ML14127A376
| ML14127A376 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 05/06/2014 |
| From: | Glenn Dentel Reactor Projects Branch 3 |
| To: | Ossing M, Walsh K NextEra Energy Seabrook |
| Dentel G | |
| References | |
| IR 14-002 | |
| Download: ML14127A376 (37) | |
See also: IR 05000443/2014002
Text
K. Walsh
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION I
2100 RENAISSANCE BLVD., SUITE 100
KING OF PRUSSIA, PA 19406-2713
May 6, 2014
Mr. Kevin Walsh
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 05000443/2014002
Dear Mr. Walsh:
On March 31, 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 April 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.
This report documents one NRC-identified finding of very low safety significance (Green). This
finding was determined to involve a violation of NRC requirements. However, because of the
very low safety significance, and because it was entered into your corrective action program
(CAP), 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 finding in this report, you should provide
a response within 30 days of the date of this inspection report, with the basis for your
disagreement, to the Regional Administrator, Region I, and the NRC Resident Inspector at
Seabrook Station.
Additionally, as we informed you in the most recent NRC integrated inspection report, cross-
cutting aspects identified in the last six months of 2013 using the previous terminology were
being converted in accordance with the cross-reference in Inspection Manual Chapter (IMC) 0310. Section 4OA5 of the enclosed report documents the conversion of these cross-cutting
K. Walsh 2
aspects which will be evaluated for cross-cutting themes and potential substantive cross-cutting
issues in accordance with IMC 0305 starting with the 2014 mid-cycle assessment review. If you
disagree with the cross-cutting aspect assigned, you should provide a response within 30 days
of the date of this inspection report, with the basis for your disagreement, to the Regional
Administrator, Region I, and the NRC Resident Inspector at Seabrook Station.
In accordance with Title 10 of the Code of Federal Regulations (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. 05000443/2014002
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
PCataldo/GTD for per RBarkley/GTD for per
NAME GDentel/GTD
email email
DATE 05/05/14 05/06/14 05/06/14
1
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No.: 50-443
License No.: NPF-86
Report No.: 05000443/2014002
Licensee: NextEra Energy Seabrook, LLC
Facility: Seabrook Station, Unit No.1
Location: Seabrook, New Hampshire 03874
Dates: January 1, 2014 through March 31, 2014
Inspectors: P. Cataldo, Senior Resident Inspector
C. Newport, Resident Inspector
E. Burket, Emergency Preparedness Inspector
T. Burns, Reactor Inspector
B. Dionne, Health Physicist
W. Cook, Senior Reactor Analyst
Approved by: Glenn T. Dentel, Chief
Reactor Projects Branch 3
Division of Reactor Projects
Enclosure
2
TABLE OF CONTENTS
SUMMARY .................................................................................................................................... 3
REPORT DETAILS ....................................................................................................................... 4
1. REACTOR SAFETY .............................................................................................................. 4
1R01 Adverse Weather Protection ....................................................................................... 4
1R04 Equipment Alignment .................................................................................................. 4
1R05 Fire Protection ............................................................................................................. 5
1R06 Flood Protection Measures ........................................................................................ 6
1R11 Licensed Operator Requalification Program ............................................................... 6
1R12 Maintenance Effectiveness ......................................................................................... 7
1R13 Maintenance Risk Assessments and Emergent Work Control .................................. 7
1R15 Operability Determinations and Functionality Assessments ....................................... 8
1R18 Plant Modifications ................................................................................................... 10
1R19 Post-Maintenance Testing ....................................................................................... 11
1R22 Surveillance Testing ................................................................................................. 11
1EP4 Emergency Action Level and Emergency Plan Changes ......................................... 12
1EP6 Drill Evaluation ......................................................................................................... 12
2. RADIATION SAFETY .......................................................................................................... 13
2RS1 Radiological Hazard Assessment and Exposure Controls ....................................... 13
2RS2 Occupational ALARA Planning and Controls ........................................................... 16
4. OTHER ACTIVITIES ............................................................................................................ 17
4OA1 Performance Indicator Verification ........................................................................... 17
4OA2 Problem Identification and Resolution ..................................................................... 18
4OA3 Follow-Up of Events and Notices of Enforcement Discretion ................................... 22
4OA5 Other Activities .......................................................................................................... 23
4OA6 Meetings, Including Exit ............................................................................................ 24
ATTACHMENT: SUPPLEMENTARY INFORMATION................................................................ 24
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-1
LIST OF ACRONYMS ............................................................................................................. A-10
Enclosure
3
SUMMARY
IR 05000443/2014002; 01/01/2014-03/31/2014; Seabrook Station, Unit No. 1; Operability
Determinations and Functionality Assessments.
This report covered a three-month period of inspection by resident inspectors and announced
inspections performed by regional inspectors. Inspectors identified one finding of very low
safety significance (Green), which was an NCV. The significance of most findings is indicated
by their color (i.e., greater than Green, or Green, White, Yellow, Red) and determined using IMC 0609, Significance Determination Process (SDP), 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 June 7, 2012. The NRCs program for overseeing the safe operation
of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight
Process, Revision 4.
Cornerstone: Mitigating Systems
Green. The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion V,
Procedures, because NextEra did not ensure adequate separation was maintained
between temporary scaffolding and safety-related equipment. Specifically, six instances
of scaffolding installed in the plant were identified with less than the minimum standoff
distance to safety-related equipment specified in NextEra procedures and no corresponding
engineering evaluation to support these deviations. NextEra entered this NCV into their
CAP as AR 01933827 and assessed the six deviations for any impact on the associated
safety-related systems.
This performance deficiency was considered 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, NextEra did not
evaluate scaffolding installations when insufficient separation to safety-related equipment
existed after procedural requirements were revised to a more restrictive value. Additionally,
it was similar to example 4.a in IMC 0612, Appendix E, Examples of Minor Issues, which
states that the issue of failing to appropriately evaluate scaffold installation as required by
procedures is more than minor if the licensee routinely failed to perform engineering
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 involve the loss or degradation of equipment
or function specifically designed to mitigate a seismic event. This finding has a cross-cutting
aspect in the area of Problem Identification and Resolution, Evaluation, because NextEra
personnel did not perform an adequate extent of condition review after revision of their
erection of scaffold procedure. This performance deficiency directly contributed to multiple
instances of scaffold members erected within two inches of safety-related equipment without
an engineering evaluation [P.2]. (Section 1R15)
Enclosure
4
REPORT DETAILS
Summary of Plant Status
Seabrook operated essentially at full power for the entire assessment period, with the exception
of minor downpowers for turbine control valve testing. However, on March 31, 2014, plant load
was reduced to approximately 15% for turbine generator testing prior to a shutdown and entry
into refueling outage No. 16 at midnight, March 31, 2014. 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 - 1 sample)
Readiness for Impending Adverse Weather Conditions
a. Inspection Scope
The inspectors reviewed NextEras preparations for the onset of cold weather and snow
on February 5, 2014. The inspectors reviewed the implementation of adverse weather
preparation procedures before the onset of and during this adverse weather
condition. The inspectors verified that operator actions defined in NextEras adverse
weather procedure maintained the readiness of essential systems. The inspectors
discussed readiness and staff availability for adverse weather response with operations
and work control personnel.
b. Findings
No findings were identified.
1R04 Equipment Alignment
Partial System Walkdowns (71111.04Q - 3 samples)
a. Inspection Scope
The inspectors performed partial walkdowns of the following systems:
'A' emergency diesel generator (EDG) while 'B' EDG was out of service (OOS) for
annual maintenance on January 27, 2014
Supplemental emergency power system (SEPS) while 'A' EDG was OOS for annual
maintenance on February 12, 2014
'A' emergency feedwater (EFW) pump return to service on March 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 Updated Final Safety Analysis
Enclosure
5
Report (UFSAR), technical specifications (TSs), work orders (WOs), condition reports,
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 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 CAP for resolution with the appropriate significance
characterization.
b. Findings
No findings were identified.
1R05 Fire Protection
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 OOS, degraded, or
inoperable fire protection equipment, as applicable, in accordance with procedures.
Residual heat removal (RHR) containment spray safety injection (CSSI) equipment
vault train 'A' RHR-F-1B-Z, RHR-F-2B-Z, RHR-3B-Z, RHR-F-4B-Z, RHR-F-4B-Z1,
RHR-F-4B-Z2 on January 15, 2014
RHR CSSI equipment vault train 'B' RHR-F-1A-Z, RHR-F-2A-Z, RHR-3A-Z,
RHR-F-4A-Z, RHR-F-4A-Z1, RHR-F-4A-Z2 on January 16, 2014
SEPS-F-1-0 on January 21, 2014
Circulating water pump room SW-F-1A-Z on March 11, 2014
Primary auxiliary building (PAB) piping penetration area PAB-F-1A-Z, PAB-F-1J-Z
on March 17, 2014
b. Findings
No findings were identified.
