ML13225A471
| ML13225A471 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 08/13/2013 |
| From: | Daniel Schroeder Reactor Projects Branch 1 |
| To: | Costanzo C Constellation Energy Nuclear Group |
| Schroeder D | |
| References | |
| IR-13-003 | |
| Download: ML13225A471 (73) | |
See also: IR 05000220/2013003
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION I
2100 RENAISSANCE BOULEVARD, SUITE 100
KING OF PRUSSIA, PENNSYLVANIA 19406-2713
August 13, 2013
Mr. Christopher Costanzo, Vice President
Nine Mile Point Nuclear Station
Constellation Energy Nuclear Group, LLC
P.O. Box 63
Lycoming, NY 13093
SUBJECT: NINE MILE POINT NUCLEAR STATION - NRC INTEGRATED INSPECTION
REPORT 05000220/2013003 AND 05000410/2013003
Dear Mr. Costanzo:
On June 30, 2013, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at
your Nine Mile Point Nuclear Station (NMPNS) Units 1 and 2. The enclosed inspection report
documents the inspection results, which were discussed on July 25, 2013, 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 self-revealing apparent violation concerning the improper restoration
of a direct current electrical bus which resulted in a loss of all shutdown cooling. The safety
significance of the violation is still under review pending the outcome of a Phase III risk analysis
by NRC Senior Reactor Analysts. However, the violation does not represent an immediate
safety concern because Constellation has conducted a prompt human performance event
review, entered the issue into their corrective action program (CAP), and conducted a root
cause analysis. Additionally, corrective actions including a review of all emergency, off-normal,
and normal system operating procedures are in progress. This violation with the supporting
circumstances and details is documented in this inspection report.
This report documents two NRC-identified findings and two self-revealing findings of very low
safety significance (Green). These findings were determined to involve violations of NRC
requirements. However, because of the very low safety significance, and because they are
entered into your CAP, the NRC is treating these findings as non-cited violations (NCVs)
consistent with Section 2.3.2 of the NRC Enforcement Policy. If you contest any NCVs in this
report, you should provide a response within 30 days of the date of this inspection report with
the basis of your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control
Desk, Washington D.C. 20555-0001; with copies to the Regional Administrator, Region I; the
Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555-0001; and the NRC Resident Inspector at NMPNS. In addition, if you disagree with the
C. Costanzo 2
cross-cutting aspect assigned to any finding in this report, you should provide a response within
30 days of the date of this inspection report, with the basis for your disagreement, to the
Regional Administrator, Region I, and the NRC Resident Inspector at NMPNS.
In accordance with Title 10 of the Code of Federal Regulations 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 NRC 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 Web site at http://www.nrc.gov/reading-
rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Daniel L. Schroeder, Chief
Reactor Projects Branch 1
Division of Reactor Projects
Docket Nos: 50-220 and 50-410
License Nos: DPR-63 and NPF-69
Enclosure: Inspection Report 05000220/2013003 and 05000410/2013003
w/Attachment: Supplementary Information
cc w/encl: Distribution via ListServ
Non-Sensitive Publicly Available
SUNSI Review
Sensitive Non-Publicly Available
OFFICE klm RI/DRP RI/DRP RI/DRP
NAME KKolaczyk/DLS for ARosebrook/DLS for DSchroeder/DLS
DATE 08/13/13 08/13/13 08/13/13
1
U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket Nos: 50-220 and 50-410
License Nos: DPR-63 and NPF-69
Report No: 05000220/2013003 and 05000410/2013003
Licensee: Constellation Energy Nuclear Group, LLC (CENG)
Facility: Nine Mile Point Nuclear Station, Units 1 and 2
Location: Oswego, NY
Dates: April 1, 2013 through June 30, 2013
Inspectors: K. Kolaczyk, Senior Resident Inspector
E. Miller, Resident Inspector
B. Dionne, Health Physicist
B. Haagensen, Resident Inspector
P. Kaufman, Senior Reactor Inspector
J. Krafty, Resident Inspector
J. Laughlin, Emergency Preparedness Inspector
J. Lilliendahl, Reactor Inspector
A. Rosebrook, Senior Project Engineer
B. Scrabeck, Project Engineer
Approved by: Daniel L. Schroeder, Chief
Reactor Projects Branch 1
Division of Reactor Projects
Enclosure
2
TABLE OF CONTENTS
SUMMARY.................................................................................................................................... 3
1. REACTOR SAFETY.............................................................................................................. 7
1R01 Adverse Weather Protection ................................................................................... 7
1R04 Equipment Alignment .............................................................................................. 8
1R05 Fire Protection ......................................................................................................... 9
1R07 Heat Sink Performance ........................................................................................... 9
1R08 Inservice Inspection Activities ............................................................................... 10
1R11 Licensed Operator Requalification Program & Licensed Operator Performance .. 12
1R12 Maintenance Effectiveness ................................................................................... 13
1R13 Maintenance Risk Assessments and Emergent Work Control .............................. 13
1R15 Operability Determinations and Functionality Assessments.................................. 14
1R18 Plant Modifications ................................................................................................ 15
1R19 Post-Maintenance Testing ..................................................................................... 15
1R20 Refueling and Other Outage Activities .................................................................. 16
1R22 Surveillance Testing .............................................................................................. 17
1EP4 Emergency Action Level and Emergency Plan Changes ...................................... 20
1EP6 Drill Evaluation ...................................................................................................... 20
2. RADIATION SAFETY.......................................................................................................... 21
2RS1 Radiological Hazard Assessment and Exposure Controls .................................... 21
2RS2 Occupational ALARA Planning and Controls ........................................................ 24
2RS3 In-Plant Airborne Radioactivity Control and Mitigation .......................................... 26
2RS4 Occupational Dose Assessment ........................................................................... 27
2RS7 Radiological Environmental Monitoring Program .................................................. 30
4. OTHER ACTIVITIES ........................................................................................................... 33
4OA1 Performance Indicator Verification ........................................................................ 33
4OA2 Problem Identification and Resolution ................................................................... 33
4OA3 Follow-Up of Events and Notices of Enforcement Discretion ................................ 42
4OA6 Meetings, Including Exit ........................................................................................ 50
ATTACHMENT: SUPPLEMENTARY INFORMATION .............................................................. 50
SUPPLEMENTARY INFORMATION ........................................................................................ A-1
KEY POINTS OF CONTACT .................................................................................................... A-1
LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED ..................................... A-2
LIST OF DOCUMENTS REVIEWED ........................................................................................ A-3
LIST OF ACRONYMS............................................................................................................. A-19
Enclosure
3
SUMMARY
IR 05000220/2013003, 05000410/2013003; 04/01/2013 - 06/30/2013; Nine Mile Point Nuclear
Station (NMPNS) Units 1 and 2; Surveillance Testing, Problem Identification and Resolution,
Follow-Up of Events and Notices of Enforcement Discretion.
This report covered a 3-month period of inspection by resident inspectors and announced
inspections performed by regional inspectors. One apparent violation was identified. The
safety significance of this violation is still under review pending the outcome of a Phase III risk
analysis by NRC Senior Reactor Analysts. Additionally, two NRC-identified findings and two
self-revealing findings of very low safety significance (Green) were identified, all of which were
non-cited violations (NCVs). 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 (SDP), dated June 2, 2011. Cross-
cutting aspects are determined using IMC 0310, Components Within the Cross-Cutting Areas,
dated October 28, 2011. All violations of NRC requirements are dispositioned in accordance
with the NRCs Enforcement Policy, dated January 28, 2013. The NRCs program for
overseeing the safe operation of commercial nuclear power reactors is described in NUREG-
1649, Reactor Oversight Process, Revision 4.
Cornerstone: Initiating Events
TBD. A self-revealing apparent violation of Technical Specification (TS) 6.4.1, Procedures,
was identified at Unit 1 because CENG failed to properly recover from a loss of a vital direct
current (DC) bus in accordance with station off-normal procedures resulting in an unplanned
loss of all shutdown cooling (SDC) when time to boil was less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Specifically,
during the restoration from the loss of battery bus 12, operators failed to identify a SDC trip
signal before attempting restoration of the DC bus, which ultimately lead to a SDC pump trip
(i.e. loss of decay heat removal from the reactor). Corrective actions included conducting a
prompt human performance event review, entering the issue into their corrective action
program (CAP), and conducting a root cause analysis. Planned corrective actions include a
review of all emergency, off-normal, and normal system operating procedures.
The inspectors determined that CENGs failure to properly restore battery bus 12 in
accordance with N1-SOP-47A.1, Loss of DC, Revision 00101, and N1-OP-47A, 125 VDC
Power System, Revision 02500, was a performance deficiency that was reasonably within
CENGs ability to foresee and correct and should have been prevented. The performance
deficiency was determined to be more than minor because the inspectors determined it
affected the configuration control aspect of the Initiating Events cornerstone and adversely
affected the associated 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 significance of the finding is designated as To Be Determined (TBD) until a Phase 3
analysis can be completed by the NRCs Senior Reactor Analysts. The inspectors
determined this finding has a cross-cutting aspect in the area of Human Performance,
Resources, because CENG did not ensure that personnel, equipment, procedures, and
other resources were available and adequate to assure nuclear safety - complete, accurate
Enclosure
4
and up-to-date design documentation, procedures, work packages, and correct labeling of
components. Specifically, CENG procedures N1-SOP-47A.1 and N1-OP-47A did not
contain adequate guidance to ensure recovery from a loss of a DC bus would not result in
an unexpected plant transient H.2(c). (Section 4OA3)
Cornerstone: Mitigating Systems
Green. A self-revealing NCV of TS 5.4.1, Procedures, was identified at Unit 2 when a
CENG instrumentation and control (I&C) technician did not properly implement procedure
N2-ISP-LDS-Q010, Reactor Building General Area Temperature Instrument Channel
Functional Test, Revision 00102. As a result, a residual heat removal (RHR)/reactor core
isolation cooling (RCIC) isolation bypass switch was inadvertently left in the NORMAL
position during surveillance testing resulting in an unplanned RCIC isolation. CENG entered
this issue into their CAP as CR-2013-002461. Other corrective actions included performing
a human performance stand down that reinforced use of human performance tools and the
need to identify and mark critical steps during pre-job briefs, retraining the I&C technicians
involved in the event on proper use of human performance error prevention techniques, and
improving bypass switch verification steps for procedure N2-ISP-LDS-Q010 and other
similar lead detection system surveillances procedures.
This finding is more than minor because it is associated with the human performance
attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to
ensure the availability, reliability, and capability of systems that respond to initiating events
to prevent undesirable consequences (i.e., core damage). Specifically, the inadvertent
isolation rendered the RCIC system inoperable and unable to perform its function for
approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Additionally, this finding is similar to example 4.b of IMC 0612,
Appendix E, Examples of Minor issues, and is more than minor due to the procedural error
leading to a plant transient, i.e. an unplanned RCIC isolation. This finding was evaluated in
accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC
0609, Appendix A, The Significance Determination Process for Findings At-Power, issued
June 19, 2012. Unit 2 is a boiling-water reactor (BWR)-5, and as a result, RCIC is treated
as having a separate high-pressure injection safety function. A detailed analysis was
conducted using SAPHIRE version 8.0.8.0 and Unit 2 SPAR model 8.17. Using an
exposure period of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and conservatively assuming no recovery of the failed
equipment, this finding had a change in core damage frequency of low E-8. The dominant
accident sequence was a grid-related loss of offsite power with a failure of Division III power
and the failure to recover offsite power and the emergency diesel generators (EDGs) in 30
minutes. Since the change in core damage frequency was less than 1E-7, contributions
from large early release and external event did not need to be considered. Therefore, this
finding was of very low safety significance (Green). This finding has a cross-cutting aspect
in the area of Human Performance, Work Practices, because the I&C technicians did not
effectively employ self-checking and place-keeping when implementing the test procedure
which directly contributed to the resulting procedural error H.4(a). (Section 1R22)
Green. The inspectors identified an NCV at Unit 2 of Title 10 of the Code of Federal
Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and
Drawings, because CENG did not assure that the replacement of cells in battery 2C were
prescribed and performed by appropriate procedures which resulted in degraded accuracy
Enclosure
5
of test results and potential degradation of safety-related battery cells. In response to this
issue, CENG generated CR-2013-005235 and initiated actions to evaluate replacing the
new cells.
This finding is more than minor because it was associated with the equipment performance
attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to
ensure the availability, reliability, and capability of systems that respond to initiating events
to prevent undesirable consequences. In accordance with IMC 0609.04, Initial
Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance
Determination Process for Findings At-Power, issued June 19, 2012, the inspectors
determined this finding is of very low safety significance (Green) because the performance
deficiency was not a design or qualification deficiency, did not involve an actual loss of
safety function, did not represent actual loss of a safety function of a single train for greater
than its TS allowed outage time, and did not screen as potentially risk-significant due to a
seismic, flooding, or severe weather initiating event. This finding has a cross-cutting aspect
in the area of Human Performance, Decision-Making component, because CENG did not
use conservative assumptions in decision making. Specifically, CENG did not monitor the
cells in storage, question the adequacy of the discharged cells, charge the cells prior to
installation, or fully evaluate the implications of the test and recharge results H.1(b).
(Section 4OA2)
Green. The inspectors identified an NCV at Unit 2 of 10 CFR Part 50, Appendix B,
Criterion III, Design Control, because CENG did not verify the adequacy of the design with
respect to battery 2C. Specifically, by failing to size the battery to the most limiting time
period, the sizing calculation significantly overstated the available design margin. CENGs
corrective actions included generating condition report CR-2013-005117 and evaluating the
condition for operability.
This finding is more than minor because it was associated with the design control attribute of
the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the
availability, reliability, and capability of systems that respond to initiating events to prevent
undesirable consequences. In accordance with IMC 0609.04, Initial Characterization of
Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process
for Findings At-Power, issued June 19, 2012, the inspectors determined this finding is of
very low safety significance (Green) because the performance deficiency was not a design
or qualification deficiency, did not involve an actual loss of safety function, did not represent
actual loss of a safety function of a single train for greater than its TS allowed outage time,
and did not screen as potentially risk-significant due to a seismic, flooding, or severe
weather initiating event. The inspectors did not assign a cross-cutting aspect because the
finding was not indicative of current performance. (Section 4OA2)
Cornerstone: Barrier Integrity
Green. A self-revealing NCV of TS 3.3.3, Leakage Rate, was identified for CENGs failure
from December 3 to December 13, 2012, to maintain containment leakage less than
1.5 percent by weight of the containment air per day and less than 0.6 percent by weight of
the containment air per day for all penetrations and all primary containment isolation valves
subject to 10 CFR Part 50, Appendix J, Types B and C tests, when pressurized to
Enclosure
6
35 pound per square inch gauge when reactor coolant system (RCS) temperature is above
215°F and primary containment integrity is required. CENG entered this issue into their
CAP as CR-2012-011247. Corrective actions included cleaning iron oxide from the primary
containment vent and purge valve and replacing the resilient seals.
This finding is more than minor because it is associated with the structure, system,
component (SSC), and barrier performance attribute of the Barrier Integrity cornerstone and
affected the cornerstone objective to provide reasonable assurance that physical design
barriers (fuel cladding, reactor coolant system, and containment) protect the public from
radionuclide releases caused by accidents or events. Specifically, containment leakage
exceeded the leakage limits outlined in the Unit 1 TS 3.3.3 from December 3 to December
13, 2012. This finding was evaluated in accordance with IMC 0609.04, Initial
Characterization of Findings, and Table 6.2, Phase 2 Risk Significance-Type B Findings at
Full Power, of IMC 0609, Appendix H, Containment Integrity Significance Determination
Process, issued May 6, 2004. The inspectors determined this finding was of very low
safety significance (Green) because the leakage was less than 100 percent of containment
volume per day for the duration of the leak. This finding has a cross-cutting aspect in the
area of Problem Identification and Resolution, CAP, because CENG failed to take
appropriate corrective action to address safety issues and adverse trends in a timely
manner commensurate with their safety significance. Specifically, following identification of
the adverse trend regarding the frequency of nitrogen addition to the drywell, CENG did not
assess in a timely manner the significance of the leakage and the impact on primary plant
containment P.1(d). (Section 4OA3)
Enclosure
7
REPORT DETAILS
Summary of Plant Status
Unit 1 began the inspection period at 100 percent power. On April 14, 2013, Unit 1 reduced
power to 32 percent to conduct emergency condenser testing and to down power for refueling
outage (N1R22). On April 15, Unit 1 was removed from the grid to commence N1R22. Unit 1
returned to service and synchronized to the grid on May 15. On June 21, Unit 1 down powered
to 83 percent to perform a rod pattern adjustment, turbine stop valve replacement, and a reactor
recirculation pump swap. Unit 1 returned to rated power on June 22 and remained at or near
full power for the remainder of the inspection period.
Unit 2 began the inspection period at 100 percent power. On May 28, Unit 2 down powered to
65 percent to investigate diverging feedwater flows between two operating feedwater pumps.
Following identification of a degraded automatic feedwater regulating valve and removal of the
B feedwater pump from service, Unit 2 returned to 100 percent on May 31, and remained at or
near full power for the remainder of the inspection period.
1. REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection (71111.01 - 2 samples)
.1 Readiness for Seasonal Extreme Weather Conditions
a. Inspection Scope
The inspectors performed a review of CENGs readiness for the onset of seasonal high
temperatures. The review focused on Unit 1 fire protection and diesel fire pump,
technical support center ventilation, control room and reactor building (RB) air
conditioning systems, and Unit 2 service water and heating, ventilation, and air
conditioning systems. The inspectors reviewed the Updated Final Safety Analysis
Report (UFSAR), TSs, and the CAP to determine what temperatures or other seasonal
weather could challenge these systems and to ensure CENG personnel had adequately
prepared for these challenges. The inspectors reviewed station procedures including
CENGs seasonal weather readiness 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. Documents reviewed for each section of this inspection report are
listed in the Attachment.
b. Findings
No findings were identified.
Enclosure
8
.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
changes to CENGs procedures affecting these areas and the communications protocols
between the transmission system operator and CENG implemented since the previous
sample in 2012. This review focused on changes to the established program and
material condition of the offsite and alternate AC power equipment. The inspectors
assessed whether CENG established and implemented appropriate procedures and
protocols to monitor and maintain availability and reliability 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 season readiness coordinator,
reviewing condition reports and open work orders and walking down portions of the
offsite and AC power systems including the 345 kilovolt (kV) and 115 kV switchyards.
b. Findings
No findings were identified.
1R04 Equipment Alignment
Partial System Walkdown (71111.04Q - 5 samples)
a. Inspection Scope
The inspectors performed partial walkdowns of the following systems:
Unit 1, Spent fuel pool (SFP) cooling system during the conduct of refueling
maintenance related activities on April 15, 2013
Unit 1, Core sprays 112 and 122 following the completion of surveillance activities on
April 21, 2013
Unit 1, Isolation condenser loop 12 following the completion of maintenance activities
on May 15, 2013
Unit 1, Diesel and electric fire pumps while the maintenance fire pump was operating
with a degraded discharge relief valve on May 22, 2013
Unit 1, Control room emergency ventilation system following the completion of
maintenance activities on May 30, 2013
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,
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
Enclosure
9
that there were no deficiencies. The inspectors also reviewed whether CENG 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 - 6 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
CENG 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, degraded, or
inoperable fire protection equipment, as applicable, in accordance with procedures.
