ML15245A785
| ML15245A785 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 09/02/2015 |
| From: | Christine Lipa NRC/RGN-III/DRS/EB2 |
| To: | Gardner P Northern States Power Co |
| References | |
| IR 2015007 | |
| Download: ML15245A785 (33) | |
See also: IR 05000263/2015007
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION III
2443 WARRENVILLE RD. SUITE 210
LISLE, IL 60532-4352
September 2, 2015
Mr. Peter A. Gardner
Site Vice President
Monticello Nuclear Generating Plant
Northern States Power Company, Minnesota
2807 West County Road 75
Monticello, MN 55362-9637
SUBJECT: MONTICELLO NUCLEAR GENERATING PLANT - NRC COMPONENT DESIGN
BASES INSPECTION (INSPECTION REPORT 05000263/2015007)
Dear Mr. Gardner:
On July 24, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed a Component
Design Bases Inspection at your Monticello Nuclear Generating Plant. The enclosed report
documents the inspection findings, which were discussed on July 24, 2015, with you and other
members of your staff.
Based on the results of this inspection, two NRC-identified findings of very low safety
significance were identified. The findings involved violations of NRC requirements. However,
because of their very low safety significance, and because the issues were entered into your
Corrective Action Program, the NRC is treating the issues as Non-Cited Violations (NCVs) in
accordance with Section 2.3.2 of the NRC Enforcement Policy.
If you contest the subject or severity of these NCVs, you should provide a response within
30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear
Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with
copies to the Regional Administrator, Region III; the Director, Office of Enforcement, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident
Inspector at Monticello Nuclear Generating Plant.
In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public
Inspections, Exemptions, Requests for Withholding, of the NRC's "Rules of Practice," a copy
of this letter, its enclosure, and your response (if any) will be available electronically for public
P. Gardner
-2-
inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS)
component of the NRC's Agencywide Documents Access and Management System (ADAMS).
ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html
(the Public Electronic Reading Room).
Sincerely,
/RA/
Christine A. Lipa, Chief
Engineering Branch 2
Division of Reactor Safety
Docket No. 50-263
License No. DPR-22
Enclosure:
Inspection Report 05000263/2015007;
w/Attachment: Supplemental Information
cc w/encl: Distribution via LISTSERV
Enclosure
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No:
50-263
License No:
Report No:
Licensee:
Northern States Power Company, Minnesota
Facility:
Monticello Nuclear Generating Plant
Location:
Monticello, MN
Dates:
June 22, 2015, through July 24, 2015
Inspectors:
A. Dunlop, Senior Engineering Inspector, Lead
B. Jose, Senior Engineering Inspector, Electrical
M. Holmberg, Senior Engineering Inspector, Mechanical
C. Phillips, Operations Inspector
S. Gardner, Electrical Contractor
G. Gardner, Mechanical Contractor
Observer:
I. Khan, Engineering Inspector, Electrical
Approved by:
Christine A. Lipa, Chief
Engineering Branch 2
Division of Reactor Safety
2
SUMMARY
Inspection Report 05000263/2015007; 06/22/2015 - 07/24/2015; Monticello Nuclear Generating
Plant; Component Design Bases Inspection.
The inspection was a 3-week onsite baseline inspection that focused on the design of
components. The inspection was conducted by regional engineering inspectors and two
consultants. Two Green findings were identified by the inspectors. The findings were
considered Non-Cited Violations (NCVs) of U.S. Nuclear Regulatory Commission (NRC)
regulations. The significance of inspection findings is indicated by their color (i.e., greater
than Green, or Green, White, Yellow, Red), and determined using Inspection Manual Chapter
(IMC) 0609, Significance Determination Process, dated April 29, 2015. Cross-cutting
aspects are determined using IMC 0310, Aspects Within the Cross-Cutting Areas, dated
December 4, 2014. All violations of NRC requirements are dispositioned in accordance with
the NRCs Enforcement Policy, dated July 9, 2013. The NRC's program for overseeing the
safe operation of commercial nuclear power reactors is described in NUREG 1649, Reactor
Oversight Process, Revision 5, dated February 2014.