Enclosure
6
1R06 Flood Protection Measures (71111.06 - 1 sample)
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 B
EDG 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.
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 January 23, 2014,
which included simulated degraded equipment and subsequent equipment failures
and initiators, which resulted in escalating degraded plant conditions that ensured
implementation of emergency operating procedures by the operating crew, as well as
implementation of the emergency plan. This emergency plan implementation included
classification of specific events that warranted an Alert Event Declaration. 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 control room supervisor.
Additionally, the inspectors assessed the ability of the crew and training staff to identify
and document crew performance problems.
b. Findings
No findings were identified.
.2 Quarterly Review of Licensed Operator Performance in the Main Control Room
a. Inspection Scope
The inspectors observed general control room activities, including alarm response and
control room shift turnovers, conducted on January 13, 2014, March 14, 2014 and
Enclosure
7
March 27, 2014. Additionally the inspectors observed turbine control valve testing on
January 17, 2014, engineered safety features actuation system (ESFAS) relay testing
on January 27, 2014, operator response to a failed open B steam generator feed
regulating bypass valve on March 14, 2014, and restoration from enclosure building
exhaust fan EAH-FN-4A testing on March 19, 2014. The inspectors observed 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, or 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.
SW pump P-41C increased vibration trending in January 2014
ED/EDE 120 VAC electrical distribution systems on February 18, 2014
b. Findings
No findings were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 5 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
Enclosure
8
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.
RHR system valve maintenance and testing on January 14, 2014
ESFAS relay testing on January 15, 2014
Inverter 1B corrective maintenance following internal transformer failure on
February 19, 2014
Planned SW cooling tower switchover on March 20, 2014
'B' feedwater regulating bypass valve M/A station replacement on March 27, 2014
b. Findings
No findings were identified.
1R15 Operability Determinations and Functionality Assessments (71111.15 - 7 samples)
a. Inspection Scope
The inspectors reviewed operability determinations for the following degraded or non-
conforming conditions:
East and west pipe chase low temperature impact on feedwater isolation valve
operability on January 3, 2014
Operability of safety-related equipment in close proximity to temporary scaffolding on
January 16, 2014
Service water pumphouse seismic monitor non-functional following monthly testing
on January 24, 2014
Containment enclosure ventilation area seal gaps identified on January 28, 2014
'A' EDG did not trip on overspeed during return to service testing on February 14, 2014
Reactor coolant system (RCS) leakage into the RHR system on March 13, 2014
Turbine-driven EFW pump P-37B oil leak identified during testing on March 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
Introduction. The inspectors identified a Green NCV of 10 CFR 50, Appendix B,
Criterion V, Procedures, because NextEra did not ensure adequate separation was
maintained between temporary scaffolding and safety-related equipment. Specifically,
Enclosure
9
six instances of scaffolding installed in the plant were identified with less than the
minimum standoff distance to safety-related equipment specified in NextEra procedures
and no corresponding engineering evaluation to support these deviations.
Description. 10 CFR 50, Appendix B, Criterion V, requires that activities affecting quality
be prescribed by documented procedures and be accomplished in accordance with
those procedures. When used in the plant, the design and installation of temporary
scaffolding must be controlled to ensure that it is not installed too close to safety-related
equipment. During a seismic event, scaffolding installed too close to safety-related
equipment can come into contact with that equipment, cause damage to it, and affect its
safety function. NextEra procedures control the installation of temporary scaffolding at
Seabrook by specifying a minimum separation between scaffolding and safety-related
equipment, and by requiring an engineering evaluation when the minimum separation
cannot be met.
NextEra mechanical maintenance procedure, MS0599.47, Erection of Scaffolding,
Revision 2, states that members of scaffolding erected adjacent to operable safety-
related equipment shall not be less than two inches unless justified by an Engineering
Evaluation. MS0599.47 was revised in February 2013, and the requirement for scaffold
separation from operable safety-related equipment was changed from ...should not be
less than 2 inches and in no case less than 1/2 inch without an engineering evaluation to
...shall not be less than 2 inches unless justified by engineering evaluation.
While performing a plant walkdown on January 15, 2014, the inspectors identified
temporary scaffold members installed less than two inches from the A Containment
Building Spray (CBS) pump discharge and suction lines. The A CBS pump and its
associated piping are classified as safety-related equipment and were operable at the
time. The identified scaffold did not include an engineering evaluation that provided
acceptance of separation of less than two inches. Subsequent plant walkdowns by
NextEra personnel identified five additional instances of scaffolding installed less than
two inches from operable safety-related equipment without an associated engineering
evaluation. NextEra personnel determined that an inadequate extent of condition review
following the February 2013 revision of MS0599.47, resulted in scaffolding being staged
in the plant at less than the new, more restrictive scaffold separation requirement of two
inches. Having identified multiple instances where NextEra personnel had not complied
with the separation requirement of the scaffolding procedure, the inspectors concluded
that NextEra had not been adequately controlling the design and installation of
temporary scaffolding.
NextEra entered the additional instances of inadequate separation identified during their
independent walkdowns into the CAP. All discrepancies were corrected and assessed
for any potential impact to the operability or functionality of affected systems. The
inspectors reviewed the CRs and determined that the safety function of each system
potentially impacted by temporary scaffolding, including those identified by the
inspectors and NextEra, would not have been affected during a seismic event.
Analysis. The inspectors determined that not providing adequate separation between
temporary scaffolding and safety-related equipment without an engineering basis was
a performance deficiency within NextEras ability to foresee and correct. Specifically,
several scaffold members were observed within two inches of safety-related equipment
without an engineering evaluation as specified by current procedural requirements.
Enclosure
10
This performance deficiency was considered more than minor because it affected the
protection against external factors attribute of the Mitigating System cornerstone and its
objective to ensure the availability, reliability, and capability of systems that respond to
initiating events to prevent undesirable consequences. Specifically, NextEra did not
evaluate scaffolding installation when insufficient separation to safety-related equipment
existed after procedural requirements were revised to a more restrictive value.
Additionally, it was similar to example 4.a in IMC 0612, Appendix E, Examples of Minor
Issues, which states that the issue of failing to appropriately evaluate scaffold
installation as required by procedures is more than minor if the licensee routinely failed
to perform engineering 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) since it did not involve the loss
or degradation of equipment or function specifically designed to mitigate a seismic event.
This finding is related to the cross-cutting area of Problem Identification and Resolution-
Evaluation, because NextEra did not thoroughly evaluate issues to ensure that
resolutions address causes and extent of conditions commensurate with their safety
significance (P.2). Specifically, NextEra personnel did not perform an adequate extent
of condition review after revision of their erection of scaffolding procedure. This
performance deficiency directly contributed to multiple instances of scaffolding members
erected within two inches of safety-related equipment without an engineering evaluation.
Enforcement. 10 CFR 50, Appendix B, Criterion V, requires, in part, that activities
affecting quality shall be prescribed by documented procedures and shall be
accomplished in accordance with those procedures. NextEra mechanical maintenance
procedure, MS0599.47, Erection of Scaffolding, Revision 2, states that members of
scaffolding erected adjacent to operable safety-related equipment shall not be less than
two inches from the equipment unless justified by an Engineering Evaluation. Contrary
to the above, on January 15, 2014, the inspectors identified that certain activities
affecting quality at Seabrook were not accomplished in accordance with documented
procedures. Specifically, following a revision of the minimum scaffolding separation
requirement in February 2013, multiple instances of scaffolding outside of the new
requirements were left uncorrected and engineering evaluations were not completed.
Installation of temporary scaffolding in the vicinity of safety-related equipment has the
potential to adversely affect that equipments performance during a seismic event
because it was installed with insufficient standoff distance. After the issue was identified
by the inspectors, NextEra performed independent walkdowns of all scaffolding,
identifying five additional instances of inadequate separation distance. All identified
discrepancies were corrected or evaluated as adequate. Because this violation is of
very low safety significance (Green) and NextEra entered this into their CAP (AR
01933827), this violation is being treated as an NCV consistent with the Enforcement
Policy. (NCV 05000443/2014002-01, Scaffolding Installed with Insufficient
Separation to Safety Related Equipment)
1R18 Plant Modifications (71111.18 - 1 sample)
Permanent Modifications
a. Inspection Scope
The inspectors evaluated a modification that replaced the SW flow element to the A
EDG jacket water cooler implemented under engineering change EC280824, and
Enclosure
11
completed on February 13, 2014. The inspectors verified that the design and licensing
bases, as well as the performance capability of the affected SW train and associated
components were not degraded by the modification. The inspectors reviewed
associated modification documents, which included topic notes, implementation work
order instructions, equivalent change revisions, applicable interface documents (for
example, drawings), and applicable post-modification testing.