Unit 1, Drywell (FA3/R1) on April 16, 2013
Unit 1, RB elevation 340 feet (FA1/R6A and FA2/R6B) on April 19, 2013
Unit 1, RB elevation 198 feet southwest (FA2/R1B) on April 21, 2013
Unit 1, RB elevation 237 feet east (FA1/R1A) on April 21, 2013
Unit 1, RB elevation 237 feet west (FA2/R1B) on April 21, 2013
Unit 1, Power board 12 (FA-17A) on April 26, 2013
b. Findings
No findings were identified.
1R07 Heat Sink Performance (71111.07 - 2 samples)
a. Inspection Scope
The inspectors reviewed the samples listed below to determine their readiness and
availability to perform their safety functions. The inspectors reviewed the design basis
for the components and verified CENGs commitments to NRC Generic Letter 89-13.
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
CENG initiated appropriate corrective actions for identified deficiencies.
Unit 1, Emergency diesel generator (EDG) 103 raw water heat exchanger on
May 3, 2013
Unit 2, 2HVY*UC2A service water pump bay A unit cooler on May 7, 2013
Enclosure
10
1R08 Inservice Inspection Activities (71111.08 - 1 sample)
a. Inspection Scope
From April 15 to 18, 2013, the inspectors conducted a review of CENGs implementation
of inservice inspection (ISI) program activities for monitoring degradation of the RCS
boundary and risk-significant piping system boundaries for Unit 1 during the N1R22.
The sample selection was based on the inspection procedure objectives and risk priority
of those components and systems where degradation would result in a significant
increase in risk of core damage. The inspectors observed in-process nondestructive
examinations (NDEs), reviewed documentation, and interviewed CENG personnel to
verify that the NDE activities performed were conducted in accordance with the
requirements of the American Society of Mechanical Engineers (ASME) Boiler and
Pressure Vessel Code,Section XI, 2004 Edition.
NDE Activities and Welding Activities
The inspectors performed direct observations of NDE activities in process and reviewed
records of NDEs listed below:
ASME Code Required Examinations
Remote visual examination (VT-3) of reactor vessel nozzle N16-1-N3A and manual
ultrasonic testing (UT) examination of three 12-inch diameter emergency condenser
supply piping welds.
Data records of manual UT phased array examination of five 28-inch diameter
reactor vessel nozzle-to-vessel dissimilar metal safe end-to-nozzle welds (32-WD-
042, N2A; 32-WD-082, N2B; 32-WD-122, N2C; 32-WD-164, N2D; 32-WD-208, N2E),
manual UT of four 12-inch diameter emergency condenser supply piping welds, dye
penetrant testing and UT of branch connection-decontamination port welds on the
recirculation system suction piping, and UT thickness readings of various diameter
RB closed loop cooling system piping located at elevation 225 foot in the drywell.
The inspectors reviewed certifications of the NDE technician, process, and equipment in
identifying the condition or degradation of risk-significant SSCs and evaluated the
activities for compliance with the requirements of Unit 1s risk informed ISI program,
ASME Boiler and Pressure Vessel Code,Section XI, and 10 CFR 50.55a.
Augmented or Industry Imitative Examinations
Based on industry operating experience, the inspectors reviewed NDE data records of
the recirculation system suction piping decontamination port branch connection welds to
verify that the activities were performed in accordance with applicable examination
procedures and industry guidance.
Modification/Repair/Replacement Consisting of Welding Activities
The inspectors reviewed the following welding activities to verify specifications and
control of the welding processes, weld procedures, welder qualifications, and NDE
examinations were in accordance with ASME code requirements.
Enclosure
11
The repair and replacement of reactor water cleanup (RWCU) dissimilar metal pipe weld
33-WD-046 was reviewed. The inspectors reviewed the associated flaw evaluation,
NDE data records, and repair/replacement WO package.
During manual phased array UT of a 6-inch diameter schedule 80 stainless steel pipe to
carbon steel RWCU pipe dissimilar metal weld, a 4.25-inch long circumferential flaw
indication was detected in the heat-affected zone of the stainless steel side of the weld.
The indication did not meet ASME Code,Section XI 2004, IWB-3514-2 acceptance
criteria so a flaw evaluation was required. The flaw evaluation concluded that sufficient
structural margin was demonstrated for the as-found flaw indication.
However, a review of construction radiographs by CENG indicated that there had been
two previous weld repairs directly adjacent to this indication. CENG determined that the
residual stresses of the weld were likely to be high due to the prior weld repairs and the
crack growth rate would be high enough to possibly propagate the flaw beyond the
ASME code limit of through-thickness. Based on this information, CENG replaced the
weld and adjacent pipe by installing a new spool piece.
The inspectors verified the welding activities and applicable NDE techniques were
performed in accordance with ASME Code requirements.
Re-examination of an Indication Previously Accepted for Service After Analysis
There were no samples available for review during this inspection that involved
examinations with recordable indications that have been accepted for continued service
from the previous Unit 1 outage through the current outage.
Drywell Visual Examination
The inspectors examined the condition of Unit 1 drywell liner surface at various elevation
levels inside the drywell. During the inspection, surface corrosion was noted on the
drywell liner and on several systems including the RB closed-cooling water system.
CENG was monitoring the condition of the liner and RB closed-cooling water system to
ensure the corrosion was not impacting system or component operability.
Identification and Resolution of Problems
The inspectors reviewed a sample of condition reports which involved ISI-related
activities to confirm that non-conformances were being properly identified, reported, and
resolved.
b. Findings
No findings were identified.
Enclosure
12
1R11 Licensed Operator Requalification Program and Licensed Operator Performance
(71111.11Q - 4 samples)
.1 Quarterly Review of Licensed Operator Requalification Testing and Training (2 samples)
a. Inspection Scope
The inspectors observed:
Unit 1, Licensed operator simulator training which included a loss of condenser
vacuum, a stuck open electro-matic relief valve (ERV), and an anticipated transient
without scram on April 2, 2013
Unit 2, Licensed operator performance during a simulator training scenario that
included high temperatures on the main transformer, degraded service water, and a
loss of the offsite electrical grid on May 23, 2013
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 classifications 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.
b. Findings
No findings were identified.
.2 Quarterly Review of Licensed Operator Performance in the Main Control Room
(2 samples)
a. Inspection Scope
The inspectors observed:
Unit 2, Control room operations during a period of increased site activity due to a
failure of an on-site power loop that supplied electrical power to several non-
essential buildings within the protected area as well as several plant information
technology systems on April 9, 2013
Unit 1, Control room operations during a plant shutdown to commence planned
refueling outage N1R22 on April 14, 2013
The inspectors reviewed CNG-OP-1.01-1000, Conduct of Operations, Revision 00900,
and verified that procedure use, crew communications, and coordination of plant
activities among work groups similarly met established expectations and standards.
Additionally, 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.
Enclosure
13
b. Findings
No findings were identified.
1R12 Maintenance Effectiveness (71111.12Q - 4 samples)
a. Inspection Scope
The inspectors reviewed the samples listed below to assess the effectiveness of
maintenance activities on SSC performance and reliability. The inspectors reviewed
system health reports, CAP documents, maintenance work orders, and maintenance
rule basis documents to ensure that CENG was identifying and properly evaluating
performance problems within the scope of the maintenance rule. For each sample
selected, the inspectors verified that the SSC was properly scoped into the maintenance
rule in accordance with 10 CFR 50.65 and verified that the (a)(2) performance criteria
established by CENG 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 CENG staff was identifying and
addressing common cause failures that occurred within and across maintenance rule
system boundaries.
Unit 1, Neutron monitoring on May 14, 2013
Unit 2, High-pressure core spray (HPCS) on May 14, 2013
Unit 1, Service water on May 16, 2013
Unit 1, Containment spray on May 17, 2013
b. Findings
No findings were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 8 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 CENG performed
the appropriate risk assessments prior to removing equipment from service. 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
CENG personnel performed risk assessments as required by 10 CFR 50.65(a)(4) and
that the assessments were accurate and complete. When CENG 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.
Enclosure
14
Unit 2, Unplanned elevated risk condition that resulted from an inadvertent isolation
of the RCIC system on April 2, 2013
Unit 2, Loss of maintenance supply power to 2VBB*UPS3B on April 5, 2013
Unit 1, Power boards 102 and 16 following electrical realignment on May 1, 2013
Unit 1, Planned maintenance on pressure safety valve 201.970, emergency
condenser vent isolation IV-05-03, and emergency condenser 112 HX HTX-60-44 on
May 2, 2013
Unit 2, Planned maintenance on the Division I control room air conditioning system
on May 13, 2013
Unit 1, Unplanned maintenance on the turbine bypass valve control system on
May 14, 2013
Unit 1, Planned maintenance on the 102 EDG raw water pump on May 23, 2013
Unit 2, Unplanned maintenance on the 2SWP*P1B service water pump on June 11,
2013
b. Findings
No findings were identified.
1R15 Operability Determinations and Functionality Assessments (71111.15 - 9 samples)
a. Inspection Scope
The inspectors reviewed operability determinations for the following degraded or non-
conforming conditions:
Unit 1, Acceptance criteria associated with N1-ST-C5, secondary containment, and
RB emergency ventilation system operability testing on April 13, 2013
Unit 1, Emergency service water 11 pump (72-04) trip during surveillance testing on
April 17, 2013
Unit 1, Damaged containment spray nozzle deflectors on May 3, 2013
Unit 1, Source range monitors due to under-vessel work on May 3, 2013
Unit 1, Steam leakage from vent valve 05-11 on May 19, 2013
Unit 2, RCIC high-energy line break barrier door on May 20, 2013
Unit 1, Core spray topping pump 122 bearing cooling water flow on June 11, 2013
Unit 2, Elevated drywell floor drain leakage on June 11, 2013
Unit 1, Elevated drywell floor drain leakage on June 25, 2013
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 CENGs 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 CENG. The
inspectors determined, where appropriate, compliance with bounding limitations
associated with the evaluations.
Enclosure
15
b. Findings
No findings were identified.
1R18 Plant Modifications (71111.18 - 3 samples)
.1 Temporary Modifications (1 sample)
a. Inspection Scope
The inspectors reviewed a temporary change to ventilation damper 2HVP*AOD5A which
supplies outside air to the Division III diesel generator room. The inspectors reviewed
10 CFR 50.59 documentation and conducted a field walkdown of the modification to
verify that the temporary modification did not degrade the design bases, licensing bases,
and performance capability of the affected systems.
b. Findings
No findings were identified.
.2 Permanent Modifications (2 samples)
a. Inspection Scope
The inspectors evaluated the following modifications:
Engineering Change Package (ECP) 12-00616 - Installation of a damper for Unit 1
downstream of BV-210-25
ECP 13-000167 - Installation of replacement pump for Unit 1 service water radiation
monitor
The inspectors verified that the design bases, licensing bases, and performance
capability of the affected system was not degraded by the modifications. In addition, the
inspectors reviewed modification documents associated with the upgrade and design
change including the post-installation test procedure, the 10 CFR 50.59 screening form,
and the operational impact assessment form.
b. Findings
No findings were identified.
1R19 Post-Maintenance Testing (71111.19 - 5 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
Enclosure
16
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.
Unit 1, Control room ventilation/smoke purge system test following installation of fire
damper BV-21-036 on April 3, 2013
Unit 1, Power board 102 following National Fire Protection Act 805 modification on
April 28, 2013
Unit 1, Isolation valve IV-39-10R following control circuit stop relay replacement on
May 9, 2013
Unit 1, Replacement of excess flow check valve CKV-32-138 on May 10, 2013
Unit 1, IV-29-07R diagnostic testing following body-to-bonnet seal replacement on
May 23, 2013
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 Unit 1
maintenance and refueling outage (N1R22) which was conducted April 14 through May
15, 2013. The inspectors reviewed CENGs 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 portions of the shutdown and cooldown processes and monitored controls
associated with the following outage activities:
Configuration management, including maintenance of defense-in-depth,
commensurate with the outage plan for the key safety functions and compliance with
the applicable TSs 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
Status and configuration of electrical systems and switchyard activities to ensure that
TSs were met
Monitoring of decay heat removal operations
Impact of outage work on the ability of the operators to operate the SFP 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
Maintenance of secondary containment as required by TSs
Refueling activities
Fatigue management
Enclosure
17
Tracking of startup prerequisites, walkdown of the drywell (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
Identification and resolution of problems related to refueling activities
b. Findings
No findings were identified.
1R22 Surveillance Testing (71111.22 - 8 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 CENG 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:
N1-ST-Q3, Unit 1, High-Pressure Coolant Injection Pump and Check Valve
Operability Test for Train 12 on April 1, 2013
N1-ST-C5, Unit 1, Secondary Containment and Reactor Building Emergency
Ventilation System Operability Test for Loop 11 on April 8, 2013
N1-ISP-LRT-TYC, Unit 1, Local Leak Rate Test for Valves IV-201-09 and IV-201-10
on April 9, 2013
N2-ISP-LDS-Q010, Unit 2, Reactor Building General Area Temperature Instrument
Channel Functional Test on April 18, 2013
Unit 2, RCS Leakage Determination Surveillance and Calculations on April 24, 2013
N2-CSP-GEN-D100, Unit 2, Reactor Water/Auxiliary Water Chemistry Surveillance
on April 24, 2013
N1-TSP-201-001, Unit 1, Integrated Leak Rate Test of Primary Containment Type A
Test on May 8, 2013
N1-ST-Q15, Unit 1, Condensate Transfer System Operability Test on May 30, 2013
b. Findings
Introduction. A self-revealing Green NCV of TS 5.4.1, Procedures, was identified at
Unit 2 when a CENG I&C technician did not properly implement procedure N2-ISP-LDS-
Q010, Reactor Building General Area Temperature Instrument Channel Functional
Test, Revision 00102. As a result, a RHR/RCIC isolation bypass switch was
inadvertently left in the NORMAL position during surveillance testing resulting in an
unplanned RCIC isolation.
Description. The RCIC system is designed to provide adequate makeup water to the
reactor pressure vessel (RPV) automatically or manually following an RPV isolation
accompanied by a loss of coolant flow from the feedwater system. In the event the
Enclosure
18
steam piping to the RCIC pump system leaks, temperature sensors in the RCIC pump
room will close isolation valves in the RCIC system stopping the leak. CENG
surveillance procedure N2-ISP-LDS-Q010 is a TS surveillance test that verifies that the
group 5 (RHR) and group 10 (RCIC) isolation trip signals will close the respective
system isolation valves if a high-temperature condition occurs. The procedure tests this
function by simulating a high temperature condition and verifying correct system
response. Actual valve movement during testing is prevented by control room operators
blocking the test signal.
On April 2, 2013, an unplanned RCIC isolation occurred when I&C technicians did not
properly implement procedure N2-ISP-LDS-Q010 to block the test signal. Specifically,
step 7.2.1 required I&C technicians to request control room operators to place channel
bypass switch E31A-S4B RHR/RCIC ISOLATION BYPASS in BYPASS and to verify the
circuit was bypassed by observing annunciator and plant computer alarms prior to lifting
thermocouple leads in the field. This was not accomplished which resulted in the
isolation of the RCIC system.
Prior to the event, a pre-job brief was conducted by CENG I&C technicians performing
the work which focused on the roles and responsibilities of personnel including the lifting
of thermocouple leads safely and error free. Placing the RHR/RCIC isolation bypass
switch in BYPASS was not identified as a critical step, and no critical steps were
annotated in the work document as required by CNG-PR-1.01-1009, Procedure and
Work Order Use and Adherence Requirements, Revision 00701. However, the
requirement for operations personnel to place the isolation switch in BYPASS was
discussed during the procedure review with the control room supervisor who assigned a
control room operator to perform the task when requested by I&C technicians. Section
3.12 of CNG-PR-1.01-1009 defines place-keeping as physically marking steps to
prevent the omission or duplication of the steps to maintain an accounting of steps in
progress, steps completed, steps not applicable, and steps not yet performed. It lists
among high-risk practices to be avoided by signing or checking off a step as completed
before it is completed. After commencing surveillance procedure N2-ISP-LDS-Q010,
technicians used improper self-checking and place-keeping by checking and initialing as
complete step 7.2.1 to request operators to place the RHR/RCIC isolation bypass switch
in BYPASS and to verify annunciator and computer alarm points were in alarm without
that step having been performed. Consequently, when thermocouple leads were lifted in
the following step, a false high-temperature signal was generated resulting in the closing
of RCIC steam supply isolation valves 2ICS*MOV121, 2ICS*MOV128, 2ICS*MOV170,
and an unplanned isolation of RCIC. The surveillance test was immediately stopped, the
required TS action statements were entered for the RCIC system, and the system was
restored to an operable status after approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The isolation signal was also
sent to the RHR system for SDC supply and return valves and for reactor head spray
isolation valve which were already closed at power. There was no impact on operability
of low-pressure coolant injection or containment spray functions of RHR.
A CENG investigation concluded human error as the primary cause for the inadvertent
isolation of the RCIC system. A contributing cause was the failure to implement
adequate corrective actions following a similar RCIC isolation event in 2007. Immediate
corrective actions for this event included a human performance stand down that
reinforced use of human performance tools and the need to identify and mark critical
steps during pre-job briefs, retraining the I&C technicians involved in the event on proper
use of human performance error prevention techniques, and improving bypass switch
Enclosure
19
verification steps for procedure N2-ISP-LDS-Q010 and other similar leak detection
system surveillance procedures. CENG entered this issue in their CAP as CR-2013-
002461.
Analysis. The inspectors determined that CENGs failure to correctly implement
surveillance test procedure N2-ISP-LDS-Q010 was a performance deficiency that was
within CENGs ability to foresee and correct and should have been prevented. This
finding is more than minor because it is associated with the human performance attribute
of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure
the availability, reliability, and capability of systems that respond to initiating events to
prevent undesirable consequences (i.e., core damage). Specifically, the inadvertent
isolation rendered the RCIC system inoperable and unable to perform its function for
approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Additionally, this finding is similar to Example 4.b. of IMC 0612,
Appendix E, Examples of Minor Issues, and is more than minor due to the procedural
error leading to a plant transient, i.e. an unplanned RCIC isolation.
In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of
IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power,
issued June 19, 2012, this finding represents a loss of safety function. Unit 2 is a
BWR-5, and as a result, RCIC is treated as having a separate high- pressure injection
safety function. A detailed analysis was conducted using SAPHIRE Version 8.0.8.0 and
Unit 2 SPAR Model 8.17. Using an exposure period of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and conservatively
assuming no recovery of the failed equipment, this finding had a change in core damage
frequency of low E-8. The dominant accident sequence was a grid- related loss of off-
site power with a failure of Division III power and the failure to recover off-site power and
the EDGs in 30 minutes. Since the change in core damage frequency was less than
1E-7, contributions from large early release and external event did not need to be
considered. Therefore, this finding was determined to be of very low safety significance
(Green).