Cornerstone: Mitigating Systems
Green. The inspectors identified a finding having very-low safety significance, and
an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50,
Appendix B, Criterion III, Design Control, for the failure to assure the nitrogen supply
for the alternate nitrogen (AN2) system was controlled as safety-related in system
specifications, drawings, procedures, and instructions. Specifically, the licensee did not
confirm effective quality assurance controls were in place to ensure the bottled nitrogen
was acceptable to support the safety-related functions of this system. The licensee
entered this finding into the Corrective Action Program (CAP), and subsequently
contacted the commercial nitrogen gas supplier to confirm that the vendors quality
controls provided a sufficient basis to conclude that the AN2 system was operable.
The finding was determined to be more than minor because if left uncorrected, the
issue had the potential to lead to a more significant safety concern. Specifically, if the
commercial (e.g., non-safety) gas supply vendor quality controls were not adequate to
ensure contaminants such as moisture or particulates were excluded from the nitrogen
gas bottles, it could potentially disable the AN2 systems capability to support manual
operation of safety relief valves during post loss-of-coolant-accident mitigation. The
inspectors did not identify a cross-cutting aspect associated with this finding as it did
not reflect current performance. (Section 1R21.3.b.(1))
Green. The inspectors identified a finding of very-low safety significance, and an
associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the
failure to assure measures were established for the selection and review for suitability
of application of materials, parts, equipment and processes that were essential to the
safety-related functions of structures, systems and components. Specifically, the
licensee failed to review for suitability of application of safety-related Agastat and
General Electric relays that had exceeded their service life, a condition non-conforming
to their design basis, to justify their continued service considering in-service
deterioration. The licensee previously entered this finding into the CAP, and
completed corrective actions to replace or evaluate some relays and implemented
a program to address the remaining relays in a timely manner.
3
The finding was determined to be more than minor because, if left uncorrected, the
issue had the potential to lead to a more significant safety concern. Specifically, these
safety-related relays were installed in protective circuits such as reactor protection
system, etc., and their failure could impact the proper operation of these protective
schemes. The inspectors did not identify a cross-cutting aspect associated with this
finding as it was not reflective of the licensees current performance.
(Section 1R21.3.b.(2))
4
REPORT DETAILS
1.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R21 Component Design Bases Inspection (71111.21)
.1
Introduction
The objective of the Component Design Bases Inspection (CDBI) is to verify that design
bases have been correctly implemented for the selected risk-significant components and
that operating procedures and operator actions are consistent with design and licensing
bases. As plants age, their design bases may be difficult to determine and an important
design feature may be altered or disabled during a modification. The Probabilistic Risk
Assessment (PRA) model assumes the capability of safety systems and components to
perform their intended safety function successfully. This inspectable area verifies
aspects of the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones
for which there are no indicators to measure performance.
Specific documents reviewed during the inspection are listed in the Attachment to this
report.
.2
Inspection Sample Selection Process
The inspectors used information contained in the licensees PRA and the Monticello
Standardized Plant Analysis Risk Model to identify internal flooding scenarios to
use as the basis for component selection. Based on these scenarios, a number of
risk-significant components, including those with large early release frequency (LERF)
implications, were selected for the inspection.
The inspectors also used additional component information such as a margin
assessment in the selection process. This design margin assessment considered
original design reductions caused by design modification, power uprates, or reductions
due to degraded material condition. Equipment reliability issues were also considered in
the selection of components for detailed review. These included items such as
performance test results, significant corrective actions, repeated maintenance activities,
Maintenance Rule (a)(1) status, components requiring an operability evaluation, system
health reports, and U.S. Nuclear Regulatory Commission (NRC) resident inspector input
of problem areas/equipment. Consideration was also given to the uniqueness and
complexity of the design, operating experience, and the available defense in depth
margins. A summary of the reviews performed and the specific inspection findings
identified are included in the following sections of the report.