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.
Thermal barrier cooling pump monthly surveillance following electrical breaker
testing on January 11, 2014
SEPS diesel generator DG-2A maintenance on January 21, 2014
Portable diesel driven pump B.5.b functional test following repairs on February 3, 2014
EDG A exhaust valve Belleville washer replacement on February 11, 2014
1B vital inverter Ferro-Resonant transformer replacement on February 19, 2014
Service water pump P-41C following shaft sleeve replacement on March 6, 2014
b. Findings
No findings were identified.
1R22 Surveillance Testing (71111.22 - 7 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:
Enclosure
12
Rod control testing and verification of proper operation of digital rod position
indication on January 9, 2014
Portable diesel driven pump B.5.b annual functional test on January 14, 2014
'A' cooling tower pump comprehensive inservice test on January 16, 2014 (IST)
RCS leak rate surveillance test on March 5, 2014 (RCS leak rate)
RCS pump seal monthly controlled leakage surveillance on March 12, 2014
Containment enclosure exhaust fan EAH-FN-4A monthly testing on March 19, 2014
'B' EDG operability surveillance and ESFAS slave relay testing March 25, 2014
b. Findings
No findings were identified.
Cornerstone: Emergency Preparedness
1EP4 Emergency Action Level and Emergency Plan Changes (71114.04 - 1 sample)
Emergency Preparedness Drill Observation
a. Inspection Scope
NextEra implemented various changes to the Seabrook Emergency Action Levels
(EALs), Emergency Plan, and Implementing Procedures. NextEra had determined that,
in accordance with 10 CFR 50.54(q)(3), any change made to the EALs, Emergency
Plan, and its lower-tier implementing procedures, had not resulted in any reduction in
effectiveness of the Plan, and that the revised Plan continued to meet the standards in
50.47(b) and the requirements of 10 CFR 50, Appendix E.
The inspectors performed an in-office review of all EAL and Emergency Plan changes
submitted by NextEra as required by 10 CFR 50.54(q)(5), including the changes to
lower-tier emergency plan implementing procedures, to evaluate for any potential
reductions in effectiveness of the Emergency Plan. This review by the inspectors was
not documented in an NRC Safety Evaluation Report and does not constitute formal
NRC approval of the changes. Therefore, these changes remain subject to future NRC
inspection in their entirety. The requirements in 10 CFR 50.54(q) were used as
reference criteria.
b. Findings
No findings were identified.
1EP6 Drill Evaluation (71114.06 - 2 samples)
.1 Emergency Preparedness Drill Observation
a. Inspection Scope
The inspectors evaluated a routine NextEra emergency drill on March 12, 2014, to
identify any weaknesses and deficiencies in the event classification and notification
Enclosure
13
activities. The inspectors observed emergency response operations in the simulator and
technical support center to determine whether the event classification and notification
activities were performed in accordance with procedures. The inspectors also attended
the individual facility drill critique to compare inspector observations with those identified
by NextEra staff, to evaluate NextEras critique and to verify whether NextEra staff was
properly identifying weaknesses and entering them into the CAP.
b. Findings
No findings were identified.
.2 Emergency Preparedness Training Observations
a. Inspection Scope
The inspectors observed a simulator training evolution for Unit 1 licensed operators on
January 23, 2014, which involved simulated emergency plan implementation by an
operations crew. NextEra planned for this evolution to be evaluated and included in
performance indicator data regarding drill and exercise performance. The inspectors
observed event classification and notification activities performed by the crew. The
inspectors also attended the post-evolution critique for the scenario. The focus of the
inspectors activities was to note any weaknesses and deficiencies in the crews
performance and ensure that NextEra evaluators noted the same issues and entered
them into the CAP.
b. Findings
No findings were identified.
2. RADIATION SAFETY
Cornerstone: Public Radiation Safety
2RS1 Radiological Hazard Assessment and Exposure Controls (71124.01)
a. Inspection Scope
From March 4 to March 8, 2014, the inspectors reviewed NextEras performance in
assessing the radiological hazards and exposure control in the workplace. 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 the NextEra procedures required by TSs, as criteria for determining
compliance.
Inspection Planning
The inspectors reviewed 2013 NextEra performance indicators for the occupational
exposure cornerstone for Seabrook Nuclear Station. The inspectors reviewed the
results of radiation protection program audits. The inspectors reviewed any reports of
operational occurrences related to occupational radiation safety since the last inspection.
Enclosure
14
Radiological Hazard Assessment
The inspectors determined if there have been changes to plant operations since the last
inspection that may result in significant new radiological hazards. The inspectors
evaluated whether NextEra assessed the potential impact of these changes and has
implemented periodic monitoring, as appropriate, to detect and quantify the radiological
hazard.
The inspectors reviewed the last two radiological surveys from the Fuel Transfer Canal
and Letdown Line in the Demineralizer Alley. The inspectors evaluated whether the
thoroughness and frequency of the surveys were appropriate for the given new
radiological hazard.
The inspectors conducted walk-downs and independent radiation measurements in the
facility, including radioactive waste processing, storage, and handling areas to evaluate
material and radiological conditions.
The inspectors reviewed one risk-significant work activity that involved exposure to
radiation. This activity was the initial entry, survey and decontamination of the fuel
transfer canal following the second dry spent fuel storage cask loading campaign. For
this work activity, the inspectors assessed whether the pre-work surveys performed were
appropriate to identify and quantify the radiological hazard and to establish adequate
protective measures. The inspectors evaluated the radiological survey program to
determine if radiological hazards were properly identified (e.g., discrete radioactive hot
particles, transuranics and hard to detect nuclides in air samples, transient dose rates
and large gradients in radiation dose rates).
The inspectors observed work in potential airborne radioactivity areas, and evaluated
whether the air samples from the fuel transfer canal air sample locations were
representative of the breathing air zone and were properly evaluated.
Instructions to Workers
The inspectors selected three containers of radioactive materials and assessed whether
the containers were labeled and controlled in accordance with 10 CFR Part 20
requirements.
The inspectors reviewed the following radiation work permits (RWP) used to access high
radiation areas (HRA) and evaluated if the specified work control instructions and control
barriers were consistent with TS requirements for HRA.
RWP 14-0015 High Integrity Container/Liner Shipping Preparation to include
Capping, Weighing and Transfer to Waste Processing Building, January 1, 2014
RWP 14-0022 Inspect CS Valves inside Letdown Valve Room at Power,
January 15, 2014
RWP 14-0027 Primary Auxiliary Building Demineralizer Alley Work/Entry,
December 31, 2013
RWP 14-0058 Fuel Storage Building Transfer Canal Radiation Protection Survey,
Decontamination and Maintenance Support Activities, March 2, 2014
Enclosure
15
For these RWPs, the inspectors assessed whether allowable stay-times or permissible
dose for radiologically significant work under each RWP were clearly identified. The
inspectors evaluated whether electronic personnel dosimeter (EPD) alarm set-points
were in conformance with survey indications and plant procedural requirements.
The inspectors reviewed two occurrences where a workers EPD malfunctioned or
alarmed. The inspectors evaluated whether workers responded appropriately. The
inspectors assessed whether the issue was included in the corrective action program
and whether compensatory dose evaluations were conducted as appropriate.
For work activities that could suddenly increase radiological conditions, the inspectors
assessed the NextEra means to inform workers of these changes.
Contamination and Radioactive Material Control
The inspectors reviewed NextEras criteria for the survey and release of potentially
contaminated material. The inspectors evaluated whether there was sufficient
procedural guidance on alarm response.
The inspectors reviewed NextEras procedures and records to verify that the radiation
detection instrumentation was used at an appropriate sensitivity level. The inspectors
selected six sealed sources from the NextEra inventory records and reviewed whether
the sources were accounted for and were tested for loose surface contamination.
The inspectors evaluated whether any recent transactions involving nationally tracked
sources were reported in accordance with 10 CFR Part 20 requirements.
Radiological Hazards Control and Work Coverage
The inspectors evaluated ambient radiological conditions and performed independent
radiation measurements during walk-downs of the facility. The inspectors assessed
whether the conditions were consistent with applicable posted surveys, RWPs, and
associated worker briefings.
The inspectors examined NextEra physical and programmatic controls for highly
activated or contaminated materials stored within spent fuel and other storage pools.