This finding had a cross-cutting aspect in the area of Human Performance, Work
Practices, because the I&C technicians did not effectively employ self-checking and
place-keeping when implementing N2-ISP-LDS-Q010 which directly contributed to the
resulting procedural error H.4(a).
Enforcement. TS 5.4.1, Procedures, requires written procedures to be established,
implemented, and maintained covering the applicable procedures recommended in
Regulatory Guide (RG) 1.33, Quality Assurance Program Requirements (Operation),
Appendix A, Revision 2, dated February 1978. Section 8.b(2)(b) of RG 1.33 requires, in
part, specific procedures for surveillance tests on containment isolation. CENG
surveillance test procedure N2-ISP-LDS-Q010, Reactor Building General Area
Temperature Instrument Channel Functional Test, directed that the RHR/RCIC
ISOLATION BYPASS switch be placed in BYPASS to prevent an inadvertent
containment isolation while lifting thermocouple leads. Contrary to above, on April 2,
2013, technicians lifted thermocouple leads without ensuring the isolation switch was
bypassed, resulting in an unplanned isolation of the RCIC system. Because this issue is
of very low safety significance (Green) and was entered into CENGs CAP as CR-2013-
002461, this violation is being treated as an NCV, consistent with Section 2.3.2 of the
NRC Enforcement Policy. (NCV 05000410/2013003-01, Failure to Follow
Containment Isolation System Surveillance Procedure Resulting in Isolation of the
Reactor Coolant Isolation Cooling System)
Enclosure
20
Cornerstone: Emergency Preparedness
1EP4 Emergency Action Level and Emergency Plan Changes (71114.04 - 1 sample)
a. Inspection Scope
The Office of Nuclear Security and Incident Response headquarters staff performed an
in-office review of the latest revisions of various emergency plan implementing
procedures and the emergency plan located under ADAMS accession number
ML131071146 as listed in the Attachment.
CENG determined that in accordance with 10 CFR 50.54(q), the changes made in the
revisions resulted in no reduction in the effectiveness of the plan and that the revised
plan continued to meet the requirements of 10 CFR 50.47(b) and Appendix E to 10 CFR
Part 50. The NRC review was not documented in a safety evaluation report and did not
constitute approval of CENG-generated changes; therefore, this revision is subject to
future inspection.
b. Findings
No findings were identified.
1EP6 Drill Evaluation (71114.06 - 1 sample)
Training Observation
a. Inspection Scope
The inspectors observed a simulator training evolution for CENGs licensed operators on
April 2, 2013, which required emergency plan implementation by an operations crew.
The inspectors observed Unit 1 licensed operator performance during an evaluated
simulator scenario that included a loss of condenser vacuum, a stuck open ERV, and an
anticipated transient without scram. CENG 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 focus of the inspectors activities was to note any weaknesses and
deficiencies in the crews performance and ensure that CENG evaluators noted the
same issues and entered them into the CAP.
b. Findings
No findings were identified.
Enclosure
21
2. RADIATION SAFETY
Cornerstone: Public Radiation Safety and Occupational Radiation Safety
2RS1 Radiological Hazard Assessment and Exposure Controls (71124.01)
a. Inspection Scope
From April 22 to 25, 2013, the inspectors reviewed and assessed CENGs performance
in assessing the radiological hazards and exposure control in the workplace associated
with licensed activities and the implementation of appropriate monitoring and exposure
control measures for both individual and collective exposures.
The inspectors interviewed the radiation protection manager, radiation protection
supervisors, radiation protection technicians (RPTs), and radiation workers. The
inspectors performed walkdowns of various portions of the plant, performed independent
radiation dose rate measurements, observed work activities in radiological control areas,
and reviewed CENG documents during the N1R22 outage. The inspectors used the
requirements in 10 CFR 20, guidance in Regulatory Guide (RG) 8.38, Control of Access
to High and Very High Radiation Areas of Nuclear Plants, TSs, and CENGs procedures
required by TSs as criteria for determining compliance.
Inspection Planning
The inspectors reviewed the results of radiation protection program audits. The
inspectors reviewed reports of operational occurrences related to occupational radiation
safety since the last inspection on March 21, 2013.
Radiological Hazard Assessment
The inspectors conducted walkdowns and independent radiation measurements to
evaluate material, work and radiological conditions in the facility including the drywell,
RB, refueling floor, and turbine building (TB).
The inspectors selected the following radiological risk-significant work activities that
involved exposure to radiation:
Refueling floor activities
Drywell control rod drive under-vessel work
Drywell scaffolding
Drywell ISI
RWCU valve repairs
For these work activities, 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.
The inspectors observed work in potential airborne radioactivity areas and evaluated
whether the air samples from under the reactor vessel, from the reactor cavity and for
Enclosure
22
entries into the tent for repair of the SFP gate, were representative of the breathing air
zone and were properly evaluated. The inspectors evaluated whether continuous air
monitors on the refueling floor in the RB and at the drywell entrance were located to
ensure appropriate detection sensitivity and were representative of actual work areas.
The inspectors evaluated CENGs program for monitoring levels of loose surface
contamination in areas of the plant.
Instructions to Workers
The inspectors reviewed the following radiation work permits (RWPs) used to access
high radiation areas and evaluated if the specified work control instructions and control
barriers were consistent with TS requirements for locked high radiation areas:
RWP 113330H, RB 261 RWCU Valve Work
RWP 113802H, Drywell Under-Vessel Work
RWP 113890A, RB 340 Reactor Disassembly and Reassembly
RWP 113890B, RB 340 Underwater Work on Refuel Floor
RWP 113890E, RB 340 Reactor Cavity and Equipment Storage Pit Decon
RWP 113815, RB 261 Flow Accelerated Corrosion (FAC) ISI
RWP 113810, Drywell General Scaffolding Activities
The inspectors assessed whether permissible dose for radiological-significant work
under each RWP was clearly identified. The inspectors evaluated whether electronic
personal dosimeter alarm set points were in conformance with survey indications and
plant procedural requirements.
The inspectors reviewed CR-2013-002474 and CR-2012-002974 for occurrences where
a workers electronic personal dosimeter noticeably malfunctioned or alarmed. The
inspectors evaluated whether workers responded appropriately to the off-normal
condition. The inspectors assessed whether the issue was included in the CAP and
whether compensatory dose evaluations were conducted.
For work activities that could suddenly and severely increase radiological conditions, i.e.,
upper elevation of drywell during spent fuel movement and low power range monitor
moves, the inspectors assessed CENGs means to inform workers of these changes that
could significantly impact their occupational dose.
Contamination and Radioactive Material Control
The inspectors observed the access control point where CENG monitors potentially
contaminated material leaving the radiological control area and inspected the methods
used for control, survey, and release from these areas. The inspectors observed the
performance of personnel surveying and releasing material for unrestricted use and
evaluated whether the release surveys were performed in accordance with plant
procedures and process knowledge concerning the equipment.
Enclosure
23
Radiological Hazards Control and Work Coverage
The inspectors evaluated ambient radiological conditions and performed independent
radiation measurements during plant walkdowns. The inspectors assessed whether the
conditions were consistent with applicable posted surveys, RWPs, and associated
worker briefings.
The inspectors assessed whether radiation monitoring devices were placed on the
individuals body consistent with CENG procedures. The inspectors assessed whether
the dosimeter was placed in the location of highest expected dose and that CENG
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; e.g., RWCU
repairs and workers under vessel in the control rod drive area.
The inspectors reviewed the following RWPs for work within airborne radioactivity areas
with the potential for individual worker internal exposures:
RWP 113802H, Under-Vessel Control Rod Drive Work
RWP 113890E, RB 340 Reactor Cavity and Equipment Storage Pit Decontamination
The inspectors evaluated airborne radioactive controls and monitoring including potential
for significant airborne levels. The inspectors assessed applicable containment barriers
integrity and the operation of temporary high-efficiency particulate air ventilation system.
Risk-Significant High Radiation Area and Very High Radiation Area Controls
The inspectors discussed the controls and procedures for high risk high radiation areas
and very high radiation areas with the radiation protection manager. 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 during refueling outages.
The inspectors evaluated the controls for very high radiation areas and areas with the
potential to become a very high radiation area to ensure that an individual was not able
to gain unauthorized access to these areas.
Radiation Worker Performance
The inspectors observed the performance of radiation workers during the N1R22 with
respect to stated radiation protection work requirements. The inspectors assessed
whether workers were aware of the radiological conditions in their workplace, the RWP
controls and limits, and whether their behavior reflected the level of radiological hazards
present.
Radiation Protection Technician Proficiency
The inspectors observed the performance of the RPTs during the N1R22 with respect to
controlling radiation work. The inspectors evaluated whether technicians were aware of
Enclosure
24
the radiological conditions in their workplace, the RWP controls and limits, and whether
their behavior was consistent with their training and qualifications with respect to the
radiological hazards and work activities.
Problem Identification and Resolution
The inspectors evaluated whether problems associated with radiation monitoring and
exposure control were being identified by CENG at an appropriate threshold and were
properly addressed for resolution in the CENGs CAP. The inspectors assessed the
appropriateness of the corrective actions for a selected sample of problems documented
by CENG that involved radiation monitoring and exposure controls. The inspectors
assessed CENGs process 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 reasonably achievable (ALARA)
during the N1R22. The inspectors used the requirements in 10 CFR 20, RG 8.8,
Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear
Power Stations will be As Low As Is Reasonably Achievable, RG 8.10, Operating
Philosophy for Maintaining Occupational Radiation Exposures As Low As Is Reasonably
Achievable, TSs, and CENGs procedures required by TSs as criteria for determining
compliance.
Inspection Planning
The inspectors reviewed pertinent information regarding CENGs collective dose history,
current exposure trends, and ongoing or planned activities in order to assess current
performance and exposure challenges.
The inspectors reviewed changes in the radioactive source term by reviewing the trend
in average contact dose rates on reactor recirculation piping for the time period between
1984 and the present Unit 1 outage. The inspectors reviewed ALARA procedures that
specified the processes used to estimate and track exposures for radiological work
activities.
Radiological Work Planning
The inspectors selected the following work activities that had the highest exposure
significance:
ALARA Plan 2013-1-002, Drywell Under-Vessel Activities and Associated Activities
N1R22
ALARA Plan 2013-1-004, Drywell Operations and Local Leak Rate Test Activities
Enclosure
25
ALARA Plan 2013-1-006, Drywell ISI Activities
ALARA Plan 2013-1-007, Recirculation Pump Seals Replacement and Motor PMs
(Numbers 11, 13, and 15)
ALARA Plan 2013-1-010, Drywell Scaffold Activities
ALARA Plan 2013-1-014, Drywell Emergency Relief Valve and Pilot Valve Work
Activities
ALARA Plan 2013-1-024, Main Steam Isolation Valve 01-02 Stem Replacement
Actuator Remove/Replace and Testing
ALARA Plan 2013-1-029, Balance of Plant FAC Activities in RWCU Heat Exchanger
Room and Valve Aisles
ALARA Plan 2013-1-030, Refuel Floor Activities
ALARA Plan 2013-1-031, RWCU Miscellaneous Maintenance, Preventive
Maintenance, Surveillance Testing, Operations N1R22
The inspectors reviewed the ALARA work activity evaluations, exposure estimates, and
exposure reduction requirements. The inspectors determined whether CENG
reasonably grouped the radiological work into work activities based on historical
precedence, industry norms, and/or special circumstances.
The inspectors assessed when CENGs planning identified appropriate dose reduction
techniques, considered alternate dose reduction features, and estimated reasonable
dose goals. The inspectors evaluated whether the ALARA assessment had taken into
account decreased worker efficiency from use of respiratory protective devices and/or
heat stress mitigation equipment. The inspectors determined whether 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.
Verification of Dose Estimates and Exposure Tracking Systems
The inspectors reviewed the assumptions and basis for the current annual collective
dose estimate and outage collective dose estimate for accuracy. 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 CENG had established measures to track, trend, and
reduce occupational doses for ongoing work activities. The inspectors assessed
whether dose threshold criteria were established to prompt additional reviews and/or
additional ALARA planning and controls.
The inspectors evaluated CENGs 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 radiation protection and ALARA principles or if they were just
adjusted to account for failures to plan/control the work.
Enclosure
26
Source Term Reduction and Control
The inspectors used station records to determine the historical trends and current status
of plant source term known to contribute to elevated facility collective exposure. The
inspectors assessed whether CENG had made allowances or 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.
Radiation Worker Performance
The inspectors observed radiation workers and RPTs performance during refueling
outage activities in radiation areas, airborne radioactivity areas, and high radiation areas.
The inspectors evaluated whether workers demonstrated the ALARA philosophy in
practice and whether there were any procedure or RWP compliance issues.
Problem Identification and Resolution
The inspectors evaluated whether problems associated with ALARA planning and
controls were being identified by CENG at an appropriate threshold and were properly
addressed for resolution in the CENGs CAP.
b. Findings
No findings were identified.
2RS3 In-Plant Airborne Radioactivity Control and Mitigation (71124.03)
a. Inspection Scope
This area was inspected to verify in-plant airborne concentrations were being controlled
consistent with ALARA principles and the use of respiratory protection devices on-site
does not pose an undue risk to the wearer. The inspectors used the requirements in
10 CFR 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 CENGs procedures
required by TSs as criteria for determining compliance.
Inspection Planning
The inspectors reviewed the UFSAR 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 such that actions to prevent an overexposure may be taken. The
review included an overview of the respiratory protection program and a description of
the types of devices used. The inspectors reviewed procedures for maintenance,
inspection, and use of respiratory protection equipment as well as procedures for
maintenance and testing of breathing air quality.
Enclosure
27
Engineering Controls
The inspectors reviewed CENGs use of permanent and temporary ventilation to
determine whether CENG uses ventilation systems as part of its engineering controls to
control airborne radioactivity. The inspectors reviewed procedural guidance 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 setups on the refuel floor used
to support work in contaminated areas. The inspectors assessed whether the use of
these systems is consistent with procedural guidance and ALARA principles.
The inspectors reviewed airborne monitoring protocols for the drywell and refueling floor
continuous air monitors used to monitor and warn of changing airborne concentrations in
the plant and evaluating whether the alarms and set points are sufficient to prompt
worker action to ensure that doses are maintained within the limits of 10 CFR 20 and the
ALARA concept.
The inspectors assessed whether CENG had established threshold criteria for
evaluating levels of airborne beta-emitting and alpha-emitting radionuclides.
Use of Respiratory Protection Devices
The inspectors selected RWCU repairs and under-vessel control rod drive work activities
where respiratory protection devices were used to limit the intake of radioactive
materials and assessed whether CENG performed an evaluation concluding that further
engineering controls were not practical and that the use of respirators is ALARA. The
inspectors also evaluated whether CENG had established means (such as routine
bioassay) to determine if the level of protection (protection factor) provided by the
respiratory protection devices during use was at least as good as that assumed in work
controls and dose assessment.
Problem Identification and Resolution
The inspectors evaluated whether problems associated with the control and mitigation of
in-plant airborne radioactivity were being identified by CENG at an appropriate threshold
and were properly addressed for resolution in CENGs CAP. The inspectors assessed
whether the corrective actions were appropriate for a selected sample of problems
involving airborne radioactivity and were appropriately documented.
b. Findings
No findings were identified.
2RS4 Occupational Dose Assessment (71124.04)
a. Inspection Scope
From April 22 to 25, 2013, the inspectors reviewed occupational doses to ensure they
were appropriately monitored and assessed. The inspectors used the requirements in
10 CFR 20, RG 8.13, Instruction Concerning Prenatal Radiation Exposure, RG 8.36,
Enclosure
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Radiation Dose to the Embryo/Fetus, RG 8.40, Methods for Measuring Effective Dose
Equivalent from External Exposure, TSs, and CENGs procedures required by TSs as
criteria for determining compliance.
Inspection Planning
The inspectors reviewed the results of Unit 1 radiation protection program audits related
to internal and external dosimetry. A review was conducted of procedures associated
with dosimetry operations including issuance/use of external dosimetry, assessment of
internal dose, and evaluation of and dose assessment for radiological incidents. The
inspectors evaluated whether CENG had established procedural requirements for
determining when external dosimetry and internal dose assessments are required.
External Dosimetry
The inspectors evaluated whether CENGs dosimetry vendor was accredited with the
National Voluntary Laboratory Accredited Program and if the approved irradiation test
categories for each type of personnel dosimeter used were consistent with the types and
energies of the radiation present and the way the dosimeter is being used.
The inspectors evaluated the onsite storage of dosimeters before issuance, during use,
and before processing and reading. The inspectors also reviewed the guidance
provided to radiation workers with respect to care and storage of dosimeters.
The inspectors assessed the use of electronic personal dosimeters to determine if
CENG uses a correction factor to address the response of the electronic personal
dosimeter as compared to the dosimeter of legal record for situations when the
electronic personal dosimeter is used to assign dose and whether the correction factor is
based on sound technical principles.
The inspectors reviewed two CAP documents for adverse trends related to electronic
personal dosimeters. The inspectors assessed whether CENG 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 counting 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 CENGs 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
29
Special Bioassay (In Vitro)
There was no internal dose assessments obtained using whole body count results for
the inspectors to review. There was no internal dose assessments obtained using
urinalysis or fecal sample results for the inspectors to review.
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
The inspectors reviewed CENGs program for dose assessment based on airborne
monitoring and calculations of derived air concentration calculations. The inspectors
determined whether flow rates and collection times for air sampling equipment were
adequate to allow appropriate lower limits of detection to be obtained. CENG had
performed internal dose assessments using airborne/derived air concentration
monitoring for some work in the cavity during the N1R22.
Internal Dose Assessment - Whole Body Count Analyses
CENG has not documented any internal dose assessments using whole body count
results during the period reviewed.
Special Dosimetry Situations
Declared Pregnant Workers
The inspectors assessed the process used by CENG to inform 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 monitoring and controlling exposure to a
declared pregnant worker. CENG has not documented any internal dose assessments
for declared pregnant workers during this inspection period.
Dosimeter Placement and Assessment of Effective Dose Equivalent for External
Exposures
The inspectors reviewed CENGs methodology for monitoring external dose in non-
uniform radiation fields or where large dose gradients exist. The inspectors evaluated
CENGs criteria for determining when alternate monitoring such as use of multi-badging
is to be implemented.
The inspectors reviewed dose assessments performed for workers performing under-
vessel work and RWCU repairs. These workers used multi-badging to evaluate effective
dose equivalent and the dose assessment was performed consistent with CENG
procedures and dosimetry standards.
Enclosure
30
Shallow Dose Equivalent
There were no dose assessments for shallow dose equivalent available for review. The
inspectors evaluated CENGs method for calculating shallow dose equivalent from
distributed skin contamination or discrete radioactive particles.
