of service
OR Refueling Outage
PCCW primary component cooling water
PAB primary auxiliary building
Attachment
A-22
RAT reserve auxiliary transformer
RCP reactor coolant pump
REMP radiological environmental monitoring program
RG Regulatory Guide
RIY re-inspection year
RP radiation protection
RVCH reactor vessel upper closure head
RWP radiation work permit
SSC structure, system, or component
TS technical specification
UAT unit auxiliary transformer
UFSAR Updated Final Safety Analysis Report
UT ultrasonic testing
VHRA very high radiation area
VT visual examination
VUHP vessel upper head penetration
WBC whole body count
WO work order
Attachment