The inspectors assessed whether appropriate controls were in place to preclude
inadvertent removal of these materials from the pool.
The inspectors examined the posting and physical controls for selected HRAs, locked
high radiation area (LHRA) and very high radiation areas (VHRA) to verify conformance
with the occupational performance indicator.
Risk-Significant HRA and VHRA Controls
The inspectors discussed with the Radiation Protection Manager (RPM) the controls and
procedures for high-risk HRAs and VHRAs. The inspectors assessed whether any
changes to NextEra relevant procedures reduce the effectiveness of worker protection.
Enclosure
16
The inspectors evaluated NextEra controls for VHRAs and areas with the potential to
become a VHRA to ensure that an individual was not able to gain unauthorized access
to these areas.
Problem Identification and Resolution
The inspectors evaluated whether problems associated with radiation monitoring and
exposure control were being identified by NextEra at an appropriate threshold and were
properly addressed for resolution in the licensees corrective action program. The
inspectors assessed the appropriateness of the corrective actions for problems
documented by NextEra that involve radiation monitoring and exposure controls. The
inspectors assessed NextEra processes for applying operating experience to their plant.
b. Findings
No findings were identified.
2RS2 Occupational ALARA Planning and Controls (71124.02)
a. Inspection Scope
The inspectors assessed performance with respect to maintaining occupational
individual and collective radiation exposures as low as is reasonably achievable
(ALARA). 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.
Inspection Planning
The inspectors reviewed information regarding Seabrooks collective dose history,
current exposure trends, and ongoing or planned activities in order to assess current
performance and exposure challenges. The inspectors reviewed the plants three year
rolling average collective radiation exposure.
The inspectors compared the site-specific trends in collective exposures against the
industry average values and from similar nuclear power plants. In addition, the
inspectors reviewed any changes in the radioactive source term by reviewing the trend
in average contact dose rate with reactor coolant piping and steam generator primary
channel head space and manways. The inspectors reviewed site-specific procedures
associated with maintaining occupational exposures ALARA, which included a review of
processes used to estimate and track exposures from specific work activities.
Radiological Work Planning
The inspectors assessed whether NextEra planning identified appropriate dose
reduction techniques; considered alternate dose reduction features; and estimated
reasonable dose goals. The inspectors evaluated whether NextEras ALARA
assessment had taken into account decreased worker efficiency from use of respiratory
protective devices and/or heat stress mitigation equipment. The inspectors determined
Enclosure
17
whether NextEra work planning considered the use of remote technologies as a means
to reduce dose and the use of dose reduction insights from industry operating
experience and plant-specific lessons learned. The inspectors assessed the integration
of ALARA requirements into work procedure and RWP documents.
Source Term Reduction and Control
The inspectors used licensee records to determine the historical trends and current
status of plant source term known to contribute to elevated facility collective dose. The
inspectors assessed whether the licensee had developed contingency plans for
expected changes in the source term as the result of changes in plant fuel performance
issues or changes in plant primary chemistry.
Problem Identification and Resolution
The inspectors evaluated whether problems associated with ALARA planning and
controls are being identified by the licensee at an appropriate threshold and were
properly addressed for resolution in the licensees corrective action program. The
inspectors assessed NextEras process for applying operating experience to their plant.
b. Findings
No findings were identified.
4. OTHER ACTIVITIES
4OA1 Performance Indicator Verification (71151)
Unplanned Scrams, Unplanned Power Changes, and Unplanned Scrams with
Complications (3 samples)
a. Inspection Scope
The inspectors reviewed NextEras submittals for the following Initiating Events
Cornerstone performance indicators for the period of January 1, 2013 through
December 31, 2013.
Unplanned scrams per 7000 critical hours
Unplanned scrams with complications
Unplanned power changes per 7000 critical hours
To determine the accuracy of the performance indicator data reported during those
periods, inspectors used definitions and guidance contained in Nuclear Energy Institute
(NEI) Document 99-02, Regulatory Assessment Performance Indicator Guideline,
Revision 6. The inspectors reviewed NextEras operator narrative logs, event reports,
and NRC integrated inspection reports to validate the accuracy of the submittals.
b. Findings
No findings were identified.
Enclosure
18
4OA2 Problem Identification and Resolution (71152 - 2 samples)
.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.
b. Findings
No findings were identified.
.2 Annual Sample: Increasing Frequency of Leaks in Service Water Piping in the Vicinity of
Installation/Fabrication Welds
a. Inspection Scope
During the period January 27 to January 31, 2014, inspectors reviewed a root cause
evaluation (RCE AR 16379222) completed by NextEra staff for a service water pipe leak
that occurred in August 2013. This problem was described in a licensee event report
submitted to the NRC dated December 23, 2013. The inspectors determined the
effectiveness of actions by NextEra staff to identify, characterize, correct and prevent
reoccurrence of SW system leaks.
The inspectors assessed problem identification threshold, apparent cause analysis,
extent of condition reviews, and timeliness of corrective actions. The inspectors
reviewed documents listed in the Attachment to this report and interviewed NextEra
engineering personnel to assess the effectiveness of the planned, scheduled, and
completed corrective actions to resolve the identified deficiency.
The inspectors reviewed non-destructive test procedures, procedure qualifications
including test personnel qualifications to determine compliance with the applicable
American Society of Mechanical Engineers codes and standards. Also, the inspectors
reviewed system health reports, work orders, procurement documents, drawings and
photographs to determine if the nonconforming condition was appropriately identified,
documented, characterized and entered into NextEras corrective action process.
The inspectors reviewed root cause evaluation AR 16379222 and interviewed members
of the evaluation team. The inspectors interviewed the qualified non-destructive test
examiner to evaluate the ultrasonic test method used. Test results were reviewed with
the test examiner to assess the remaining wall thickness for continued operation without
encroaching on minimum wall requirements.
Enclosure
19
b. Findings and Observations
No findings were identified. The root cause evaluation and corrective actions were
reasonable, appropriate and timely.
NextEras root cause evaluation addressed a history of SW degradation (corrosion/
erosion) resulting in wall thinning and pressure boundary penetration and leakage. The
areas where wall thinning and leakage occurred was determined to be associated with
the loss of protective coating and/or liner failure at fabrication/installation welds which
typically results in turbulent fluid flow. This turbulent flow was particularly aggressive in
the attack of base metals and protective coatings at these weld locations and
configuration changes. The inspectors assessed the root cause determination, results of
the extent of condition investigation of other locations within the SW system and other
fluid (circulating water) systems with similar piping materials, operating parameters and
configurations.
The inspectors noted that examination using ultrasonic testing was performed at
selected locations with known change in flow patterns and velocity changes. The results
of this testing identified areas exhibiting variable wear rates. An evaluation of these test
results was made to determine pipe structural and pressure retaining integrity.
The inspectors visually examined several portions of previous SW pipe and fittings that
had been removed from the SW system in prior outages due to identified leaks. The
removed samples provided confirmatory evidence of corrosive/erosive attack from
turbulent flow at root locations of field welds, configuration changes and pipe to fitting
intersections. These locations revealed characteristic pin hole leaks at weld locations
and a general wastage of pipe and fitting interior diameters. These locations were
evaluated for compliance with minimum wall thickness requirements. Those locations
which were identified as active leaks at weld locations or, where areas exhibiting loss
of wall thickness and were encroaching on minimum wall requirements, were
dispositioned for repair/replacement in the CAP.
The inspectors determined that this issue received appropriate management attention as
indicated by the corrective action that was taken to perform a temporary leak repair by
the installation of a weldolet encapsulating the leak location. At the next outage, (OR16)
the weldolet will be removed and replaced with a more suitable flush patch. The patch
will be coated internally with a corrosion/erosion resistant material. The inspectors
discussed the licensee plans to systematically remove and replace the SW piping with a
base metal that is significantly more resistant to erosion/corrosion attack. The
inspectors examined numerous lengths of pipe and fittings which were staged for
replacement in the plant during the April 2014 refueling outage and subsequent outages.
.3 Annual Sample: Review of Activities Associated with Alkali Silica Reaction Affected
Structures
a. Inspection Scope
March 12 to 13, NRC inspectors from Region I and a structural engineer from the
Division of License Renewal, NRR, witnessed testing conducted at the Ferguson
Structural Engineering Laboratory (FSEL) at the University of Texas - Austin. The
testing was conducted in support of the Seabrook Alkali-Silica Reaction (ASR)
Enclosure
20
Integrated Corrective Action Plan. Specifically, the inspectors witnessed load testing of
the control beam for reinforcement anchorage (lap-splice) capacity. The testing was
performed in accordance with MPR Project 0326-0063, Procedure 5-7, Structural
Testing of Shear and Anchorage Specimens, Revision 1.