Assigning Dose of Record
For the special dosimetry situations reviewed in this section, the inspectors assessed
how CENG 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 assessment was based on these techniques.
Problem Identification and Resolution
The inspectors assessed whether problems associated with occupational dose
assessment are being identified by CENG at an appropriate threshold and are properly
being addressed for resolution in CENGs CAP. The inspectors assessed the
appropriateness of the corrective actions for a selected sample of problems documented
by CENG involving occupational dose assessment.
b. Findings
No findings were identified.
2RS7 Radiological Environmental Monitoring Program (71124.07)
a. Inspection Scope
From May 6 to 10, 2013, the inspectors verified that the radiological environmental
monitoring program (REMP) quantifies the impact of radioactive effluent released to the
environment and sufficiently validates the integrity of the radioactive gaseous and liquid
effluent release program.
The inspectors used the requirements in 10 CFR 20; 10 CFR 50, Appendix A, Criterion
60, Control of Release of Radioactivity to the Environment; 10 CFR 50, Appendix I,
Numerical Guides for Design Objectives and Limiting Conditions for Operations to Meet
the Criterion As Low As Is Reasonably Achievable for Radioactive Material in Light-
Water-Cooled Nuclear Power Reactor Effluents; 40 CFR 190, Environmental Radiation
Protection Standards for Nuclear Power Operations; 40 CFR 141, Maximum
Contaminant Levels for Radionuclides; RG 1.23, Meteorological Monitoring Programs
for Nuclear Power Plants; RG 4.1, Radiological Environmental Monitoring for Nuclear
Power Plants; RG 4.15, Quality Assurance for Radiological Monitoring Programs;
NUREG-1302, Offsite Dose Calculation Manual Guidance: Standard Radiological
Effluent Controls for Boiling Water Reactors; applicable industry standards; and CENG
procedures as criteria for determining compliance.
Enclosure
31
Inspection Planning
The inspectors reviewed CENGs annual radiological environmental operating reports for
2011 and 2012 and the results of any assessments since the last inspection to verify that
the REMP was implemented and reported in accordance with requirements. This review
included changes to the offsite dose calculation manual (ODCM) in environmental
monitoring, sampling locations, monitoring and measurement frequencies, land-use
census, inter-laboratory comparison program, and analysis of environmental data.
The inspectors reviewed Units 1 and 2 ODCMs to identify locations of environmental
monitoring stations. The inspectors reviewed Units 1 and 2 UFSARs for information
regarding the environmental monitoring program and meteorological monitoring
instrumentation. The inspectors reviewed quality assurance audits and technical
evaluations performed on the vendor analytical laboratory program.
The inspectors reviewed Units 1 and 2 radioactive effluent release reports for 2011 and
2012 and the most recent results from waste stream analysis to determine if CENG was
sampling and analyzing for the predominant radionuclides released in plant effluents.
Site Environmental Inspection
The inspectors walked down five air sampling stations and five environmental thermo
luminescent dosimeter (TLD) monitoring stations to determine whether they were
located as described in the ODCM and to determine the equipment material condition.
For the air samplers and TLD stations selected, the inspectors reviewed the calibration
and maintenance records to verify that they demonstrated adequate operability for these
components. Additionally, the review included the calibration and maintenance records
of four composite water samplers.
The inspectors performed an assessment of any compensatory environmental sampling
upon loss of a required sampling station.
The inspectors observed the collection and preparation of four environmental samples
from surface water and fish to verify that environmental sampling was representative of
the effluent release pathways as specified in the ODCM and that sampling techniques
were in accordance with procedures.
Based on direct observation and review of records, the inspectors assessed whether the
meteorological instruments were operable, calibrated, and maintained in accordance
with procedures. The inspectors assessed whether the meteorological data readout and
recording instruments in the control room and at the meteorological tower were operable
and accurate.
The inspectors evaluated whether missed and/or anomalous environmental samples
were identified and reported in the annual radiological environmental operating reports.
The inspectors selected five events that involved a missed sample or inoperable sampler
to verify that CENG had identified the cause and had implemented corrective actions.
The inspectors reviewed the assessment of any sample results detected above the
lower limits of detection and reviewed CENGs evaluation of associated radioactive
effluent release data that was the potential source of the released material. The 2011
Enclosure
32
radiological environmental operator report noted the detection of Iodine from the
Fukushima Daiichi accident during March and April 2011.
The inspectors selected the following five SSCs that contained licensed material for
which there was a credible mechanism for radioactive material to reach ground water:
Unit 1 drywell, reactor, and turbine building sumps
Unit 2 drywell, reactor, and turbine building sumps
Unit 2 stack condensate transfer line to radwaste
Old radwaste sumps W 11, 12, and 13, and concentrator waste tank cubicle
Waste water treatment facility clarified tanks and sludge pits
The inspectors assessed whether CENG had implemented a sampling, inspection, and
monitoring program to provide early detection of leakage from these SSCs to ground
water.
The inspectors evaluated whether decommissioning records of leaks, spills, and
environmental remediation since the previous inspection were retained in a retrievable
manner in the 10 CFR 50.75(g) decommissioning file. Two records were added to the
decommissioning file in 2012. The first was Unit 1 turbine building roof replacement,
and the second was tritium in-leakage to the Unit 1 screen house.
The inspectors reviewed any significant changes made by CENG to the ODCM as the
result of changes to the land census, long-term meteorological conditions, or
modifications to the sampler stations since the last inspection. The inspectors reviewed
technical justifications for any changed sampling locations to ensure that the changes
did not affect CENGs ability to monitor the impact of plant operations on the
environment.
The inspectors assessed whether the detection sensitivities for environmental samples
were below the lower limits of detection specified in the ODCM. The inspectors
reviewed quality control charts for laboratory radiation measurement instrument and
actions taken for degrading detector performance. The inspectors also reviewed the
results of the vendors quality control program including the inter-laboratory comparison
to assess the adequacy of the vendors program.
The inspectors reviewed the results of Entergy Nuclear Northeast (Entergy) inter-
laboratory and intra-laboratory comparison program to verify the adequacy of
environmental sample analyses performed by James A. Fitzpatrick Nuclear Power Plant
environmental laboratory. The inspectors assessed whether the results included for the
media radionuclide mix was appropriate for the facility.
Identification and Resolution of Problems
The inspectors assessed whether problems associated with the REMP and
meteorological monitoring programs were being identified by CENG at an appropriate
threshold and correction actions were assigned for resolution in CENGs CAP.
Enclosure
33
b. Findings
No findings were identified.
4. OTHER ACTIVITIES
4OA1 Performance Indicator Verification (71151)
RCS Specific Activity and RCS Leak Rate (4 samples)
a. Inspection Scope
The inspectors reviewed CENGs submittal for the RCS specific activity (BI01) and RCS
leak rate (BI02) performance indicators for both Unit 1 and Unit 2 for the period of April
1, 2011, through March 31, 2013. (Note: An additional 12 months of BI02 data was
reviewed due to CENG having updated and revised the BI02 performance indicator data
since the previous inspection.) To determine the accuracy of the performance indicator
reported during those periods, the inspectors used definitions and guidance contained in
Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance
Indicator Guideline, Revision 6. The inspectors also reviewed RCS sample analysis
and control room logs of daily measurements of RCS leakage and compared that
information to the data reported by the performance indicator. Additionally, the
inspectors observed surveillance activities that determined the RCS identified leakage
rate, and chemistry personnel taking and analyzing an RCS sample.
b. Findings
No findings were identified.
4OA2 Problem Identification and Resolution (71152 - 4 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 CENG 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.
b. Findings
No findings were identified.
Enclosure
34
.2 Semi-Annual Trend Review
a. Inspection Scope
The inspectors performed a semi-annual review of site issues, as required by Inspection
Procedure 71152 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 CENG outside of the CAP such as trend reports,
performance indicators, major equipment problem lists, system health reports,
maintenance rule assessments, and maintenance or CAP backlogs. The inspectors also
reviewed CENGs CAP database for the first and second quarters of 2013 to assess
condition reports written in various subject areas (equipment problems, human
performance issues, etc.) as well as individual issues identified during the NRCs daily
condition report review (Section 4OA2.1). The inspectors reviewed CENGs quarterly
trend report for the first quarter of 2013 conducted under CNG-QL-1.01-1008, Periodic
QPA Performance Reporting Process, Revision 00500, to verify that CENG personnel
were appropriately evaluating and trending adverse conditions in accordance with
applicable procedures.
b. Findings and Observations
No findings were identified.
Two trends were identified by the inspectors that had not been identified by CENG.
The inspectors noted a negative trend in the reliability and availability of the emergency
core cooling system (ECCS) keep-fill pumps on Unit 2. The low-pressure core spray
keep-fill pump 2CLS*P2 failed on January 9, 2013, due to motor overload (CR-2013-
000218). On February 28, the HPCS keep-fill pump suddenly failed (CR-2013-001633).
As part of an extent-of-condition review for the low-pressure core spray keep-fill pump
failing, operators identified that Division II RHR system keep-fill pump 2RHS*P2 motor
had an abnormal noise. On April 12, CENG replaced 2RHS*P2 motor. The ECCS
keep-fill pumps are Goulds Pump Model 3196ST with 215T Westinghouse motors rated
for 575 volts. Westinghouse investigations determined that each motor had a turn-to-
turn failure. The failure of the HPCS keep-fill pump resulted in Licensee Event Report
(LER) 2013-002, Failure of High-Pressure Core Spray System Pressure Pump due to a
Motor Winding Failure, in accordance with 10 CFR Part 50.73(a)(2)(v)(D) and 10 CFR
Part 21. All three keep-fill pump motors have been replaced, and CENG has entered
these issues into their CAP as noted by the condition reports above.
The inspectors noted a decrease in the reliability of the Unit 1 RB sumps, and as a
result, an increase in the number of emergency operating procedure entries by control
room operators due to sump failures. The decrease in reliability was noted by three
separate events regarding Unit 1 RB sumps that resulted in emergency operating
procedure entries. These events occurred on January 20, April 12, and April 24, and
were documented in CR-2013-000532, CR-2013-002743 and CR-2013-003371,
respectively. The inspectors review identified that although CENG had properly
assessed sump performance per the NRC maintenance rule 10 CFR 50.65 for the train
level criteria, CENG did not assess sump performance against the system level criteria.
CENG documented this issue in CR-2013-004828 and entered this issue into their CAP.
A subsequent CENG evaluation determined the RB floor and equipment sumps had
exceeded their performance monitoring group functional failure criteria and the systems
Enclosure
35
were placed into (a)(1) status. The inspectors determined that this issue was not more
than minor because the train level criteria were appropriately being monitored and
placing the RB sumps into (a)(1) status for exceeding system level criteria would not
have resulted in additional maintenance-related corrective actions being taken by
CENG.
.3 Annual Sample: Review of Repetitive Valve Packing Leakage Issues
a. Inspection Scope
The inspectors performed an in-depth review of CENGs root cause analysis and
corrective actions associated with CR-2011-007171 and CR-2011-010906 regarding two
forced shutdowns of Unit 2 due to excessive unidentified leak rates in 2011. The
inspectors focused on the implementation of corrective actions and extent-of-condition
and cause reviews as it applied to both units.
The inspectors assessed CENGs problem identification threshold, cause analyses,
extent-of-condition reviews, compensatory actions, and the prioritization and timeliness
of CENGs corrective actions to determine whether CENG was appropriately identifying,
characterizing, and correcting problems associated with this issue and whether the
planned or completed corrective actions were appropriate. The inspectors compared the
actions taken to the requirements of CENGs CAP and 10 CFR 50, Appendix B. In
addition, the inspectors performed field walkdowns and interviewed engineering
personnel to assess the effectiveness of the implemented corrective actions.
b. Findings and Observations
No findings were identified.
On August 6 and December 9, 2011, Unit 2 conducted forced shutdowns due to
excessive unidentified leakage rate. In both cases, the increased unidentified leakage
was determined to be from the failure of the recirculation discharge gate valve,
2RCS*MOV18A. CENG completed separate root cause analysis for both events and
determined the August 6 event was due to a design issue which subjects the packing to
excessive vibrations due to the valve gate being exposed to RCS system flow. The
December 9 event was determined to be the result of a workmanship error following the
August 6 event which resulted in a burr forming on the valve stem and eventually led to
the second packing failure.
The inspectors reviewed the root cause analysis and the ECP associated with the 2001
change in packing design for this valve. The inspectors reviewed photos and drawings
of the valve and interviewed engineering personnel. The inspectors concluded that
CENGs determination of the root cause and major contributing causes were reasonable
and had a sound technical basis. The inspectors also determined that corrective actions
for the August 6 event would not have been expected to preclude the December 9 event.
The inspectors reviewed CENGs extent-of-condition reviews and corrective actions
related to similar valves on both Units 1 and 2. The inspectors concluded that CENG
conducted an appropriate extent-of-condition review and identified other valves which
Enclosure
36
may be susceptible to the same failure mechanism. CENG also developed corrective
actions to enhance their valve packing program and designated an engineer to oversee
this program.
The inspectors conducted an independent review of condition reports from 2000 until the
present looking for excessive leakage issues associated with valve packing. The
inspectors confirmed that a large percentage of issues prior to 2001 and since 2007
have been related to RCS*MOV18A and the underlying design vulnerability. Corrective
actions related to this issue included enhancing torque specification values for the
packing, developing preventive maintenance items to re-torque the packing periodically,
and revising work packages. The inspectors determined these corrective actions were
reasonable and had been implemented appropriately and in a timely manner.
The inspectors also observed that appropriate effectiveness reviews were either
completed or were scheduled to be completed in a timely manner.
.4 Annual Sample: Human Performance Safety Culture Themes
a. Inspection Scope
This inspection focused on CENGs evaluation and resolution of an emerging theme in
the number of human performance cross-cutting issues associated with NRC inspection
findings. Specifically, in the third quarter of 2012, four NRC Green inspection findings
across multiple cornerstones were identified as having common cross-cutting aspects in
the area of Human Performance, Resources, H.2(c), because CENG did not provide
complete, accurate, and up-to-date procedures that were adequate to assure nuclear
safety. On August 9, 2012, CENG initiated CR-2012-007529 and performed an
apparent cause evaluation to assess this trend. The NRC completed Inspection
Procedure 71152 in the form of a problem identification and resolution annual sample to
assess this trend during the fourth quarter of 2012 to provide information to support the
end of cycle assessment. Subsequently, on November 7, CENG initiated CR-2012-
010211, A Cross-Cutting Theme Exists in the Aspect of Human Performance,
Resources, Documentation H.2(c), to further assess and address this adverse trend.
A root cause analysis was completed and corrective actions were recommended for
implementation. The inspectors selected this emerging trend for further review to
develop more recent insights into CENGs progress in addressing the cross-cutting
theme to provide meaningful input to the mid-cycle assessment process. The inspectors
reviewed CENG condition reports, the root cause evaluation, and corrective, preventive,
and compensatory actions associated with the emerging theme. The inspectors also
interviewed plant personnel. The four findings associated with cross-cutting theme
H.2(c) are summarized as follows:
Unit 1 - Inadequate torque applied to SDC isolation valve closure bolts (CR-2012-
001441)
Unit 2 - Loss of SFP cooling due an inadequate procedure (CR-2012-004850)
Unit 2 - Inadequate special operating procedure for loss of SFP cooling (CR-2012-
007811)
Unit 2 - Inadequate evaluation and implementation of design modification to the
turbine gland seal supply system (CR-2012-006615)
Enclosure
37
b. Findings and Observations
No findings were identified.
CENG identified an adverse trend existed in the cross-cutting aspect H.2(c) and
recognized that the theme affected broad areas of performance as assessed in the
fourth quarter of 2012. CENG completed the root cause assessment for the adverse
trend in the H.2(c) cross-cutting aspect in December 2012. The root cause analysis
evaluated the four Green findings and also independently determined the common
causes of these findings.
CENG concluded that the work and administrative control documents and processes
were adequate, but the implementation of these processes was not adequate. Formal
techniques were used to reach this conclusion. The 46 specific causal factors from the
four findings were generalized into 13 general causal areas which were further
condensed (or binned) into five causal themes. The process of generalization of the
causal factors resulted in the majority of causal factors (53 percent) having the theme of
lack of engineering /challenge assumptions /mindset (willingness to accept answer with
no challenge). CENG further concluded a less rigorous standard resulted in products
that were of insufficient quality. The error drivers may be both process and behavior;
however, the results of the common cause analyses did not indicate that process
problems were significant errors.
CENG determined that the root cause of the trend was that site leadership had not
identified marginal performance relative to the technical rigor in the production of work
execution documents and, as such, has not put in place corresponding corrective or
mitigating strategies. A contributing cause was listed that existing administrative
controls governing changes to work orders and reviews of said changes are too lenient
to ensure high quality documents are consistently prepared to support plant operations
and maintenance activities.
The root cause team recommended 22 corrective actions in the report. CENG
management translated these recommendations into 20 unique corrective actions to be
implemented, 18 of which had been completed by the end of the first quarter 2013. The
two remaining corrective actions were to complete quarterly effectiveness reviews and a
final effectiveness review. The assigned corrective action to prevent recurrence
(CAPR159) was formulated to develop and communicate a station policy addressing
work documentation quality.
The corrective actions focused substantially on training plant personnel to properly
implement their procedures and to hold them accountable if they did not follow the
procedures. Three of the recommended corrective actions involved development of or
changes to work procedures. CA #59 was to define the term skill of the craft in a
procedure and was completed on June 12, using guidance obtained from an industry
group; CA #55 was to develop and implement a fleet conduct of engineering
administrative procedure and was closed to CA #244 to reinforce current expectations
for engineering roles and responsibilities; and CA #64 was to develop a process tool to
assist in screening pen and ink changes to procedures. This corrective action was also
changed to revise site procedures to add a requirement to initiate a condition report if a
procedure could not be completed as written. All but one corrective action relied on
knowledge-based corrective actions. The only rule-based corrective action was CA #59.
Enclosure
38
Although the majority of the corrective actions were knowledge-based activities that
relied upon one-time training presentations, only two corrective actions were
implemented to conduct a needs analysis for the specified training. The needs analysis
for CA #58 (improve the use of SDS-006 for bolt-torque requirements) and CA #164
(understanding the work order process) both concluded that no additional or recurring
training were required. The one-time training that had been administered would be
sufficient to correct the adverse trend. As a result, no changes to the initial site training
program will be made and these training topics will not be refreshed periodically during
proficiency training.
The inspectors noted the implemented corrective actions rely almost entirely upon a
series of one-time training activities to result in institutionalized changes to personnel
behavior and organizational culture into the future. Therefore, the effectiveness of the
corrective actions could diminish over time as personnel turnover occurs.
The effectiveness reviews for the corrective actions are scheduled to start in the third
quarter of 2013. There have been no effectiveness reviews completed on the efficacy of
the corrective actions for this cross-cutting aspect theme as of June 2013.
The inspectors could not conclude that CENGs root cause analysis and resultant
corrective actions are correct and effective since they have only recently been fully
implemented. However, the number of findings with a cross-cutting aspect in procedure
adequacy has declined from four to two from the end of cycle to mid cycle NRC reviews.