The inspectors also reviewed the results of the December 2013, Combined Crack
Indexing (CCI) measurements and the supporting engineering analysis. Proprietary data
sheets and associated evaluations were made available for inspector review.
Additionally, the inspectors review included discussions with the responsible Seabrook
engineers, as well as petrography specialists consulting for the University of Texas
Ferguson Structural Engineering Laboratory
Lastly, the inspectors reviewed NextEras revised Prompt Operability Determinations
(PODs) that address additional Seabrook structures identified as being affected by ASR
via the Phase 3 ASR walkdown program.
b. Findings and Observations
The inspectors identified no findings.
Review of CCI and Crack Width Measurements
The inspectors examined the December 2013 CCI measurement results documented in
Foreign Print (FP) 100847 and FP 100848, dated January 30, 2014. As documented in
these NextEra reports, there are 32 areas currently being monitored for ASR progression
using the CCI and crack width methodology. As of December 2013, 26 of the 32 areas
have been monitored on a six-month basis for approximately two years. Based upon the
data collected to date, NextEra has concluded the following: 1) the data suggests a slow
increasing trend in CCI and crack width over the past two years; 2) at 14 interior ASR
locations, the horizontal and vertical CCI data indicates an overall upward trend (an
average increase of 0.04 mm/m, with a measurement tolerance of 0.05 mm); 3) at nine
exterior locations, the horizontal and vertical CCI data indicates no significant change
over the two-year period; 4) the six floor/ceiling/roof locations indicate flat to upward
trends early in the period, but no change later in the two-year period; and, 5) some
fluctuation in the measured CCI and crack width values have been observed. The
fluctuations may be attributed to thermal effects, cyclic or constant moisture exposure,
measurement device accuracy, or the condition of the measured surfaces as impacted
by cleaning/preparation and weathering effects. Independent inspector review of CCI
and crack width measurement data and photographs of selected areas confirmed
NextEras conclusions. As stated in FP 100847 and FP 100848, NextEra will continue
the six-month data collection to comply with the Structures Monitoring Program and
validate these observed ASR progression trends.
Review of Operability Determinations
Based on the result of recent Phase 3 walkdowns, NextEra identified six additional areas
with CCI > 1.0 millimeter per meter (mm/m). In accordance with the Seabrook
Structures Monitoring Program (SMP), NextEra staff completed evaluations of the
affected structures to assess the potential impact of ASR on continued operability. The
six additional areas identified with a CCI value greater than 1.0 mm/m were: cooling
tower exterior (elevation 25, reference CTE-02); primary auxiliary building (PAB)
Enclosure
21
penetration area (elevation -26, reference MF105-01); west pipe chase (elevation 12,
reference MF202-02); and three areas in electrical manholes (below grade elevations,
reference CI-W03-Wall, CI-W05-Wall, and CI-W11-Wall). The structural evaluations
were documented as Supplement IV and V to FP 100716, Seabrook Station: Impact of
ASR on Concrete Structures and Attachments, and utilized the same design capacity
versus calculated demand margin analysis approach as the previously completed PODs
(reference Section 9.1, NRC Inspection Report 05000443/2012010). Inspector review of
the six additional ASR-affected area PODs concluded that the impacted structures have
adequate strength margin available and are fully capable of performing their safety
functions.
Control Beam Testing Observations
The inspectors witnessed the performance of load testing of the first control beam (a
specimen that has not undergone ASR aging). The beam (A-7, reinforcement
anchorage control specimen) was tested in accordance with MPR Project 0326-0063,
Procedure 5-7, Structural Testing of Shear and Anchorage Specimens, Revision 1, on
March 13, 2014. Beam failure occurred at a load of approximately 251,000 pounds, as
compared to the estimated failure load of 214,000 pounds. The failure load was slightly
higher than the estimated design capacity, but within the accuracy of the design
calculations. No test anomalies were identified. The results of the control beam test will
be used to compare subsequent ASR-affected specimen tests to evaluate the impact of
ASR degradation on structural performance. The inspectors observed proper procedural
adherence, good test coordination and proper communications and safety practices
exhibited by the testing staff, supervisory personnel and quality assurance overseers.
The inspectors verified proper testing preparations and quality control oversight as
specified by MPR Project 0326-0063, Procedure 5-6, General Preparation of Test
Facilities and Specimens, Revision 2, and MPR Procedure 0326-0062-46, Procedure
for In-process Inspections of FSEL Reinforcement Anchorage Test Setup for Seabrook
Station, Revision 0.
Initial Test Specimen ASR Expansion Results
NextEra and the UT-Austin FSEL staff have observed in the large-scale test specimens
that the X- and Y-direction deep pin expansion measurements (comparable to the
Seabrook vertical and horizontal wall surface CCI measurements) do not appear to
correlate with the through-wall (e.g., out-of-plane, or Z-direction) deep pin expansion
measurements after the initial phase of ASR expansion. X- and Y-direction expansion
appears to plateau while the Z-direction expansion continues to trend upward (increase).
All large-scale reinforcement anchorage and shear specimens have demonstrated this
expansion trend. The Z-direction expansion in the test specimens has been observed to
be 10 times greater than the X- and Y- expansions after approximately one year.
The preliminary implication of these test specimen expansion measurement trends is
that the X- and Y- expansion measurement methods (CCI and crack width) currently
used for monitoring the progression of ASR on Seabrook Station structure surfaces (per
the Structures Monitoring Program) may not provide alone, an adequate means to
monitor (1) ASR progression and (2) by inference (pending the completion of the testing
program), the ASR impact on the affected buildings structural performance. The
validation of the use of the CCI and crack width measurements for monitoring the
structural impact of ASR has been an objective of the large specimen testing program.
Enclosure
22
In considering these initial test program results, NextEra staff initiated an Action Report
(No. 01952162) to address this issue. In addition to evaluating the future impact of using
CCI to monitor ASR progression on affected structures, NextEra staff conducted a
preliminary assessment of this test data for impact to their PODs completed for ASR-
affected Seabrook structures. NextEra staff concluded this initial test program data does
not adversely impact the POD margins analyses, principally because the PODs are not
dependent upon measured CCI values for assessing the level of ASR degradation.
Instead, those structures demonstrating the most significant ASR progression (as
assumed by CCI and crack width measurements of >1.0 mm/m and >1.0 mm,
respectively) were evaluated using conservative and bounding degradation values
derived from published research and testing data developed from non-reinforced
concrete specimens. At the close of the inspection period, NextEra staff had initiated
actions to re-evaluate the use of CCI and crack width for the SMP and to re-evaluate
their methods for monitoring and assessing ASR progression of test specimens in order
to correlate test data to Seabrook Station.
NextEra staff communicated with the inspectors their plans to fabricate an additional
large-scale test specimen to instrument with strain gages and allow ASR progression in
order to validate the use of strain gages for through-wall (Z-direction) expansion
monitoring. The purpose of this activity is to validate the use of one or more strain gage
designs that can subsequently be installed in Seabrook structures to accurately monitor
through-wall expansion. In addition, NextEra staff described plans to increase the core
sampling of control and ASR-affected large scale test specimens in order to more
accurately measure ASR impact on concrete compressive and tensile strength and
modulus of elasticity. NextEra staff further communicated plans to conduct petrographic
examination of through-wall core samples from the test specimens. This additional
concrete material property testing and data collection is intended to be used to support
the correlation of testing program structural performance data to Seabrook structures
(along with additional, but not yet defined, core sampling of Seabrook ASR-affected
structures).
In summary, the inspectors concluded that the PODs completed for ASR-affected
Seabrook structures remain unaffected by the X-,Y- and Z-direction expansion data
measured, to date, in the test specimens. Actions planned by NextEra to assess the
adequacy of the SMP structural evaluation criteria and modify the ASR testing program
were viewed appropriate by the inspectors and the Seabrook ASR Issue Technical
Team, at this time. As stated above, the PODs use bounding assumptions not
dependent on the degree of expansion measured. Additional inspections are planned by
the NRC to evaluate NextEras ongoing corrective actions to resolve the non-
conformance related to ASR-affected structures at the Seabrook Station.
4OA3 Follow-Up of Events and Notices of Enforcement Discretion (71153 - 1 sample)
Plant Events
a. Inspection Scope
For the degraded plant equipment transient listed below, the inspectors reviewed and/or
observed plant parameters, reviewed personnel performance, and evaluated
performance of mitigating systems. The inspectors communicated the plant events to
Enclosure
23
appropriate regional personnel, and compared the event details with criteria contained in
IMC 0309, Reactive Inspection Decision Basis for Reactors, for consideration of
potential reactive inspection activities. The inspectors reviewed NextEras follow-up
actions related to the events to assure that NextEra implemented appropriate corrective
actions commensurate with their safety significance.