.5 Annual Sample: Battery Low Specific Gravities
a. Inspection Scope
The inspectors performed an in-depth review of CENGs evaluations and corrective
actions associated with low-specific gravity in the safety-related station batteries.
Specifically, an adverse trend of low-specific gravity readings for cells in all three
safety-related 125 volts direct current (VDC) station batteries at Unit 2 were identified in
CR-2012-001315.
The inspectors assessed CENGs problem identification threshold, extent-of-condition
reviews, compensatory actions, and the prioritization and timeliness of CENGs
corrective actions to determine whether CENG was appropriately identifying,
characterizing, and correcting problems associated with this issue and whether the
planned and completed corrective actions were appropriate. The inspectors compared
the actions taken to the requirements of 10 CFR 50, Appendix B. In addition, the
inspectors performed field walkdowns and interviewed engineering personnel to assess
the effectiveness of the implemented corrective actions.
b. Findings and Observations
CENG determined the most probable cause of the low-specific gravities was that the
battery vendors had removed some electrolyte prior to shipping the battery cells to
NMPNS; and then once at NMPNS, water was added to the cells that diluted the
concentration of sulfuric acid.
Enclosure
39
CENG performed a thorough review of the low-specific gravity issue and obtained
information from the battery vendors to support the probable cause. Corrective actions
included adjusting the method for calculating specific gravity and evaluating adding
electrolyte to restore the specific gravity to the manufacturers recommended level.
CENG verified, based on surveillance testing, that although the specific gravities were
lower than normal, the concentration of sulfuric acid was adequate to obtain sufficient
battery capacity to meet the design basis requirements of the batteries.
The inspectors reviewed condition reports, selected battery test results, and
correspondence from the battery vendors regarding the low-specific gravity issue. The
inspectors determined CENGs overall response to the issue was commensurate with
the safety significance, was timely and included appropriate compensatory actions. The
inspectors determined that the actions taken were reasonable to resolve the low-specific
gravity issue. As part of the review, the inspectors determined that two findings existed
as described below.
b.1 Inadequate Procedural Implementation for Battery Cell Replacement
Introduction. The inspectors identified a Green NCV of 10 CFR 50, Appendix B,
Criterion V, Instructions, Procedures, and Drawings, because CENG did not assure
that the replacement of cells in battery 2C was prescribed and performed by appropriate
procedures which resulted in degraded accuracy of test results and potential degradation
to safety-related battery cells.
Description. The Division III emergency battery bank, battery 2C, at Unit 2 uses jars that
contain three cells each to provide reliable direct current (DC) power for essential DC
loads required during normal and abnormal conditions. CENG determined that two jars
required replacing (a total of six cells). In preparation for this activity, CENG procured
three jars and stored them in the warehouse. The inspectors determined that several
procedural inadequacies existed during storage and subsequent cell replacement.
The cells in the warehouse were not monitored or maintained in accordance with vendor
recommendations. Specifically, the vendor requires that cells stored in spaces that are
not air conditioned should have individual cell voltages checked monthly and charged
when needed to prevent excessive discharge. Although CENG had previously noted
their poor practices with regards to battery storage and has ongoing corrective actions to
provide better storage facilities (as documented in CR-2010-012200), CENG did not take
action to adequately monitor cells in the warehouse. As a result, when the three jars for
battery 2C were obtained from the warehouse, one was found to be visibly sulfated and
had to be discarded, and the other two were found undercharged. Sulfation is an
indication of chronic undercharging and eventually results in permanent loss of capacity.
Although CR-2012-010907 identified the poor condition of the cells, the cell replacement
was continued with potentially degraded cells.
The newly installed cells were not charged prior to or upon installation. This is required
in the vendor manual and the station battery cell replacement procedure, N2-EMP-GEN-
673, 24/48 VDC and 125 VDC Batteries - Cell and Connector Replacement, Revision
00400.
Battery 2C was then subjected to a modified performance test with the newly installed
and uncharged cells. This resulted in over-discharging the new cells. Of the new cells,
Enclosure
40
the two lowest reached 0.903 VDC and 1.167 VDC as opposed to the expected end
voltage of approximately 1.75 VDC. This resulted in a battery capacity of 95 percent. In
comparison a normal battery at the age of battery 2C would have a capacity of
approximately 105 percent. Using uncharged cells artificially lowered the test results
which diminished the ability to use the test results for future trending and could mask
poor performance of the remaining cells.
Finally, after the modified performance test, one of the new cells did not recharge
properly. Specifically the vendor states that an equalization charge should be performed
until the lowest cell is within 0.05 volt of the average of all of the cells. During the
equalization charge for battery 2C after the modified performance test, one of the new
cells did not rise to within 0.05 volt of the average of all of the cells. Although CR-2012-
010901 recognized that the acceptance criteria had not been met, the acceptance
criteria was determined to be unnecessary. CENG did not recognize that the failure to
recharge properly was an indication that the previous procedural inadequacies may have
degraded the cell.
CENG entered these inspector-identified issues into the CAP as CR-2013-005235.
CENG corrective actions included reviewing the previous battery 2C test results and the
work order for the next scheduled modified performance test and verifying battery 2C will
remain operable until the next test scheduled for September 2013. CENG also initiated
CR-2013-005074 to replace the two newly installed jars.
Analysis. The inspectors determined that the failure to assure that the replacement of
cells in battery 2C was prescribed and performed by appropriate procedures was a
performance deficiency that was reasonably within CENGs ability to foresee and correct
and should have been prevented. This finding was more than minor because it was
associated with the equipment performance attribute of the Mitigating Systems
cornerstone and affected the cornerstone objective of ensuring the availability, reliability,
and capability of systems that respond to initiating events to prevent undesirable
consequences.
In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of
IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power,
issued June 19, 2012, the inspectors determined this finding to be of very low safety
significance (Green) because the performance deficiency was not a design or
qualification deficiency, did not involve an actual loss of safety function, did not represent
actual loss of a safety function of a single train for greater than its TS allowed outage
time, and did not screen as potentially risk significant due to a seismic, flooding, or
severe weather-initiating event.
This finding has a cross-cutting aspect in the area of Human Performance, Decision-
Making Component, because CENG did not use conservative assumptions in decision
making. Specifically, CENG did not monitor the cells in storage, question the adequacy
of the discharged cells, charge the cells prior to installation, or fully evaluate the
implications of the test and recharge results H.1(b).
Enforcement. 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and
Drawings, requires, in part, that activities affecting quality shall be prescribed by
documented instructions, procedures, or drawings of a type appropriate to the
circumstances and shall be accomplished in accordance with these instructions,
Enclosure
41
procedures, or drawings. Contrary to the above, CENG did not assure that the
November 2012 replacement of cells in battery 2C was prescribed and performed by
appropriate procedures which resulted in degraded accuracy of test results and potential
degradation to safety-related battery cells. Because this violation was of very low safety
significance (Green) and has been entered into CENGs CAP (CR-2013-005235), this
violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC
Enforcement Policy. (NCV 05000410/2013003-02, Inadequate Procedural
Implementation for Battery Cell Replacement)
b.2 Inadequate Design Control for Battery 2C
Introduction. The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B,
Criterion III, Design Control, because CENG did not verify the adequacy of the design
with respect to battery 2C. Specifically, by failing to size the battery to the most limiting
time period, the sizing calculation significantly overstated the available design margin.
Description. The Division III emergency battery bank, battery 2C, uses jars that contain
three cells each to provide reliable DC power for essential DC loads required during
normal and abnormal conditions at Unit 2. The inspectors reviewed EC-145,
Verification of Adequacy of Division III Battery 2BYS*BAT2C, Revision 2, to determine
if the calculation appropriately verified the adequacy of the size of the installed battery
2C. The inspectors noted that the calculation evaluated the battery based on two time
periods, a 1-minute period and a 119-minute period. In accordance with Institute of
Electrical and Electronics Engineers (IEEE) Standard 485-1997, IEEE Recommended
Practice for Sizing Lead-Acid Batteries for Stationary Applications, and EC-145, the
battery should be sized based upon the most demanding time period. The inspectors
determined that the sizing was incorrect. Specifically, although EC-145 determined that
the first time period (1 minute) was the most demanding, the battery sizing was based
upon the less demanding second time period (119 minutes).
In response to this issue, CENG agreed that the calculation was incorrect, entered this
issue into their CAP (CR-2013-005117), and evaluated the condition for operability.
CENG performed the battery sizing calculation based upon the correct time period and
determined that the battery capacity margin reduced from 26 percent to negative
11 percent (i.e., the battery was undersized by 11 percent). CENG reduced the battery
design and aging margins from the calculation and were able to increase the capacity
margin to positive 10 percent which demonstrated a reasonable expectation of
operability. The significance of reducing the design margin was that the original
calculation would have permitted modifications to the Division III DC system that could
have actually overloaded the battery. The significance of reducing the aging margin is
that the battery would not have been able to perform its design function as the battery
aged.
The inspectors independently performed battery sizing calculations and agreed with
CENGs results.
Analysis. The inspectors determined that the failure to verify the adequacy of the design
with respect to battery 2C was a performance deficiency that was reasonably within
CENGs ability to foresee and correct and should have been prevented. This finding was
more than minor because it was associated with the design control attribute of the
Enclosure
42
Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the
availability, reliability, and capability of systems that respond to initiating events to
prevent undesirable consequences.
In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of
IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power,
issued June 19, 2012, the inspectors determined this finding is of very low safety
significance (Green) because the performance deficiency was not a design or
qualification deficiency, did not involve an actual loss of safety function, did not represent
actual loss of a safety function of a single train for greater than its TS allowed outage
time, and did not screen as potentially risk-significant due to a seismic, flooding, or
severe weather-initiating event.
This finding did not have a cross-cutting aspect because it was not indicative of current
performance. Specifically, EC-145 was last revised in 2008.
Enforcement. 10 CFR 50, Appendix B, Criterion III, Design Control, states, in part, that
design control measures shall provide for verifying or checking the adequacy of design.
Contrary to the above, from July 17, 2008, to June 12, 2013, CENGs design control
measures had not appropriately verified the adequacy of the design regarding battery
2C. Specifically, by failing to size the battery to the most limiting time period, the sizing
calculation significantly overstated the available design margin. Because this violation
was of very low safety significance (Green) and has been entered into CENGs CAP
(CR-2013-005117), this violation is being treated as an NCV, consistent with Section
2.3.2 of the NRC Enforcement Policy. (NCV 05000410/2013003-03, Inadequate
Design Control for Battery Sizing Calculation)
4OA3 Follow-Up of Events and Notices of Enforcement Discretion (71153 - 6 samples)
.1 Plant Events
a. Inspection Scope
For the plant events 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 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. As applicable, the inspectors verified that CENG made appropriate
emergency classification assessments and properly reported the event in accordance
with 10 CFR Parts 50.72 and 50.73. The inspectors reviewed CENGs follow-up actions
related to the events to assure that CENG implemented appropriate corrective actions
commensurate with their safety significance.
Unit 1 loss of battery board 12 and SDC on April 16, 2013
Loss of all SDC pumps for 17 minutes on April 16, 2013
Enclosure
43
b. Findings
Introduction. The inspectors documented an apparent violation of Unit 1 TS 6.4.1,
Procedures, because CENG failed to properly restore from a loss of a vital DC bus in
accordance with station off-normal procedures resulting in an unplanned loss of all SDC
when time to boil was less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Specifically, operators failed to recognize a
potential for loss of SDC during battery bus 12 restoration in accordance with N1-SOP-
47A.1, Loss of DC, Revision 00101, and N1-OP-47A, VDC Power System, Revision
02500.
Description. Unit 1 shut down for a refueling outage on April 15, 2013. On April 16,
Unit 1 was in cold shutdown at 118 degrees Fahrenheit with a temperature band of 110
to 120 degrees Fahrenheit. The reactor vessel head was installed, and the head bolts
were in the process of being detensioned in preparation for reactor cavity flood up and
reactor vessel head removal. Primary containment was open for planned maintenance.
Decay heat removal was via the SDC pump 12. SDC pumps 11 and 13 were secured
with their breakers racked out to the test position for planned loss of offsite power/loss of
coolant accident testing (LOOP/LOCA).
During LOOP/LOCA testing, the SDC pumps and ECCS pumps in train associated with
the bus are racked to their test position. Operators are stationed in the field to restore
these pumps to normal so the pumps are still considered to be available. This is
permitted by NMPNS TSs; however, automatic functions of the pumps are not available
(such as auto start on a low-low reactor vessel level signal).
At 2:45 p.m. on April 16, a contractor walking down a tagout associated with an ERV
modification made an error and opened the breaker cabinet door for the vital DC bus 12.
The vital DC bus 12 cabinet door contains a mechanical interlock which opens battery
breaker 12 and the static battery charger DC output breaker, de-energizing the DC
switchgear when the door is open. Upon opening the breaker cabinet door and hearing
the breakers trip, the contractor realized he was in the incorrect cabinet and immediately
contacted the control room and notified them of the event. The vital bus was considered
protective equipment and a sign on the cabinet door cautioned that the door interlock
would trip the breakers in that cabinet. The loss of the vital DC bus 12 resulted in a
partial loss of indication in the main control room, loss of DC control power for the
associated bus, and a high-temperature trip signal for the SDC 12 being generated.
However, since DC power to the trip solenoid was also lost, the SDC pump 12 continued
to run. The ECCS pumps associated with the #12 bus were inoperable due to loss of
control power.
In response to the event, operators entered procedure N1-SOP-47A, Loss of DC,
Revision 00101. The flowchart in SOP-47A.1 directs the operator to transfer selected
loads normally powered from battery bus 12 to their alternate power supplies and then
directs restoration of the bus. However, a decision was made to not take actions
specified in N1-SOP-47A.1 and pursue restoring the vital DC bus 12 using system
operating procedure N1-OP-47A, 125 VDC Power System, Revision 02500. The
inspectors noted that N1-SOP-47A.1 Section 5.1 contains two caution statements stating
that pump trip signals may have been generated while the bus was de-energized and
those signals must be cleared prior to restoration or a pump trip may occur when the bus
is restored and power is supplied to the DC trip coils. However, neither N1-SOP-47A.1
nor N1-OP-47A contained a list of tripping circuits and tripping actions which are
Enclosure
44
associated with the vital DC bus 12. Operators failed to recognize the bus 12
high-temperature trip signal present on the alarm log and the plant process computer
displays prior to attempting to restore bus 12. The presence of the trip signal was also
indicated by a control room annunciator which was locked-in since the loss of battery
bus 12 at 2:45 p.m.
At 3:45 p.m., field operators attempted to close static battery charger 171A DC output
breaker to restore the battery bus from its alternate power supply. Due to the high-
temperature trip signal already being present on the SDC pump 12, when operators
attempted to close the static battery charger 171A output breaker, the DC trip coil
received enough power to energize the relay and trip the SDC pump 12 just before the
static battery charger 171A output breaker tripped due to the mechanical interlock.
Operators did not immediately recognize that they had lost SDC pump 12 via their
indications at the control panel (i.e.; annunciator, pump current, pump flow). Upon
recognizing the loss of SDC at approximately 3:50 p.m., operators entered N1-SOP-6.1
Loss of SFP/RX Cavity Level/Decay Heat Removal, Revision 00501.
At 3:50 p.m., the control room directed the breakers for SDC pumps 11 and 13 to be
racked to their normal positions and that SDC be restored using the 11 and 13 SDC
pumps. The 11 SDC pump breaker was restored at 4:03 pm and SDC flow was restored
at 4:17 pm when the SDC 11 temperature control valve was opened, restoring cooling
flow to the reactor. Reactor vessel temperature rose from 118 to 145 degrees
Fahrenheit as a result of the loss of SDC. At 5:11 p.m., the normal DC power
distribution lineup was restored.
CENG immediately conducted prompt investigations of both the loss of battery bus 12
and loss of SDC events, entered both events into their CAP as CR-2013-002926 and
CR-2013-002916, and conducted a root cause analysis. CENG determined the root
cause for the loss of SDC was inadequate procedural guidance for restoring the DC
power. Contributing causes included operators proceeding in the face of uncertainty,
management oversight of operations, and inadequate use of operational experience
which could have precluded this event. Corrective actions to prevent recurrence
included a review of operations procedures to ensure those procedures contain
adequate levels of detail to safely recover from the event and restore the system to
normal operation.
Analysis. The inspectors determined that CENGs failure to properly restore the battery
bus 12 in accordance with plant procedures was a performance deficiency that was
reasonably within CENGs ability to foresee and correct and should have been
prevented. The performance deficiency was determined to be more than minor because
the inspectors determined it affected the configuration control aspect of the Initiating
Events cornerstone and adversely affected the associated cornerstone objective to limit
the likelihood of events that upset plant stability and challenge critical safety functions
during shutdown as well as power operations. Specifically, operators failed to recognize
a potential for loss of SDC during battery bus 12 restoration in accordance with N1-SOP-
47A.1, Loss of DC, Revision 00101, and N1-OP-47A, VDC Power System, Revision
02500. This performance deficiency initiated a plant transient, loss of shutdown cooling.
The inspectors evaluated the finding using IMC 0609 Attachment 0609.04, Initial
Characterization of Findings, issued June 19, 2012, and IMC 0609 Appendix G,
Shutdown Operations Significance Determination Process, issued February 28, 2005.
Enclosure
45
IMC 0609 Appendix G Table 1, Losses of Control, states a quantitative analysis is
required for:
Loss of Thermal Margin (PWRs and BWRs)
(Inadvertent change in RCS temperature due to loss of RHR)/(change in temperature
that would cause boiling) > 0.2 (temperature margin to boil)
In this case, RCS temperature changed 27 degrees (145 to 118 degrees Fahrenheit)
and the change in temperature to boiling was 94 degrees (212 to 118 degrees
Fahrenheit). Temperature margin to boil was greater than 0.2 (0.2872); thus, a
quantitative analysis was required. The significance of the finding is designated as To
Be Determined (TBD) until a Phase 3 analysis can be completed by Regional and
Headquarters Senior Reactor Analysts.
The inspectors determined this finding had a cross-cutting aspect in the area of Human
Performance, Resources, because CENG did not ensure that personnel, equipment,
procedures, and other resources were available and adequate to assure nuclear safety -
complete, accurate and up-to-date design documentation, procedures, and work
packages, and correct labeling of components. Specifically, CENG procedures
N1-SOP-47A.1 and N1-OP-47A did not contain adequate guidance to ensure recovery
from a loss of a DC bus would not result in an unexpected plant transient H.2(c).