B steam generator feedwater bypass valve failed to open on March 14, 2014
b. Findings
No findings were identified.
4OA5 Other Activities
.1 Cross-Cutting Aspects
The table below provides a cross-reference from the 2013 and earlier findings and
associated cross-cutting aspects to the new cross-cutting aspects resulting from the
common language initiative. These aspects and any others identified since January
2014 will be evaluated for cross-cutting themes and potential substantive cross-cutting
issues in accordance with IMC 0305 starting with the 2014 mid-cycle assessment
review.
Finding Old Cross- New Cross-
Cutting Aspect Cutting Aspect
05000443/2013004-01, Inadequate H.1(b) H.14
Operability Determination Regarding Service
Water Leakage and Associated TS Violation
.2 Buried Piping, TI-2515/182, Phase 2 (1 sample)
a. Inspection Scope
The licensees buried piping and underground piping and tanks program was inspected
in accordance with paragraphs 03.02.a of the Temporary Instruction (TI) 2515/182, and
it was confirmed that activities which correspond to the completion dates specified in the
program, that have passed since the Phase 1 inspection was conducted, have been
completed.
The licensees buried piping and underground piping and tanks program was inspected
in accordance with paragraph 03.02.b of the TI and responses to specific questions
found in http:www.nrc.gov/reactors/operating/ops-experience/buried-pipe-ti-phase-2-
insp-req-2011-11-16.pdf were submitted to NRC headquarters staff.
b. Findings
No findings were identified.
Enclosure
24
4OA6 Meetings, Including Exit
On April 10, 2014, the inspectors presented the inspection results to Mr. Kevin Walsh,
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.
ATTACHMENT: SUPPLEMENTARY INFORMATION
Enclosure
A-1
SUPPLEMENTARY INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
K. Walsh, Site Vice President
T. Vehec, Plant General Manager
V. Brown, Senior Licensing Engineer
M. Chevalier, RP Supervisor
J. Connolly, Site Engineering Director
D. Currier, Emergency Planning Manager
K. Douglas, Maintenance Director
D. Flahardy, Radiation Protection Manager
M. Ossing, Licensing Manager
V. Pascucci, Nuclear Oversight Manager
D. Robinson, Chemistry Manager
T. Waechter, Nuclear Plant Shift Manager
LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED
Opened/Closed
05000443/2014002-01 NCV Scaffolding Installed with Insufficient Separation to
Safety Related Equipment (Section 1R15)
LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection
Procedures
ON1090.13, Response to Natural Phenomena Affecting Plant Operations, Revision 1
OS1200.03, Severe Weather Conditions, Revision 20
OS1090.09, Station Cold Weather Operations, Revision 2
Section 1R04: Equipment Alignment
Procedures
OS1026.02, Operating the DG 1A Lube Oil System, Revision 14
OS1026.03, Operating DG 1A Jacket Water Cooling System, Revision 11
OS1026.04, Operating DG 1A Starting Air System, Revision 12
OS1026.05, Operating the DG 1A Fuel Oil System, Revision 14
OS1026.06, Operating the DG 1A Air Intake, Exhaust and Vacuum System, Revision 9
OS1036.01, Aligning the Emergency Feedwater System for Automatic Operation, Revision 17
OX1461.03, SEPS Operational Readiness Status Surveillance, Revision 1
Attachment
A-2
Condition Reports
1880681 1904097 1936382 1934562 1952234
Section 1R05: Fire Protection
Procedures
Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, PAB-F-1A-Z, PAB-F-1J-Z
Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, RHR-F-1B-Z, RHR-F-2B-Z,
RHR-3B-Z, RHR-F-4B-Z, RHR-F-4B-Z1, RHR-F-4B-Z2
Seabrook Station Fire Protection Pre-Fire Strategies, Volume I, RHR-F-1A-Z, RHR-F-2A-Z,
RHR-3A-Z, RHR-F-4A-Z, RHR-F-4A-Z1, RHR-F-4A-Z2
Seabrook Station Fire Protection Pre-Fire Strategies, Volume II, SEPS-F-1-0 21-0
Seabrook Station Fire Protection Pre-Fire Strategies, Volume II, SW-F-1A-Z
Section 1R06: Flood Protection Measures
Condition Reports
00158490 01939967
Miscellaneous
Seabrook Station Moderate Energy Line Break Study
Section 1R11: Licensed Operator Requalification Program
Procedures
ER 1.1, Classification of Emergencies, Revision 53
Section 1R12: Maintenance Effectiveness
Procedures
ER-AA-201-2001, System and Program Health Reporting, Revision 4
PEG-40, Scoping Changes and Program Interfaces, Revision 5
PEG-45, Maintenance Rule Program Monitoring Activities, Revision 17
Condition Reports
1625261 1927188 1927781 1932096 1932711 1933065
1936449 1945056
Miscellaneous
ED/EDE 120 VAC System Health Report
EE-10-010, Maintenance Rule-PRA Basis Document PRA Risk Ranking and performance
Criteria based on SSPSS-2009, Revision 1
NEI-99-02, Revision 7
SW-P-41C In-Service Testing Pump Data Sheet and Data Logs
Service Water System Health Report
Section 1R13: Maintenance Risk Assessments and Emergent Work Control
Procedures
OS1046.24, Removing EDE-I-1B from Service during Power Operation, Revision 3
Miscellaneous
Maintenance Rule a(4) Risk Assessment Report for Workweek 1407-12
PRA-301, MR (a)(4) Process for On-Line Maintenance Group Instruction, Revision 0
WM-AA-100-1000, Work Activity Risk Management, Revision 0
Attachment
A-3
Maintenance Orders/Work Orders
40248951 40295189
Section 1R15: Operability Determinations and Functionality Assessments
Procedures
IX1670.905, Seismic Monitoring Data Retrieval Following a Seismic Event, Revision 6
IX1670.919, SWPH Seismic Monitor Calibration, Revision 6
MA 4.8, Control of Scaffolding, Revision 10
MA 5.7, Station Barriers, Penetration Seals, and Fire Barrier Wrap, Revision 17
MS0599.16, Construction, Repair and Rework of Silicone Base Penetration Seals, Revision 7
MS0599.47, Erection of Scaffolding, Revision 2
MX0539.50, Emergency Diesel Generator Engine 34-Month Preventative Maintenance,
Revision 6
MX0599.02, 18 Month Inspection of Technical Requirement Fire Rated Assembly Penetration
Seals, Revision 2
OS1005.05, Safety Injection System Operation, Revision 25
OX1413.01, A Train RHR Quarterly Flow and Valve Stroke Tests, and 18 Month Valve Stroke
Observation, Revision 20
OX1413.03, B Train RHR Quarterly Flow and Valve Stroke Tests, and 18 Month Valve Stroke
Observation, Revision 11
OX1426.34, Diesel Generator 1A 18 Month Operability Surveillance, Revision 10
OX1436.02, Turbine Driven Emergency Feedwater Pump Quarterly and Monthly Valve
Alignment, Revision 20
Condition Reports
0182029 0216388 0214364 0221647 1612785 1804255
1833819 1914234 1918208 1928775 1930569 1930855
1933827 1934585 1935442 1936576 1937513 1937679
1941086 1941153 1942147 1945355 1945771 1946400
1947827 1949876
Maintenance Orders/Work Orders
40153374 40191167 40199273 40228515 40245357 40250128
40291172 40291812 40291925 40292272
Miscellaneous
Calculation C-S-1-61035, Allowable CEVA Penetration Seal Opening Size, Revision 3
Colt-Pielstick PC2V Engine Vendor Manual
EE-98-019, Control of Temporary Loads in Seismic Areas
Engineering Evaluation 14-001, Scaffold & Temporary Equipment Engineering Evaluation
FP22849, Terry Turbine Instruction Manual, Revision 1
Preventive Maintenance Activity MS-CAT-12-SEAL-INSP
Penetration Seal Design, Seal No. PB-021-EV101-7502, FP4490R-01
Penetration Seal Design, Seal No. PB-021-EV101-7504, FP4492R-01
Tech Spec and Commitment Logs dated January 2-3, 2014
Technical Requirement TR21-4.3.3.3.1
Drawings
1-RH-B20662, Residual Heal Removal Sys. Train A Detail, Revision 22