Enforcement. Unit 1 TS 6.4.1, Procedures, requires, in part, that written procedures
and administrative policies shall be established, implemented, and maintained that meet
or exceed the requirements and recommendations of Sections 5.1 and 5.3 of American
National Standards Institute N18.7-1972 Administrative Controls and Quality Assurance
for the Operational Phase of Nuclear Power Plants, and cover the following activities:
the applicable procedures recommended in RG 1.33, Quality Assurance Program
Requirements (Operation), Appendix A, Typical Procedures for Pressurized-Water
Reactors and Boiling-Water Reactors, dated November 3, 1972. RG 1.33, Appendix A,
Section 4, Procedure for Startup, Operation, and Shutdown of Safety-Related BWR
Systems, requires procedures for onsite DC system, and Section 6, Procedures for
Combating Emergencies and Other Significant Events, requires, in part, procedures for
including loss of electrical power (and/or degraded power sources). CENG procedures
N1-OP-47A, 125 VDC Power System, Revision 02500, and N1-SOP-47A.1, Loss of
DC, Revision 00101, implement this requirement. Contrary to the above, on April 16,
2013, operators were unable to properly implement N1-OP-47 and N1-SOP-47A.1
following a loss of the battery bus 12 resulting in a temporary loss of all decay heat
removal. This issue is being characterized as an apparent violation in accordance with
the NRC's Enforcement Policy, and its final significance will be dispositioned in a
separate future correspondence. (Apparent Violation 05000220/2013003-04,
Improper Bus Restoration Results in a Loss of Shutdown Cooling)
.2 (Closed) LER 05000220/2012-006-00: Technical Specification Required Shutdown Due
to Containment Leakage
a. Inspection Scope
On December 13, 2012, Unit 1 commenced a shutdown after observing nitrogen leakage
from primary containment over a period of 10 days. NRC Inspection Report
Enclosure
46
05000220/2012005 documented CENGs immediate response and the NRCs initial
review of the event. As of the end of the inspection documented in that report, CENGs
evaluation of the causes for the leakage was still ongoing. The inspectors had identified
an issue of concern regarding the total amount of leakage from primary containment
vent and purge valves and its relation to exceeding the required value in TS 3.3.3. The
NRC opened URI 05000220/2012005-03 to track CENGs completion of the root cause
evaluation, the quantification of the amount of leakage from primary containment for the
event, and the NRCs subsequent review of CENGs completed evaluation.
To close URI 05000220/2012005-03 the inspectors reviewed and independently verified
CENGs calculation regarding the quantity of leakage from primary containment from
December 3 - December 13. The inspectors also reviewed Appendix J Type B and C
testing of the primary containment vent and purge valves to determine leakage
quantities and how they impacted overall primary containment leakage. The inspectors
also reviewed the cause of the leakage and CENGs actions to address the cause which
was included in CR-2012-011157. URI 05000220/2012005-03 is closed to the violation
discussed below. The enforcement actions associated with this LER are discussed
below. This LER is closed.
b. Findings
Introduction. A self-revealing Green NCV of TS 3.3.3, Leakage Rate, was identified for
CENGs failure from December 3 to December 13, 2012, to maintain containment
leakage less than 1.5 percent by weight of the containment air per day and less than 0.6
percent by weight of the containment air per day for all penetrations and all primary
containment isolation valves subject to 10 CFR Part 50, Appendix J, Types B and C
tests, when pressurized to 35 pounds per square inch gauge (psig) when RCS
temperature is above 215 degrees Fahrenheit and primary containment integrity is
required.
Description. On December 3, 2012, at 11:31 a.m., Unit 1 established primary
containment integrity and commenced a reactor startup from an unplanned outage. The
following day at 2:40 a.m., CENG operators commenced adding nitrogen gas into the
primary containment as part of a planned activity to reduce primary containment oxygen
concentration to less than 4 percent as required by TS 3.3.1, Oxygen Concentration.
This activity was completed at 10:55 a.m. on December 4. Once an appropriate nitrogen
concentration has been achieved in the containment, additional makeup is generally not
required. However, from December 6 through December 8, on three occasions,
operators added additional nitrogen to the containment to maintain pressure within
procedural limits. This issue was documented in CR-2012-011157, Adverse Trend in
Unit 1 Nitrogen Usage. CENG commenced initial troubleshooting activities which
included examining systems and components that were possible sources of nitrogen
leakage; however, a definitive source for the leakage was not identified. On
December 12, following a fourth addition of nitrogen, CENG increased the importance of
the issue, formed an issue response team, and staffed the outage control center. As
part of the investigation process, operators cycled several containment isolation valves
in the nitrogen purge and vent system and attempted to quantify the amount of seat
leakage through the valves by opening test fittings located between isolation valves. In
parallel with the troubleshooting efforts, CENG and vendor personnel began to develop
analytical tools that could be used to quantify the amount of containment leakage.
Enclosure
47
On December 13, at 6:47 p.m., after observing a decrease in containment pressure
following a fifth nitrogen addition and receiving preliminary data that a containment
isolation valve local leak-rate test between reactor containment inert gas purge and fill
drywell cooling system isolation valves IV-201-31 and IV-201-32 may fail, CENG
commenced a plant shutdown because primary containment integrity as required in TS 3.3.3 could not be assured. On December 13, at 11:33 p.m., the plant reached cold
shutdown and exited plant TS 3.3.3.
Subsequent testing of containment isolation valves revealed that three valves in the
reactor containment inert gas purge and fill drywell cooling system, valves IV-201-10,
IV-201-31, and IV-201-32 had unacceptable seat leak rates. These conditions were
documented in condition reports 2012-011210 and 2012-011288. When the valves were
disassembled and examined, CENG identified that iron oxide (i.e., rust) buildup on the
valve resilient seats had prevented the valves from closing tightly and adversely
impacted seat leakage performance. The reactor containment inert gas purge and fill
drywell cooling system is a carbon steel system and the internal piping surface adjacent
to the valves had visible signs of iron oxide degradation. CENG corrective actions
included removing the loose surface rust, installing new resilient seats on the valves,
and successfully performing as-left local leak-rate tests on the subject valves. Additional
corrective actions were outlined in CR-2012-011247.
CENG analysis determined that based upon the nitrogen supplied to the drywell,
containment leakage from December 3 through December 13, 2012, exceeded the limits
in TS 3.3.3 which requires containment leakage to be less than 1.5 percent by weight of
the containment air per day and less than 0.6 percent by weight of containment air per
day for all penetrations and all primary containment isolation valves subject to 10 CFR Part 50, Appendix J, Types B and C tests, when pressurized to 35 psig when RCS
temperature is above 215°F and primary containment integrity is required. Specifically,
leakage was calculated to be between 1,421 and 2,023 standard cubic feet per hour
verses a calculated limit of 647 standard cubic feet per hour.
Analysis. The inspectors determined that CENGs failure to maintain containment
leakage from December 3 through December 13, 2012, within the limits required by TS 3.3.3 was a performance deficiency that was within CENGs ability to foresee and
correct and should have been prevented. This finding is more than minor because it is
associated with the SSC and barrier performance attribute of the Barrier Integrity
cornerstone and affected the cornerstone objective to provide reasonable assurance that
physical design barriers (fuel cladding, RCS, and containment) to protect the public from
radionuclide releases caused by accidents or events. Specifically, containment leakage
from December 3 through December 13 exceeded the leakage limits outlined in Unit 1
In accordance with IMC 0609.04, Initial Characterization of Findings, and Table 6.2,
Phase 2 Risk Significance-Type B Findings at Full Power, of IMC 0609, Appendix H,
Containment Integrity Significance Determination Process, issued May 6, 2004, the
inspectors determined this finding was of very low safety significance (Green) because
the leakage was less than 100 percent of containment volume per day for the duration of
the leak.
This finding has a cross-cutting aspect in the area of Problem Identification and
Resolution, CAP, because CENG failed to take appropriate corrective action to address
Enclosure
48
safety issues and adverse trends in a timely manner commensurate with their safety
significance. Specifically, following identification of the adverse trend regarding the
frequency of nitrogen addition to the drywell, CENG did not assess in a timely manner
the significance of the leakage and the impact on primary plant containment. As a
result, plant operation continued for several days with drywell leakage that exceeded the
limits outlined in TS 3.3.3 P.1(d).
Enforcement. TS 3.3.3, Leakage Rate, requires containment leakage to be less than
1.5 percent by weight of the containment air per day and less than 0.6 percent by weight
of the containment air per day for all penetrations and all primary containment isolation
valves subject to 10 CFR Part 50, Appendix J, Types B and C tests, when pressurized
to 35 psig when RCS temperature is above 215 degrees Fahrenheit and primary
containment integrity is required. Contrary to the above, from December 3 through 13,
2012, containment leakage exceeded 1.5 percent by weight. Specifically, following a
December 13 plant shutdown, CENG determined containment leakage during this period
to have been between 1,421 and 2,023 standard cubic feet per hour verses a calculated
limit of 647. Because this violation is of very low safety significance (Green) and CENG
entered this issue into their CAP as CR-2013-011247, this finding is being treated as an
NCV consistent with consistent with Section 2.3.2 of the NRC Enforcement Policy.
(NCV 05000220/2013003-05, Containment Leakage Exceeds Technical
Specification 3.3.3 Limits)
.3 (Closed) LER 05000220/2012-006-01: Technical Specification Required Shutdown Due
to Containment Leakage
This LER was revised on June 14, 2013, to reflect changes in corrective actions that
were outlined in the original LER submittal. In the original LER, CENG indicated that
during the spring 2013 refueling outage, the internal surfaces of the horizontal drywell
vent and purge piping that contained valves IV-201-09, IV-201-10, IV-201-31, and
IV-201-32 would be coated with a material that would minimize the recurrence of rust
buildup on the piping. Further, during the outage, the vertical piping that contained
valves IV-201-07, IV-201-08, IV-201-16, and IV-201-17 would be inspected; and based
on the inspection findings, a coating strategy (if required) would be developed for that
piping. Subsequent to submittal of the original LER, CENG determined that based upon
the difficultly associated with application of a suitable coating to the pipes and the
potential of subsequent coating failure, a protective coating would not be installed.
In lieu of the original corrective actions, CENG indicated that the horizontal section of
pipe would be inspected each refueling outage. The vertical piping would not be
inspected. These corrective actions were based, in part, on results from inspections
conducted during the 2013 N1R22 that identified rust accumulation only on the
horizontal sections of pipe. The enforcement aspects of this issue are discussed in
section 4OA3.2 of this report. The inspectors did not identify any new issues during the
review of this revised LER. This LER is closed.
.4 (Closed) LER 05000220/2012-007-00: High-Pressure Coolant Injection System Logic
Actuation Following an Automatic Turbine Trip Signal due to High Reactor Water Level
On November 6, 2012, while Unit 1 was in cold shutdown, an unexpected rise in reactor
water level occurred causing an automatic turbine trip signal and actuation of the
high-pressure coolant injection initiation logic. Operators immediately closed the 12
Enclosure
49
feedwater pump discharge blocking valve and stabilized reactor water level, stopping the
transient. At Unit 1, high-pressure coolant injection is a mode of operation of the
condensate and feedwater system that utilizes the condensate storage tanks, main
condenser hotwell, two condensate pumps, two feedwater booster pumps, and two
motor-driven feedwater pumps. The rise in reactor water level resulted from the 12
feedwater flow control valve (FCV) FCV-29-137 unexpectedly failing partially open when
instrument air was removed from the valve during a tagout in preparation for
maintenance on the valve. FCV-29-137 has a series of lockup valves that are designed
to hold the FCV stem in position in the event instrument air is lost. CENG determined
FCV-29-137 partially opened due to a degraded top cylinder lockup valve O-ring. The
enforcement aspects of this issue are discussed in NRC Integrated Inspection Report 05000220/2013002, Section 1R22. The inspectors did not identify any new issues
during the review of the LER. This LER is closed.
.5 (Closed) LER 05000410/2013-001-00: Reactor Core Isolation Cooling System Isolation
Due to a Temperature Switch Unit Failure
On January 23, 2013, at 3:16 p.m., Unit 2 was operating at 100 percent power when an
unexpected isolation signal for containment isolation valves in the RCIC and RHR
system occurred due to a failure of a RB general area temperature switch
(2RHS*TS85A). The isolation resulted in the RCIC system being unavailable for
injection into the reactor vessel if called upon during an event. The affected RHR
isolation valves were already in the closed position which is their normal position during
power operation. The failure also occurred concurrently with the HPCS system being
inoperable for planned surveillance testing. With both RCIC and HPCS inoperable,
high-pressure coolant makeup capability was lost. At 3:50 p.m., HPCS was restored
and declared operable. Temperature switch 2RHS*TS85A was replaced at 11:04 p.m.,
and on January 24, at 1:17 a.m., RCIC was declared operable. The cause of the
temperature switch failure was determined to be age-related capacitor degradation. The
enforcement aspects of this issue are discussed in NRC Integrated Inspection Report 05000410/2013002, Section 1R12. The inspectors did not identify any new issues
during the review of the LER. This LER is closed.
.6 (Closed) LER 05000410/2013-002-00: Failure of High-Pressure Core Spray System
Pressure Pump Due to Motor Winding Failure
On February 28, 2013, Unit 2 was operating at 100 percent power when the HPCS
system pressure pump failed. At the time of the failure, the HPCS system was
inoperable for planned maintenance. The pump failure was due to turn-to-turn short in
the motor winding. The HPCS system pressure pump is designed to maintain a positive
pressure on the HPCS discharge header to prevent voids from forming. CENG replaced
the HPCS pressure pump motor and returned the HPCS system to an operable status
on March 6. The HPCS system discharge piping remained full during the period when
the pressure pump was OOS. The inspectors reviewed the maintenance history of the
HPCS pressure pump motor and determined that when the motor bearings were
replaced in January 2011, the work order documented a satisfactory visual inspection
and meggar testing of the motor windings. The inspectors reviewed the LER and
determined that no findings or violations of NRC requirements were identified. This LER
is closed.
Enclosure
50
4OA6 Meetings, Including Exit
Exit Meeting
On July 25, 2013, the inspectors presented the inspection results to Mr. Christopher
Costanzo, Site Vice President, and other members of the NMPNS staff. The inspectors
verified that no propriety 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
C. Costanzo, Vice President
J. Stanley, Plant General Manager
P. Bartolini, Supervisor, Design Engineering
K. Clark, Director, Security
S. Dack, Seasonal Readiness Coordinator / Cycle Manager
J. Dean, Supervisor, Quality Assurance
S. Dhar, Design Engineering
J. Dosa, Director, Licensing
J. Gillard, Emergency Preparedness Analyst
J. Holton, Supervisor, Systems Engineering
G. Inch, Principle Engineer,
M. Kunzwiler, Security Supervisor
J. Leonard, Supervisor Design Engineering
C. McClay, Senior Engineer
F. Payne, Manager, Operations
P. Politzi, Work Week Manager
J. Reid, Design Engineer
B. Scaglione, System Engineer
J. Schulz, System Engineer
M. Shanbhag, Licensing Engineer
R. Staley, System Engineer
T. Syrell, Manager, Nuclear Safety and Security
J. Thompson, General Supervisor, Mechanical Maintenance
A. Verno, Director, Emergency Preparedness
Attachment
A-2
LIST OF ITEMS OPENED, CLOSED, DISCUSSED, AND UPDATED
Opened
05000220/2013003-04 AV Improper Bus Restoration Results in a Loss of
Shutdown Cooling (Section 4OA3)
Opened/Closed
05000410/2013003-01 NCV Failure to Follow Containment Isolation System
Surveillance Procedure Resulting in Isolation of the
Reactor Coolant Isolation Cooling System
(Section 1R22)05000410/2013003-02 NCV Inadequate Procedural Implementation for Battery
Cell Replacement (Section 4OA2)05000410/2013003-03 NCV Inadequate Design Control for Battery Sizing
Calculation (Section 4OA2)05000220/2013003-05 NCV Containment Leakage Exceeds Technical
Specification 3.3.3 Limits (Section 4OA3)
Closed
05000220/2012005-03 URI Assessment of Containment Leakage Due to
Containment Isolation Valve Failure (4OA3)
05000220/2012-006-00 and LER Technical Specification Required Shutdown Due
05000220/2012-006-01 to Containment Leakage (Section 4OA3)
05000220/2012-007-00 LER High-Pressure Coolant Injection System Logic
Actuation Following an Automatic Turbine Trip
Signal Due to High Reactor Water Level
(Section 4OA3)
05000410/2013-001-00 LER Reactor Core Isolation Cooling System Isolation
Due to a Temperature Switch Unit Failure
(Section 4OA3)
05000410/2013-002-00 LER Failure of High-Pressure Core Spray System
Pressure Pump Due to Motor Winding Failure
(Section 4OA3)
Attachment
A-3
LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection
Procedures
N1-OP-64, Meteorological Monitoring, Revision 00603
N2-OP-102, Meteorological Monitoring, Revision 01103
N2-OP-102, Attachment 3, Hot Weather Preparation Checklist, Revision 01102
NAI-PSH-11, Seasonal Readiness Program, Revision 00700
Condition Reports
CR-2010-008430 CR-2011-010519 CR-2012-004448
CR-2011-008564 CR-2012-001034 CR-2012-007341
CR-2011-009058 CR-2012-002008 CR-2013-000154
CR-2011-009946 CR-2012-004258
Work Orders
WO C90679919 WO C91901545 WO C92110489
WO C91178423 WO C91919260 WO C92116209
WO C91425002 WO C91920244 WO C92133487
WO C91570604 WO C91966877 WO C92135500
WO C91570606 WO C92033133 WO C92139868
WO C91711577 WO C92008152 WO C92154168
WO C91847825 WO C92008169 WO C92156668
WO C91860534 WO C92015166 WO C92156894
WO C91862547 WO C92044771 WO C92161257
WO C91862559 WO C92067054 WO C92221738
WO C91883258 WO C92073630 WO C92226912
WO C91883511 WO C92073671 WO C92285675
WO C91883613 WO C92073704 WO C92292596
Miscellaneous
Diesel Trend Analysis
Summer Readiness Status, Attachment 1
System Seasonal Readiness Evaluations, Attachment 2
Unit 1 Scheduler Evaluation for Summer Readiness from June 15 to September 15
Unit 2 Scheduler Evaluation for Summer Readiness from June 15 to September 15
Section 1R04: Equipment Alignment
Procedures
N1-OP-13, Emergency Cooling System, Revision 03700
N1-OP-48, Control Room Ventilation System, Revision 02400
NIP-OUT-01, Shutdown Safety, Revision 03700
Attachment
A-4
Condition Reports
CR-2013-004333
CR-2013-004347
Drawings
B-69017-C, Emergency Condenser Number 11 Steam Flow, Revision 1
C-180007-C, Reactor Core Spray Piping and Instrumentation Drawing (P&ID), Revision 58
C-18008-C, Spent Fuel Storage Pool Filtering and Cooling System, Revision 38
C-18030-C, Fire Protection Water System, Revision 38
C-18047-C, Control Room Heating Ventilation and Air Conditioning System, Revision 48
C-181017-C, Emergency Cooling System, Revision, Revision 55
Miscellaneous
Plant Configuration Change 1M00888
Section 1R05: Fire Protection
Procedure
N1-PFP-0101, Unit 1 Pre-Fire Plans, Revision 00200
Condition Report
CR-2013-002902
Miscellaneous
USAR Section 10, Revision 16
Section 1R07: Heat Sink Performance
Procedure
N1-ST-Q25, Emergency Diesel Generator Cooling Water Quarterly Test, Revision 02201
Work Order
WO C91454468
Section 1R08: In-Service Inspection
Procedures
NDEP-PT-3.00, Liquid Penetrant Examination, Revision 01900
NDEP-UT-6.23, UT Examination of Ferritic Piping Welds, Revision 01100
NDEP-UT-6.24, UT Examination of Austenitic Piping Welds, Revision 01101
NDEP-VT-2.01, ASME Section XI Visual Examination, Revision 19
NDEP-VT-2.07, In-Vessel Visual Examination, Revision 1300
NIP-IIT-02, ASME Section XI Repair and Replacement Program, Revision 00701
SI-UT-130, Phased Array Ultrasonic Examination of Dissimilar Metal Welds, Revision 0
Condition Reports
CR-2012-000816
Attachment
A-5
CR-2012-003805
CR-2012-010291
CR-2013-000506
CR-2013-001573
CR-2013-002975
CR-2013-002977
CR-2013-002978
CR-2013-003442
Drawing
C-18009, Reactor Water Cleanup P&ID, Revision 60, Sheet 1
Work Order
WO C92260831
NDE Records
BOP-UT-13-014, UT Calibration/Thickness Examination Records of RBCLC System Piping to
Recirculation Pump 11 Motor MOT-32-187, dated April 21, 2013
BOP-UT-13-015, UT Calibration/Thickness Examination Records of RBCLC System Piping to
Recirculation Pump 12 Motor MOT-32-188, dated April 21, 2013
BOP-UT-13-016, UT Calibration/Thickness Examination Records of RBCLC System Piping to
Recirculation Pump 13 Motor MOT-32-189, dated April 21, 2013
BOP-UT-13-017, UT Calibration/Thickness Examination Records of RBCLC System Piping to
Recirculation Pump 14 Motor MOT-32-190, dated April 21, 2013
BOP-UT-13-018, UT Calibration/Thickness Examination Records of RBCLC System Piping to
Recirculation Pump 15 Motor MOT-32-191, dated April 21, 2013
BOP-UT-13-021, UT Calibration/Thickness Examination Records of General Corrosion of
RBCLC System Piping Inside U1 Drywell 225 Feet Elevation, dated April 24, 2013
ISI-PT-13-003, Liquid Penetrant Examination Record of Branch Connection - Decontamination
Port Weld 32-WD-011 on Recirculation System Suction Piping, dated April 24, 2013
ISI-PT-13-004, Liquid Penetrant Examination Record of Branch Connection - Decontamination
Port Weld 32-WD-091 on Recirculation System Suction Piping, dated April 24, 2013
ISI-UT-13-032, UT Calibration/Examination Records of Branch Connection - Decontamination
Port Weld 32-WD-051 on Recirculation System Suction Piping, dated April 22, 2013
ISI-UT-13-033, UT Calibration/Examination Records of 12-Inch Diameter Emergency Condenser
Supply Piping, Pipe-to-Pipe Weld 39-WD-108, dated April 24, 2013
ISI-UT-13-034, UT Calibration/Examination Records of 12-Inch Diameter Emergency Condenser
Supply Piping, Pipe-to-Tee Weld 39-WD-109, dated April 24, 2013
ISI-UT-13-035, UT Calibration/Examination records of 12-Inch Diameter Emergency Condenser
Supply Piping, Tee-to-Pipe Weld 39-WD-110, dated April 24, 2013
ISI-UT-13-036, UT Calibration/Examination Records of 12-Inch Diameter Emergency Condenser
Supply Piping, Pipe-to-Elbow Weld 39-WD-112, dated April 20, 2013
NMP U1 33-WD-046, Phased Array UT Calibration/Examination Records of 6-Inch Diameter
RBCLC Pipe-to-Pipe DM Weld, dated April 29, 2013
UT Calibration/Examination Records of Unit 1 32-WD-042, N2A Safe End-Nozzle DM Weld,
Phased Array Ultrasonic Examination Record, dated April 30, 2013
UT Calibration/Examination Records of Unit 1 32-WD-042, N2A Safe End-to-Nozzle DM Weld
on Recirc Discharge, Phased Array Ultrasonic Examination Record, dated April 30, 2013
UT Calibration/Examination Records of Unit 1 32-WD-082, N2B Safe End-to-Nozzle DM Weld
on Recirc Discharge, Phased Array Ultrasonic Examination Record, dated April 30, 2013
Attachment
A-6
UT Calibration/Examination Records of Uni5 1 32-WD-122, N2C Safe End-to-Nozzle DM Weld
on Recirc Discharge, Phased Array Ultrasonic Examination Record, dated April 30, 2013
UT Calibration/Examination Records of Unit 1 32-WD-164, N2D Safe End-to-Nozzle DM Weld
on Recirc Discharge, Phased Array Ultrasonic Examination Record, dated April 30, 2013
UT Calibration/Examination Records of Unit 1 32-WD-208, N2E Safe End-to-Nozzle DM Weld
on Recirc Discharge, Phased Array Ultrasonic Examination Record, dated April 30, 2013
Miscellaneous
Audit Report SPC-12-01-N, Special Processes, Testing, & Inspection, dated November 28, 2012
ASME, 2004 Edition
Section 1R11: Licensed Operator Requalification Program and Licensed Operator
Performance
Procedure
CNG-OP-1.01-1000, Conduct of Operations, Revision 00900
Condition Reports
CR-2013-002697
CR-2013-002698
CR-2013-002647
CR-2013-002652
Section 1R12: Maintenance Effectiveness
Procedures
CNG-AM-1.01-1023, Maintenance Rule Program, Revision 00201
N2-OP-33, High Pressure Core Spray System, Revision 01201
N2-OSP-CSH-Q@002, HPCS Pump and Valve Operability and System Integrity Test,
Revision 00500
Condition Reports
CR-2011-006564 CR-2012-002176 CR-2012-009400
CR-2011-006930 CR-2012-002198 CR-2012-009982
CR-2011-007084 CR-2012-002249 CR-2012-010499
CR-2011-007313 CR-2012-002711 CR-2013-000159
CR-2011-007654 CR-2012-005017 CR-2013-000563
CR-2011-007830 CR-2012-005119 CR-2013-001491
CR-2011-009790 CR-2012-005999 CR-2013-001633
CR-2011-010817 CR-2012-006141 CR-2013-002768
CR-2012-000359 CR-2012-007193 CR-2013-002945
CR-2012-001459 CR-2012-008548 CR-2013-002969
CR-2012-001614 CR-2012-008816
Miscellaneous
ACE for CR-2011-006930
Attachment
A-7
ACE for CR-2012-002176
Eval-NMP-PRM-03046, (a)(1) Evaluation for 1-PRM-F01
Unit 1 Containment Spray System Health Report, 1st Quarter 2013
Unit 1 Neutron Monitoring System Health Report, 1st Quarter 2013
Unit 1 Service Water System Health Report, 1st Quarter 2013
Unit 2 High-Pressure Core Spray System Health Report, 1st Quarter 2013
Section 1R13: Maintenance Risk Assessments and Emergent Work Control
Procedures
CNG-MN-4.01-1004, On-Line T-Week Process, Revision 00302
N2-ISP-LDS-Q010, Reactor Building General Area Temperature Instrument Channel Functional
Test, Revision 00102
N2-OP-71D, Uninterruptible Power Supplies, Revision 00800
N2-SOP-29.1, Reactor Recirculation Pump Seal Failure, Revision 00101
N2-SOP-97, Reactor Protection Systems Failures, Revision 00401
NIP-OUT-01, Shutdown Safety, Revision 03700
S-ODP-OPS-0122, Posting and Control of Protected Equipment during Online and Outage
Operations, Revision 00500
Condition Reports
CR-2013-002461
CR-2013-002916
CR-2013-002926
CR-2013-002958
CR-2013-002998
CR-2013-005021
CR-2013-005077
Work Orders
WO C90962110
WO C91488068
WO C90648733
Miscellaneous
Control Room Operator Logs for Tuesday April 16, 2013
NIP-OUT-01, Shutdown Safety, Revision 03700, Attachment 1, Shutdown Safety Contingency
Plan (or Equivalent), Contingency Plan No. N1R22-003
NIP-OUT-01, Shutdown Safety, Revision 03700, Attachment 1, Shutdown Safety Contingency
Plan (or Equivalent), Contingency Plan No. N1R22-004
NIP-OUT-01, Shutdown Safety, Revision 03700, Attachment 1, Shutdown Safety Contingency
Plan (or Equivalent), Contingency Plan No. N1R22-005
Outage Control Center Logs for Tuesday April 16, 2013
Work Control Center Turnover Sheet for April 16, 2013, Days to Night.
Section 1R15: Operability Determinations and Functionality Assessments
Procedures
CNG-OP-1.01-1002, Conduct of Operability Determinations/Functionality Assessments,
Attachment
A-8
Revision 00200
N1-IPM-092-100, SRM Detector Drive Maintenance and Limit Switch Calibration, Revision 00700
N1-OP-18, Service Water System, Revision 02902
N1-OP-38A, Source Range Monitor, Revision 02000
N1-ST-C5, Secondary Containment and Reactor Building Emergency Ventilation System
Operability Testing, Revision 01600
N1-ST-C6, Source Range Monitor Operability Test, Revision 01100
Condition Reports
CR-2013-002637 CR-2013-003186 CR-2013-003698
CR-2013-002945 CR-2013-003445 CR-2013-004481
CR-2013-002969 CR-2013-003504 CR-2013-005079
CR-2013-002978 CR-2013-003520 CR-2013-004807
CR-2013-003107 CR-2013-003548
CR- 2013-003116 CR-2013-003567
CR-2013-003124 CR-2013-003589
Drawing
RX-147741, 10HN-18 Refinery Pump Elevation, Revision 0
Documents
UFSAR Section VI-2.0, Secondary Containment, Revision 15
UFSAR Section VII-3.0, Emergency Ventilation System, Revision 18
UFSAR Section VII-B, Containment Spray System, Revision 18
UFSAR Section XVI-2.0, Containment Spray System, Revision 20
Section 1R18: Plant Modifications
Procedure
N2-EPM-GEN-V786, MOD Actuator and Damper PM, Revision 00700
Condition Reports
CR-2013-002334
CR-2013-002303
Drawing
ECN Number ECP-12-000616-CN-004 LR18047C
Work Order
WO C919733104
Miscellaneous
ECP 12-000616, Installation of Bubble Tight Damper (BV-210-36)
ECP 13-000167, Installation of Replacement Pump for Unit 1 Service Water Radiation Monitor
ECP 13-000347, Temporary Change to Plug Hand Wheel Connection for 2HVP*AOD5A
Section 1R19: Post-Maintenance Testing
Attachment
A-9
Procedures
CNG-MN-4.01-1008, Pre-/Post-Maintenance Testing, Revision 00100
N1-FST-FPP-C005, Ventilation/Smoke Purge System, Revision 00400
S-EPM-GEN-063, MOV Diagnostic Testing, Revision 00700
Condition Reports
CR-2013-003051
CR-2013-003251
CR-2013-004003
CR-2013-004052
CR-2013-004177
CR-2013-004212
CR-2013-004253
Drawings
C-19410-C, Elementary Wiring Diagram 4.16 kV Emergency Power Boards and Diesel
Generators (102 and 103 Power Circuits), Revision 28, Sheet 1,
C-22277-C, 4160 Volt Power Board 102 Connection Diagram Unit 2-1, Diesel Generator 102,
Revision 09, Sheet 1
C-19437-C, Elementary Wiring Diagram 600V, Power Board 161B Control Circuits, Revision 25,
Sheet 2
C-19437-C, Elementary Wiring Diagram 600V, Power Board 161B Control Circuits, Revision 25,
Sheet 6
C-19017-C, Emergency Cooling System P&I Diagram, Revision 55, Sheet 1
Work Orders
WO C91473955
WO C91474635
WO C91973104
WO C92264883
WO C92279163
WO C92279776
Miscellaneous
ECP-13-000420-015-9, Removal and Replacement of Existing Cable 102-33 from EDG102 to
Power Board 102, Revision 0000
ECP-12-000575, Standard Spec for Electrical Installation Activities at NMP1, Revision 21.00
N21036, Limitorque Type SMB and SB Instruction and Maintenance Manual, NMPCNO:
N2L20000VALVOP004
SPEC NMP1-325M,Section II, Penetration Seals, Revision 1
Section 1R20: Refueling and Other Outage Activities
Procedures
CNG-OP-3.01-1000, Reactivity Management, Revision 00800
Attachment
A-10
N1-FHP-27C, Core Shuffle, Revision 00603
N1-FHP-25, General Description of Fuel Moves, Revision 02301
N1-OP-43C, Plant Shutdown, Revision 01200
N1-RESP-9, SRM Operability for Core Alterations, Revision 00001
N1-ST-V3, Rod Worth Minimizer Operability Test APRM/IRM Overlap Verification, Revision
01300
Condition Report
CR-2013-002793
Tagout
TO-30-0224
Miscellaneous
RFO22 Fuel Movement Instructions
Section 1R22: Surveillance Testing
Procedures
N1-ISP-LRT-TYC, Type C Containment Isolation Valve Local Leak Rate Test, Revision 00900
N1-ST-C5, Secondary Containment and Reactor Building Emergency Ventilation System
Operability Test, Revision 01600
N1-ST-Q15, Condensate Transfer System Operability Test, Revision 00703
N1-ST-Q3, High-Pressure Coolant Injection Pump and Check Valve Operability Test,
Revision 01300
N1-TSP-201-001, Integrated Leak Rate Test of Primary Containment Type A Test, Revision
00600
N2-CSP-GEN-D100, Reactor Water/Auxiliary Water Chemistry Surveillance, Revision 00601
N2-ISP-LDS-Q010, Reactor Building General Area Temperature Instrument Channel Functional
Test, Revision 00103
N22-CSP-W@101, Weekly Conductivity Monitor Channel Check, Revision 1
S-CAD-CHE-101, Chemistry Sample Conduct, Revision 0100
Condition Reports
CR-2013-002788
CR-2013-002637
Drawings
C-18013-C, Reactor Building Heating and Ventilation System, Revision 33
C-18014-C, Reactor Containment (Drywell and Torus) Inert Gas N2 Purge and Fill Drywell
Cooling System, Revision 58
Work Orders
WO C91214116
WO C92182070
Attachment
A-11
Miscellaneous
NUREG-1493, Performance-Based Containment Leak Test Program, September 1995
Section 1EP4: Emergency Action Level and Emergency Plan Changes
Procedures
EPIP-EPP-01, Classification of Emergency Conditions at Unit 1, Revision 23
EPIP-EPP-02, Classification of Emergency Conditions at Unit 2, Revision 22
EPMP-EPP-0101, Unit 1 Emergency Classification Technical Bases, Revision 01700
EPMP-EPP-0102, Unit 2 Emergency Classification Technical Bases, Revision 01900
Section 1EP6: Drill Evaluation
Procedure
EPIP-EPP-01, Classification of Emergency Conditions at Unit 1, Revision 02000
Section 2RS1: Radiological Hazard Assessment and Exposure Controls
Procedures
CNG-TR-1.01-1025, Radiation Protection Technician Training Program, Revision 00100
GAP-RPP-08, Control of High Locked High and Very High Radiation Areas, Revision 16
S-RAP-RPP-0103, Posting and Barricading Radiological Areas, Revision 02800
S-RAP-RPP-0201, Radiation Work Permit Initiation, Preparation, Control and Use,
Revision 02300
S-RAP-RPP-0801, High Locked High and Very High Radiation Area Monitoring and Control,
Revision 03000
S-RPIP-3.0, Radiological Surveys, Revision 01900
Condition Reports
CR-2013-002520
CR-2013-002781
CR-2013-003098
Audits, Self Assessments, and Surveillances
Q&PA Assessment Report 13-010, Assess Station Preparedness for Managing and Executing
N1R23
SA-2013-000005, Snapshot Assessment of 2012 4th Quarter Dose and Dose Rate Alarms
SA-2013-000034, Snapshot Assessment of Radiation Protection Job Hazard Analysis Process
Usage
Miscellaneous
BRAC Survey Trends in Discharge Piping Dose Rates, Unit 1, 1984 to 2013
BRAC Survey Trends in Recirc Suction Piping Dose Rates, Unit 1, 1984 to 2013
High Radiation Area/Locked High Radiation Area Gate Door Checklist, Unit 1, April 20, 2013
Personnel Qualification Form Verification, Employee Badge 38016, April 8, 2013
Personnel Qualification Form Verification, Employee Badge 38359, April 1, 2013
Personnel Qualification Form Verification, Employee Badge 4127, April 8, 2013
Personnel Qualification Form Verification, Employee Badge 4169, April 1, 2013
Personnel Qualification Form Verification, Employee Badge 4196, March 29, 2013
Attachment
A-12
Personnel Qualification Form Verification, Employee Badge 54337, February 25, 2013
RWP 113330H, RB 261 Reactor Water Cleanup Valve Work
RWP 113802H, Drywell Under-Vessel Work
RWP 113806H, Drywell In-Service Inspection
RWP 113810, Drywell General Scaffolding Activities
RWP 113815, RB 261 FAC In-Service Inspection
RWP 113890A, RB 340 Reactor Disassembly and Reassembly
RWP 113890B, RB 340 Underwater Work on Refuel Floor
RWP 113890E, RB 340 Reactor Cavity and Equipment Storage Pit Decon
RWP 113891, Spent Fuel Pool Gate Repair
Section 2RS2: Occupational ALARA Planning and Controls
Procedures
CNG-RP-1.01-1001, Station ALARA Committee, Revision 00000
CNG-RP-1.01-2003, Operational ALARA Planning and Controls, Revision 00000
N1-OP-34, Refueling Procedure (Includes Primary Chemistry Controls), Revision 03000
S-RAP-ALA-0101, Temporary Shielding, Revision 10
S-RAP-ALA-0102, ALARA Reviews, Revision 01500
Condition Reports
CR-2013-002267
CR-2013-003168
Self Assessment
SA-2012-000283, 4th Quarter 2012 ALARA Committee Effectiveness Review
Miscellaneous
5-Year Collective Radiation Exposure Reduction Plan, 2012 to 2016
ALARA Plan 2013-1-002, Drywell Under-Vessel Activities and Associated Activities N1R22,
April 10, 2013
ALARA Plan 2013-1-004, Drywell Operations and LLRT/ILRT Activities, April 10, 2013
ALARA Plan 2013-1-006, Drywell ISI Activities, April 10, 2013
ALARA Plan 2013-1-007, Recirc Pump Seals Replacement and Motor PMs (Numbers 11, 13 and
15), April 10, 2013
ALARA Plan 2013-1-010, Drywell Scaffold Activities, April 10, 2013
ALARA Plan 2013-1-014, Drywell Emergency Relief Valve and Pilot Valve Work Activities,
April 10, 2013
ALARA Plan 2013-1-024, Main Steam Isolation Valve 01-02 Stem Replacement Actuator
Remove/Replace and Testing, April 10, 2013
ALARA Plan 2013-1-029, Balance of Plant FAC Activities in RWCU HX Room and Valve Aisles,
April 10, 2013
ALARA Plan 2013-1-030, Refuel Floor Activities, dated April 10, 2013
ALARA Plan 2013-1-031, RWCU Miscellaneous Maintenance, PM, ST, Operations RFO 22,
April 10, 2013
ALARA Work In-Progress Review, 2013-1-006, Drywell ISI Activities, April 21, 2013
ALARA Work In-Progress Review, 2013-1-007, Recirc Pump Seals Replacement and Motor PMs,
April 22, 2013
ALARA Work In-Progress Review, 2013-1-010, Drywell Scaffold Activities, April 20, 2013
Attachment
A-13
ALARA Work In-Progress Review, 2013-1-011, Drywell Insulation, April 22, 2013
ALARA Work In-Progress Review, 2013-1-014, Drywell Emergency Relief Valve and Pilot Valve
Work Activities, April 22, 2013
ALARA Work In-Progress Review, 2013-1-024, Main Steam Isolation Valve 01-02 Stem
Replacement Actuator Remove/Replace and Testing, April 22, 2013
ALARA Work In-Progress Review, 2013-1-029, Balance of Plant FAC Activities in RWCU HX
Room and Valve Aisles, April 18, 2013
ALARA Work In-Progress Review, 2013-1-030, Refuel Floor Activities, April 20, 2013
Unit 1 Radiation Protection Pre-Outage Report, dated April 15, 2013
Section 2RS3: In-Plant Airborne Radioactivity Control and Mitigation
Procedures
GAP-RPP-04, Respiratory Protection Program, Revision 11
N1-RTP-76, Operation and Calibration of the Eberline PING-1A PING-1AMT Particulate Iodine
Noble Gas Monitor, Revision 02
S-RAP-RPP-0402, Selection and Issuance of Radiological Respiratory Protection Equipment,
Revision 12
S-RPIP-4.2, Respiratory Protection Quality Assurance Control Program, Revision 00200
S-RPIP-4.4, Maintenance Inspection and Testing of Respiratory Protection Equipment,
Revision 00700
S-RPIP-4.5, Use of Respiratory Protection Equipment, Revision 09
S-RPIP-6.0, Control and Use of HEPA Vacuum Cleaners and Portable HEPA Ventilation Units,
Revision 00300
Condition Reports
CR-2013-002816
CR-2013-002947
Self Assessment
SA-2011-000164, Radiological Respiratory Protection Program, November 18, 2011
Miscellaneous
Air Sample Unit 1 RB 340 Edge of Cavity, April 15, 2013, 5:43 a.m.