Attachment
A-4
Section 1R18: Plant Modifications
Procedures
IS1672.141, SW-F-6181, DG-E-42A Jacket Water Cooler Service Water Outlet Flow Calibration,
Revision 6
EN-AA-100, Design Control Program, Revision 1
EN-AA-100-1003, Control of Design Interfaces, Revision 1
EN-AA-205-1100, Design Change Packages, Revision 9
EN-AA-205-1103, Equivalent Design Package, Revision 0
Condition Reports
1860416 1881903 1939926
Maintenance Orders/Work Orders
40233615
Miscellaneous
Specification 9763-006-174-1D, Data Sheets for Electronic Transmitters (Non-Class 1E),
Revision 14
Drawings
1-SW-D20795, Service Water System Nuclear Detail, Revision 43
Section 1R19: Post-Maintenance Testing
Procedures
LS0556.08, Routine Preventative Maintenance 7.5 KVA Westinghouse Inverter, Revision 8
LS0556.09, Replacement of Ferro-Resonant Transformers and Capacitors in Westinghouse 7.5
KVA Inverters, Revision 5
MM-AA-100, Conduct of Maintenance, Revision 4
MX0539.63, Emergency Diesel Generator Exhaust Valve Removal, Replacement, and Belleville
Washer Replacement, Revision 2
ON0443.113, Portable Diesel Driven Pump Annual Functional Test, Revision 5
OS1046.24, Removing EDE-I-1B from Service during Power Operation, Revision 2
OS1047.01, Vital Inverter Operation, Revision 14
OS1247.01, Loss of a 120VAC Instrument Panel, Revision 17
OS1412.10, Thermal Barrier Cooling Water Pump Monthly Rotation, Revision 6
OX1416.04, Service Water Quarterly Pump and Discharge Valve Test and Comprehensive
Pump Test, Revision 19
OX1446.03, Electrical Bus Weekly Operability, Revision 12
OX1456.86, Operability Testing of IST Pumps, Revision 10
SAG-9, PDDP and Hose Trailer Deployment, Revision 5
Condition Reports
0221649 1931807 1932393 1932711 1933020 1933808
1934499 1934512 1934562 1936449 1936703 1936858
1940192 1940751 1941180 1945142 1946434 1946440
1946653 1947234 1947394
Maintenance Orders/Work Orders
01168951 01207733 01210186 0305866 0320197 40196428
40228208 40239216 40259353 40290809 40295189 40298167
40298355
Attachment
A-5
Miscellaneous
Calculation C-S-1-86208, Extreme Damage Mitigating Strategy Flow Capability, Revision 3
Calculation C-S-1-50014, SW Pumps (SW-P-41A thru D) IST Uncertainties, Revision 0
Calculation 9763-3-ED-00-34-F, AC Ground Detection System
FP35465, SEPS Generator Set Technical Manual
FP500076, Godwin PDDP Instruction Manual, Revision 6
NASA TN D-8177, Apollo Experience Report-Detection and Minimization of Ignition Hazards
From Water/Glycol Contamination of Silver-Clad Electrical Circuitry
Standing Order SOO-14-001, B.5.b Pump Status, dated January 28, 2014
Westinghouse Technical Bulletin NSID-TB-87-09
Drawings
4950C70, Sheet 4, Inverter Schematic
1-NHY-310105, Sheet E02a, UPS 1-I-1B Vital Instrument Distribution Panel 1-PP-1B
1-NHY-310231, Sheet I20a, Motor/Load List Motor Control Center 1-EDE-MCC-615, Revision 7
1-NHY-310895, Sheet B4Qa, Thermal Barrier PCCW Recirc. Pump P-322B Schematic
Diagram, Revision 2
Section 1R22: Surveillance Testing
Procedures
EN-AA-205-1102, Temporary Configuration Changes, Revision 5
MA-AA-100-1011, Equipment Troubleshooting, Revision 0
ON0443.113, Portable Diesel Driven Pump Annual Functional Test, Revisions 1 and 5
ON0443.114, 18 Month B.5.b Equipment Inventory Surveillance, Revision 10
OS1001.04, RCS Unidentified Leak Rate Action Level Exceedence, Revision 0
OS1007.01, Automatic and Manual Rod Control, Revision 12
OX1408.06, Controlled Leakage Monthly Surveillance, Revision 6
OX1416.06, Service Water Cooling Tower Pumps Quarterly and 2 Year Comprehensive Test,
Revision 21
OX1423.07, Monthly Testing of Train A Enclosure Emergency Exhaust, Revision 8
OX1426.05, DG 1B Monthly Operability Surveillance, Revision 28
OX1426.19, Aligning DG 1B Controls for Auto Start, Revision 3
OX1456.46, Train B ESFAS Slave Relay K608 Quarterly Go Test, Revision 7
Condition Reports
1822620 1929096 1933872 1934512 1941467 1949526
1949825
Maintenance Orders/Work Orders
40149909 40178159 40235580 40244087 40245450 40259353
40264853 40246660 40246721 40246722 40287931 40290041
40290042
Miscellaneous
1-SW-OT-031 IST Pump Data Log
1-SW-OT-011 IST Pump Data Log
Activity 1-CP-CP-113-CRDM-1, CRDM Current Command Trace Acquisition, Revision 6
Calc 88-002, IST Calculation of Total Developed Head for Service Water and Cooling Tower
Pumps
Engineering Evaluation SS-EV-98006, Revision 1
Attachment
A-6
Drawings
1-CS-B20725, Chemical & Volume Control Sys. Seal Water Detail, Revision 20
1-NHY-310932, Cntmnt Encl Emer Exh Fan 1-FN-4A Schematic Diagram, SH-BB3a, Revision 9
1-NHY-310932, Cntmnt Encl Emer Exh Fan 1-FN-4A Legend & SW Development, SH-BB3b,
Revision 10
1-NHY-503515, EAH - Contn. Encl. Emer Exh Fltr Fan Logic Diagram, Revision 7
Section 1EP4: Emergency Action Level and Emergency Plan Changes
Procedures
ER 3.3, Emergency Operations Facility Operations, Revision 51
Section 1EP6: Drill Evaluation
Procedures
EP-AA-101-1000, Nuclear Division Drill and Exercise Procedure, Revision 5
ER 1.1, Classification of Emergencies, Revision 52
ER 1.2, Emergency Action Plan Activation, Revision 61
ER 3.1, Technical Support Center Operations, Revision 53
Condition Reports
1948051 1910629
Miscellaneous
ER 2.0B, Seabrook Station State Notification Fact Sheet, Revision 31
Form EPDP-03A, EP Cornerstone Reporting and Information Form, Revision 23
Section 2RS1: Radiation Hazard Assessment and Exposure Control
Procedures
HD0958.04, Posting of Radiologically Controlled Areas, Revision 33
HD0958.03, Personnel Survey and Decontamination Techniques, Revision 24
HD0958.13, Generation and Control of Radiation Work Permits, Revision 39
HD095817, Performance of Routine Radiological Surveys, Revision 12
HD0958.19, Evaluation of Dosimetry Abnormalities, Revision 37
HD0958.38, Evaluation of Isotopic Mix, Revision 29
HN0958.25, High Radiation Area Control, Revision 37
HN0958.30, Inventory and Control of LHRA or VHRA Keys and Locksets, Revision 26
HX0958.23, Radioactive Source Control, Revision 20
RP-AA-100-1001, RP Conduct of Ops, Revision 3
RP-AA-100-1002, Radworker Instructions and Responsibilities, Revision 1
RP-AA-101, Personnel Monitoring Program, Revision 0
RP-AA-101-1001, Radiation Protection Conduct of Operations, Revision 3
RP-AA-101-1002, Dosimetry Data Processes for Sentinel Software, Revision 3
RP-AA-101-2004, Method for Monitoring and Assigning Effective Dose Equivalent for High Dose
Gradient Work, Revision 3
RP-AA-102-1001, Area Rad Surveys, Revision 0
RP-AA-102-1000, Alpha Monitoring, Revision 0
RP-AA-102-1002, Dosimetry Data Process for Sentinel, Revision 3
RP-AA-103-1001, Posting Requirements, Revision 1
RP-AA-103-1002, High Rad Controls, Revision 1
RP-AA-107-1003, Unconditional and Conditional Release of Material, Revision 1
Attachment
A-7
Audits, Self-Assessments, and Surveillances
Seabrook 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
Assessment and ALARA Planning and Control, February 3, 2014
Seabrook Nuclear Oversight Report SBK-14-001, Radiation Protection and Radwaste
Programs, February 24, 2014
Condition Report
01836289 01855852 01898310 01903346 01906680 01934952
01940807 01941338 01943228 01945353 01945687
Miscellaneous
Seabrook Updated Final Safety Analysis Report
Seabrook Survey M-20140304-5, Initial Entry into Fuel Transfer Canal, March 4, 2014
Seabrook Survey M-20140122-3, HSYQ066A FSB-21-FB202 Quarterly, January 22, 2014
Seabrook Survey M-20140203-2, HSYQ-082B WPC-(-26)-MF-106 Quarterly, February 3, 2014
Seabrook Survey M-20130904-3, HSXQ-079A PAB-7-PB309 Quarterly, September 4, 2013
Seabrook Survey M-20131123-2, HSYQ-083A WPC-(-20)-MF102 Quarterly, November 23, 2013
Seabrook RWP 14-0015 High Integrity Container/Liner shipping Preparation to Include Capping,
Weighing and Transfer to WPB, January 1, 2014
Seabrook RWP 14-0022 Inspect CS Valves inside Letdown Valve Room at Power, January 15, 2014
Seabrook RWP 14-0027 PAB Demin Alley Work/Entry, December 31, 2013
Seabrook RWP 14-0058 FSB Transfer Canal RP Survey, Decon and Maintenance Support
Activities, March 2, 2014
Seabrook Non Exempt Source Inventory and Leak Test, September 14, 2013
Seabrook Exempt Source Index, July, 22, 2013
Seabrook Dosimetry Abnormality Occurrence Report CR 01904744, November 18, 2013
Seabrook Dosimetry Abnormality Occurrence Report CR 01906680, December 4, 2013
Seabrook 2014 Air Sample Log, March 7, 2014
Seabrook Lesson Plan HP1188C Alpha Monitoring Course, April 15, April 29 and May 13, 2013
Seabrook Log of VHRA and LHRA Access Points, March 5, 2014
Seabrook LHRA In Service Key Box Log, January 31, 2014
Seabrook LHRA/VHRA Key Issue Log, March 4, 2014
Seabrook HRA/LHRA Briefing Acknowledgement Form, March 5, 2014
Section 2RS2: Occupational ALARA Planning and Controls
Procedures
RP-AA-104 ALARA Program, Revision 2
RP-AA-104-1000, ALARA Implementing Procedure, Revision 5
Audits, Self-Assessments, and Surveillances
Seabrook 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
Assessment and ALARA Planning and Control, February 3, 2014
Seabrook Nuclear Oversight Report SBK-14-001, Radiation Protection and Radwaste
Programs, February 24, 2014
Attachment
A-8
Corrective Action Document
01836312 01843713 01856278 01867573 01872019 01883752
01890162 01893578 01896323 01904259 01930630 01944341
Miscellaneous
EPRI Standard Radiation Monitoring Program Results through OR 15, September 25, 2012
Seabrook Updated Final Safety Analysis Report
Seabrook Post Outage Critique: ALARA and Station Dose Performance, June 2013
Seabrook Station Nuclear Plant 5-Year ALARA Plan 2013-2017, July 31, 2013
Seabrook ALARA Review Board Meeting 13-04, December 11, 2013
Seabrook ALARA Review Board Meeting 14-01, March 3, 2014
Seabrook Temporary Shielding Log for OR 16, March 2014
Pre-Job ALARA Review Package: 13-01 Dry Fuel Transfer from Pool to Pad and Associated
Tasks for 8 ISFSI Casks, June 27, 2013
Post-Job ALARA Review: 13-01 Dry Fuel Transfer from Pool to Pad and Associated Tasks
for 8 ISFSI Casks, December 4, 2013
Pre-Job ALARA Review Package 14-01 OR 16 Reactor Dissassembly and Reassembly,
December 26, 2013
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 OR 16 In Service Inspection, Febuary 25, 2014
Pre-Job ALARA Review Package 14-07 OR 16 Fuel Handling Project, Febuary 25, 2014
Pre-Job ALARA Review Package 14-09 OR 16 RCP Seal Replacement, Febuary 25, 2014
Pre-Job ALARA Review Package 14-10 OR 16 Scaffolding, Febuary 25, 2014
Pre-Job ALARA Review Package 14-13 Replace Rx Ventillation Ducting Under Vessel with
New Design, Febuary 25, 2014
Post Project Critique Dry Fuel Storage, Seabrook Station 2nd Loading Campaign, October 2013:
Section 4OA1: Performance Indicator Verification
Procedures
NAP-206, NRC Performance Indicators, Revision 6
Miscellaneous
LIC-13017, Documentation Supporting the Seabrook Station NRC 1st Quarter 2013
Performance Indicator Submittal
LIC-13036, Documentation Supporting the Seabrook Station NRC 2nd Quarter 2013
Performance Indicator 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
MSPI Derivation Reports
Section 4OA2: Problem Identification and Resolution
Non-Destructive Test Reports
40265240-01, UT Extent of Condition Thickness Examination, B Train
40265234-01, UT Extent of Condition Thickness Examination, A Train
Condition Reports
01897164 01637922
Attachment
A-9
Maintenance Orders/Work Orders
40265234 40268662 40268965 40268967 94080896 94080893
Drawings
1-SW-B20794, Service Water System Nuclear Detail (Service Water Pump House)
1-SW-B20795, Service Water System Nuclear Detail (Turbine Bldg., Aux Bldg.)
SK-EC270504-2000, Installation Detail Service Water Piping Repairs
SK-EC156603-2001, Installation Detail of Weldolet Service Water Pipe Repair
SW 1802-09-EC 2080429, SW Piping Repair (Flush Patch) Line No 1-SW-1802-004
Section 4OA3: Follow-up of Events and Notices of Enforcement Discretion
Procedures
MA-AA-100-1011, Equipment Troubleshooting, Revision 0
OS1235.03, SG Level Instrument Failure, Revision 14
Condition Reports
1948268 1952067
Maintenance Orders/Work Orders
40300038
Miscellaneous
Instrument Loop Diagram ILD-1-FW-L04220, Steam Generator RC-E-11B Feedwater Bypass
Flow (Loop 2) 1-FW-L-4220, Revision 14
Operational Decision Making Bulletin, dated 3/18/2014
Section 4OA5: Other Activities
Procedures
ER-AA-102 Underground Piping and Tank Integrity Program, Revision 6
ER-AA-102-1000 Underground Piping and Tanks Integrity Examination Procedure, Revision 2
Seabrook Station Underground Piping and Tanks Inspection Program, Revision 2
SH 6.4 Dig Safe (01/06/12) Excavation of Site Locations Penetrating Plane of Ground,
Revision 13
Miscellaneous
AR 00213052-01-00, Complete Initiative Action 1 Status Complete
AR 00213052-02-00, Complete Initiative Action 2 Risk Ranking Buried Piping
AR 00213052-03-00, Complete Initiative Action 3 Develop Inspection Plan by 06/30/11
AR 00213052-04-00, Complete Initiative Action 4 Implement Inspection Plan 06/30/12
AR00213052-05-00, Develop Asset Management Plan Status Complete 01/26/10
AR00213052-06-00, Inspect Buried Piping Containing radioactive material 09/16/10
AR00213052-07-00, Underground Piping and Tanks Procedure Oversight 12/22/11
AR00213052-08-00, Prioritize Underground Piping and Tanks 06/27/12
AR 01600464 (12/09/10) Aux Steam and Aux Steam Condensate Leak w/i guard pipe
NRC Temporary Instruction 2515/182, Issue 11/17/11 and 8/8/13; Review of the Implementation
of the Industry Initiative to Control Degradation of Underground Piping and Tanks
NEI 09-14 Initial Issue, November 2009 Guideline for the Management of Underground Piping
and Tank Integrity
NEI 09-14 Guideline for the Management of Underground Piping and Tank Integrity, Revision 1,
December 2010
Attachment
A-10
NEI 09-14 Guideline for the Management of Underground Piping and Tank Integrity Inspection
and Analysis Methodologies, Revision 3
LIST OF ACRONYMS
ADAMS Agencywide Document Access and Management System
ALARA as low as reasonably achievable
CAP corrective action program
CFR Code of Federal Regulations
CSSI containment spray safety injection
EAL emergency action level
EDG emergency diesel generator
EPD electronic personal dosimeter
ESFAS engineered safety features actuation system
IMC Inspection Manual Chapter
LHRA locked high radiation area
MR Maintenance Rule
NCV non-cited violation
NEI Nuclear Energy Institute
NRC Nuclear Regulatory Commission
OOS out of service
PAB primary auxiliary building
RG Regulatory Guide
RPM Radiation Protection Manager
RWP radiation work permit
SDP significance determination process
SEPS supplemental emergency power system
SSC structure, system, or component
TI temporary instruction
TS technical specification
UFSAR Updated Final Safety Analysis Report
VAC volts alternating current
VHRA very high radiation area
WO work order
Attachment