Air Sample Unit 1 RB 340 Edge of Cavity, April 15, 2013, 7:14 a.m.
Air Sample Unit 1 RB 340 Refuel Floor during Silver Dollar Installation, April 15, 2013, 9:20 p.m.
Air Sample Unit 1 RB 340 Refuel Floor during Stud Removal, April 17, 2013, 12:20 p.m.
HEPA Ventilation Log, dated April 23, 2013
Unit 1 System Health Report for 1st Quarter Control Room Ventilation, dated April 10, 2013
Unit 1 System Health Report for 1st Quarter RB Ventilation, dated April 10, 2013
Vacuum Cleaner Issue Log, dated April 23, 2013
Section 2RS4: Occupational Dose Assessment
Procedures
CNG-RP-1.01-2002 Effective Dose Equivalent - External, Revision 00000
CNG-RP-1.01-3002, Sampling and Analysis for 10 CFR 61 Waste Classification, Revision 00000
GAP-RPP-07, Internal and External Dosimetry Program, Revision 02100
S-RAP-ALA-0103, Dosimetry and Radiological Engineering Evaluations, Revision 00900
S-RPIP-4.6, DAC Hour Tracking and Estimating Internal Exposure, Revision 00500
Attachment
A-14
S-RPIP-5.5, Processing and Evaluating Personnel Contamination, Revision 01800
S-RPIP-5.7, Bioassay and Internal Dose Assessment, Revision 00900
S-RPIP-5.20, Dosimetry Program Quality Assurance, Revision 00800
S-RPIP-5.25, Exposure Evaluation Reports, Revision 01000
Condition Reports
CR-2013-002474
CR-2013-002678
CR-2013-002974
CR-2013-003247
CR-2013-003374
CR-2013-003350
CR-2013-003413
Miscellaneous
Oak Ridge Associated University E-mail Y. McCormick to A. Moisan RE: REIRS Data Verification,
dated April 1, 2013
Sentinel Report on Personnel with Dose Greater Than 400 mrem, dated April 22, 2013
S-RPIP-5.5 Attachment 1 Contamination Occurrence Report Number 1-13-RFO-003, dated
April 24, 2013
Section 2RS7: Radiological Environmental Monitoring Program
Procedures
CNG-EV-1.01-1000, Radiological Environmental Monitoring Program, Revision 001000
NLAP-ENV-400, Radiological Environmental Monitoring Program Land Use Census,
Inter-laboratory Comparison Program and Reporting, Revision 00.00
S-ENVSP-3, Radiological Sample Collection, Processing, and Shipment Land Use Census
Quality Control (Vendor Procedure), Revision 06.00
S-ENVSP-3.1, Milk Animal Census and Milk Sample Collection, Revision 01.00
S-ENVSP-3.2, Garden/Irrigation Census and Food Product (Vegetation and Irrigation Crop)
Sample Collection, Revision 02.00
S-ENVSP-3.3, Nearest Meat Animal Census and Meat, Poultry, and Egg Sample Collection,
Revision 01.00
S-ENVSP-3.4, Soil Sample Collection, Revision 01.00
S-ENVSP-3.5, Fish Sample Collection, Revision 01.00
S-ENVSP-3.6, Shoreline Sediment and Cladophora Sample Collection, Revision 01.00
S-ENVSP-3.7, Nearest Residence Census, Revision 00.00
S-ENVSP-4.1, TLD/OSL Preparation, Collection and Analysis, Revision 01400.00
S-ENVSP-4.2, Environmental Air Monitoring Sample Collection, Revision 01001.00
S-ENVSP-4.3, Environmental Air Monitoring Station Inspection and Maintenance,
Revision 00600.00
S-ENVSP-4.4, Environmental Surface Water Sample Collection and Compositing,
Revision 00900.00
S-ENVSP-12, Environmental Surveillance Quality Assurance/Quality Control Program,
Revision 001100.00
S-ENVSP-15, Sampling and Analysis for Unmonitored Pathways, Revision 01300.00
S-ENVSP-16, Sampling and Analysis of Monitoring Wells, Revision 00500.00
S-ENVSP-18, Environmental Data Review, Revision 01000.00
S-IPM-MET-001, Meteorological Monitoring System Equipment Check, Revision 00200.00
Attachment
A-15
S-IPM-MET-201, Dew Point Calibration, Revision 00100.00
S-IPM-MET-301, Barometric Pressure Calibration, Revision 03.00
S-IPM-MET-401, Precipitation Gauge Calibration, Revision 02.00
S-IPM-MET-601, Main Meteorological Tower 30 Foot Wind Speed and Direction Calibration,
Revision 00100.00
S-IPM-MET-602, Main Meteorological Tower 100 Foot Wind Speed and Direction Calibration,
Revision 00400.00
S-IPM-MET-603, Main Meteorological Tower 200 Foot Wind Speed and Direction Calibration,
Revision 00100.00
S-IPM-MET-611, Backup Meteorological Tower Wind Speed and Direction Calibration,
Revision 00200.00
S-IPM-MET-621, Inland Meteorological Tower Wind Speed and Direction Calibration,
Revision 00100.00
S-IPM-MET-701, Temperature and Delta Temperature Instrument Calibration,
Revision 00200.00
S-MET-ENV-01, Maintenance of Meteorological Monitoring Program, Revision 00100.00
S-MET-ENV-0002, Meteorological Data Verification and Edit, Revision 00600.00
S-MET-ENV-0003, Meteorological Monitoring Program Quality Assurance Quality Control,
Revision 00600.00
Condition Reports
CR-2012-000632 CR-2012-005817 CR-2012-010132
CR-2012-000664 CR-2012-006057 CR-2013-000603
CR-2012-000734 CR-2012-007114 CR-2013-001001
CR-2012-001143 CR-2012-007684
CR-2012-001488 CR-2012-009863
Work Orders
WO C91660878
WO C91875097
Audits, Self Assessments, and Surveillances
DTE Energy NAQA-12-0036, Audit 12-006 of Environmental Dosimetry Company, July 3, 2012
Entergy CR-LO-JAFLO-2012-00045, Radiological Environmental Monitoring Program Focused
Self Assessment, February 20 to 27, 2013
NUPIC Audit 22873, GEL Laboratories, LLC, Analytical Laboratory Services, December 13, 2011
Miscellaneous
2011 Annual Report, Meteorological Monitoring Program, Murray and Trettel, Inc., Palatine, IL
2012 Annual Quality Assurance Status Report, Environmental Dosimetry Company, dated
March 13, 2013
2012 Inter-laboratory Comparison Report, Eckert and Zeigler, dated March 29, 2013
2012 Land Use Census Summary Report, dated October 25, 2012
DVP-04.01, Environmental Laboratory Quality Assurance/Quality Control Program, Revision 4
EN-CY-102, Laboratory Analytical Quality Control, Revision 4
James A. FitzPatrick Environmental Laboratory Quality Assurance Report, January to
December 2011
Licensee Event Number 48901, Power Lost to Meteorological Instrumentation, dated April 9, 2013
Quality Assurance Topical Report, dated December 11, 2011
Attachment
A-16
Radiological Environmental Operating Report January to December, 2012, dated May 15, 2013
Radiological Engineering Evaluation Number C-99-011, Revision 7, 10 CFR 50.75(g) Record -
Unit 1 TB Roof Replacement, dated September 7, 2012
Radiological Engineering Evaluation Number C-99-011, Revision 8, 10 CFR 50.75(g) Record -
Elevated Tritium Concentration in Screen House In-Leakage, dated January 27, 2013
S-ENVSP-4.4 Attachment 5A L/S 7523 Sample Pump Control Setting Determination, Serial
Number L03004172, NRG Oswego Steam Station, dated August 14, 2009
S-ENVSP-4.4, Attachment 5A, L/S 7523 Sample Pump Control Setting Determination, Serial
Number L04004587, Unit 1 Intake Canal, dated April 20, 2009
S-ENVSP-4.4, Attachment 5A, L/S 7523 Sample Pump Control Setting Determination, Serial
Number L04004590, Unit 1 Intake Canal, dated April 20, 2009
Tektronix Certificate of Calibration 6776890, American Meter Mass Flow Meter Number 10429,
dated November 16, 2012
Tektronix Certificate of Calibration 6104009, American Meter Mass Flow Meter Number 10436,
dated April 20, 2012
Tektronix Certificate of Calibration 6780305, American Meter Mass Flow Meter Number 10458,
dated November 17, 2012
Tektronix Certificate of Calibration 6114558, American Meter Mass Flow Meter Number 10870,
dated April 23, 2012
Tektronix Certificate of Calibration 6380789, American Meter Mass Flow Meter Number 10899,
dated July 18, 2012
Unit 1 ODCM, Revision 34
Unit 1 Radioactive Effluent Release Report, January to December 2012, dated May 1, 2013
Unit 2 ODCM, Revision 35
Unit 2 UFSAR Chapter 2.3, Meteorology, Revision 19, October 2010
Section 4OA1: Performance Indicator Verification
Procedures
N2-CSP-GEN-D100, Reactor Water/Auxiliary Water Chemistry Surveillance, Revision 00601
N22-CSP-W@101, Weekly conductivity Monitor Channel Check, Revision 1
S-CAD-CHE-101, Chemistry Sample Conduct, Revision 01100
Miscellaneous
Nuclear Energy Institute 99-02, Regulatory Assessment Performance Indicator Guideline,
Revision 6
Section 4OA2: Problem Identification and Resolution
Procedures
CENG-AM-1.01-1005, Engineering Role and Responsibilities/Expectations, Revision 00303
CNG-CA-1.01-1004, Root Cause Analysis, Revision 00802
CNG-CA-2.01-1000, Self-Assessment and Benchmarking Process, Revision 00700
CNG-MN-4.01-1001, Work Order Execution and Closure Process, Revision 00401
CNG-MN-1.01-1000, Conduct of Maintenance, Revision 00200
N2-EMP-GEN-673, 24/48 VDC and 125 VDC Batteries - Cell and Connector Replacement,
Revision 00400
NPAP-INV-220, Storage and Handling of Material, Revision 01001
Nine Mile Point Station Policy Number 22, Work Document Quality, Revision 0
Attachment
A-17
Procedure Review Briefing Sheet CNG-HU-1.01-1001 HU Tools and Verification Process
Understanding Human Behavior and Error, Human Reliability Associates, David Embrey
Condition Reports
CR-1997-001696 CR-2012-000060 CR-2012-009469
CR-2001-005920 CR-2012-001137 CR-2012-010774
CR-2005-003461 CR-2012-001138 CR-2012-010907
CR-2007-007514 CR-2012-001139 CR-2013-001159
CR-2010-001220 CR-2012-001315 CR-2013-002102
CR-2010-001987 CR-2012-001316 CR-2013-002360
CR-2010-003899 CR-2012-002716 CR-2013-002443
CR-2010-007337 CR-2012-003724 CR-2013-003207
CR-2011-005737 CR-2012-004600 CR-2013-003357
CR-2011-007171 CR-2012-005362 CR-2013-005074
CR-2011-007269 CR-2012-005365 CR-2013-005117
CR-2011-007655 CR-2012-006030 CR-2013-005228
CR-2011-009896 CR-2012-006242 CR-2013-005235
CR-2011-010906 CR-2012-006823 CR-2013-005245
CR-2011-010953 CR-2012-007085
CR-2011-011006 CR-2012-007765
Drawings
3.N2.1-E21.1, One Line Diagram 125 VDC Control Bus, Revision 14
EE-1CA, One Line Diagram Emergency and Vital Bus Power Distribution Unit 2, Revision 14
EE-1CM, 125 VDC One Line Diagram Emergency Switchgear 2BYS*SWG002A, Revision 19
EE-1CN, 125 VDC One Line Diagram Emergency Switchgear 2BYS*SWG002B, Revision 17
EE-MO1F, Plant Master One Line Diagram Emergency and Normal 125V and 24/48VDC Unit 2,
Revision 8
Work Orders
WO C92017475
WO C92036878
Miscellaneous
CR Search for RCS*MOV18, Excessive Unidentified Leakage, and TS Required Shutdown for
January 1, 2000, until April 25, 2013
Design Engineering Request NM-2001-5894
Equipment Reliability Return to Excellence Plan
Equivalency Evaluation Number 00230 for RCS*MOV 10A&B and RCS*MOV 18A&B, dated
April 4, 2002
GE SIL No. 620, BWR 5 and 6 Reactor Recirculation System Pump Discharge Gate Valve
N2-ESP-BYS-Q767, Quarterly Battery Surveillance Test, completed on August 16 and 31, 2012;
February 11, March 7, and May 28, 2013
N2-ESP-BYS-R685, Divisions I, II, and III Battery Modified Profile Test, completed on April 4
and 10, 2010; April 16, July 25, and November 28, 2012
Root Cause Analysis, Cross-Cutting Theme Exists in the Aspect of Human Performance,
Resources, Documentation H.2(c) dated January 18, 2013
Root Cause Analysis, Unit 1 SCRAM due to Turbine Trip on May 2, 2011, dated
September 16, 2011
Attachment
A-18
Timeline of RCS*MOV 18A Problems
Unit 1 DEP System Health Report, 1st and 2nd Quarters 2013
Unit 2 DEP System Health Report, 1st and 2nd Quarters 2013
Valve Packing Data Sheet for RCS*MOV 10A and B
Valve Packing Data Sheet for RCS*MOV 18A and B
Vendor Manuals
35.40, Specifications Nuclear Class 1E Flooded Batteries GNB, dated August 2002
RS-1476, Stationary Battery and Vented Cell Installation and Operating Instructions C&D
Technologies, dated 2009
Calculation
EC-145, Verification of Adequacy of Division III Battery 2BYS*BAT2C, Revision 2
Section 4OA3: Follow-up of Events and Notices of Enforcement Discretion
Procedures
N1-OP-47A, 125 VDC Power System, Revision 02500
N1-SOP-47A.1, Loss of DC, Revision 00101
N1-SOP-6.1, Loss of SFP/RX Cavity Level/Decay Heat Removal, Revision 00501
N1-ST-R2, LOCA and EDG Simulated Auto Initiation Test, Revision 03201
N2-EMP-GEN-609, General Small Motor Maintenance, Revision 06
NIP-OUT-01, Shutdown Safety, Revision 03700
Condition Reports
CR-2013-001633
CR-2013-002916
CR-2013-002926
CR-2013-002958
CR-2013-002998
Miscellaneous
ACE for CR-2013-001633
CENG Safety Stand Down for April 16, 2013, Loss of Battery Bus 12 Event
Control Room Operator Logs for Tuesday, April 16, 2013
E191, NMPNS Specification for Safety-Related Motor Repairs, Revision 0
Outage Control Center Logs for Tuesday, April 16, 2013
PM Template for Small and Intermediate HP Motors
Unit 1 Station Alarm Log for Tuesday, April 16, 2013
Work Control Center Turnover Sheet for April 16, 2013, Days to Night
Attachment
A-19
LIST OF ACRONYMS
10 CFR Title 10 of the Code of Federal Regulations
AC alternating current
ADAMS Agencywide Documents Access and Management System
ALARA as low as reasonably achievable
ASME American Society of Mechanical Engineers
BWR boiling-water reactor
CAP corrective action program
CENG Constellation Energy Nuclear Group, LLC
DC direct current
ECCS emergency core cooling system
ECP engineering change package
EDG emergency diesel generator
ERV electro-matic relief valve
FA fire area
FAC flow accelerated corrosion
FCV flow control valve
HPCS high-pressure core spray
I&C instrumentation and control
IEEE Institute of Electrical and Electronics Engineers
IMC Inspection Manual Chapter
ISI inservice inspection
kV kilovolt
LER licensee event report
LOCA loss of coolant accident
LOOP loss of offsite power
NDE nondestructive examination
NCV non-cited violation
NMPNS Nine Mile Point Nuclear Station, LLC
NRC Nuclear Regulatory Commission
ODCM offsite dose calculation manual
psig pounds per square inch gauge
RB reactor building
RCIC reactor core isolation cooling
REMP radiological environmental monitoring program
RG regulatory guide
RPT radiation protection technician
RWP radiation work permit
SDP significance determination process
SFP spent fuel pool
SSC structure, system, and component
Enclosure
A-20
ST surveillance testing
TLD thermo luminescent dosimeter
TS technical specification
UFSAR Updated Final Safety Analysis Report
UT ultrasonic testing
VDC volts direct current
Attachment