The inspectors also identified procedures and modifications for review associated with
the selected components. In addition, the inspectors selected operating experience
issues associated with the selected components.
The inspection reviewed 19 samples (5 operating experience, 13 components, and
1 component with LERF implications) as defined in Inspection Procedure 71111.21 05.
5
.3
Component Design
a.
Inspection Scope
The inspectors reviewed the Updated Safety Analysis Report (USAR), Technical
Specifications (TS), design basis documents, drawings, calculations and other available
design basis information, to determine the performance requirements of the selected
components. The inspectors used applicable industry standards, such as the American
Society of Mechanical Engineers Code, Institute of Electrical and Electronics Engineers
(IEEE) Standards, and the National Electric Code, to evaluate acceptability of the
systems design. The NRC also evaluated licensee actions, if any, taken in response to
NRC issued operating experience, such as Bulletins, Generic Letters, Regulatory Issue
Summaries (RISs), and Information Notices (INs). The review was to verify that the
selected components would function as designed when required and support proper
operation of the associated systems. The attributes that were needed for a component
to perform its required function included process medium, energy sources, control
systems, operator actions, and heat removal. The attributes to verify that the component
condition and tested capability was consistent with the design bases and was
appropriate may include installed configuration, system operation, detailed design,
system testing, equipment and environmental qualification, equipment protection,
component inputs and outputs, operating experience, and component degradation.
For each of the components selected, the inspectors reviewed the maintenance history,
preventive maintenance activities, system health reports, operating experience-related
information, vendor manuals, electrical and mechanical drawings, and licensee
corrective action program documents. Field walkdowns were conducted for all
accessible components to assess material condition, including age-related degradation
and to verify that the as-built condition was consistent with the design. Other attributes
reviewed are included as part of the scope for each individual component.
The following 14 components (samples) were reviewed:
Non-Safeguards Diesel Generator (DG-13): The inspectors reviewed the fuel
capacity of the day tank, the procedures, and equipment required for refueling
the day tank to determine if the DG-13 would be able to meet its required
mission time. In addition, the inspectors reviewed monthly operability testing to
determine whether the DG-13 would perform as required. Maintenance records
and trends were also reviewed to verify reliability. The inspectors reviewed the
DG-13 ability to supply power for the safety-related inverter to Battery #13 in the
event of an extended station blackout (SBO) scenario. Generator loading was
reviewed for this scenario to ensure DG-13 was capable to supply the anticipated
load per the operating procedures. A walk through of this scenario with licensee
staff was conducted to ensure the operating procedure was adequate to perform
the intended operations.
Reactor Core Isolation Cooling Pump (P-207): The inspectors reviewed the
system hydraulic calculations such as, net positive suction head (NPSH) and
minimum required flow to ensure the pumps were capable of providing their
function. The inspectors also reviewed the vendor manual for the pump to
determine whether the pumps characteristics met the design basis requirements
and these requirements were accurately incorporated in reactor core isolation
6
cooling (RCIC) system inservice testing (IST) procedures. The IST results were
reviewed to assess potential component degradation and impact on design
margins. The operation of the pump from various suction sources was reviewed
to evaluate the pumps ability to provide the required flow from each source. The
inspectors reviewed the RCIC operation during SBO compared to how various
RCIC subcomponents were modeled in the battery sizing calculation to verify
RCIC subcomponent loading was conservative.
Reactor Core Isolation Cooling Minimum Flow Valve (CV-2104): The inspectors
reviewed the air-operated valve (AOV) calculations, including required thrust,
weak link, and maximum differential pressure, to ensure the valve was capable
of functioning under design and licensing bases conditions. Diagnostic and IST
results, including the leak rate test of the air system up to the check valve were
reviewed to verify acceptance criteria were met and performance degradation
would be identified. The inspectors reviewed the capacity calculation for the
safety-related air accumulator to ensure sufficient air was available for the AOV
to function as required upon loss of normal air. In addition, the accumulator
check valve testing was reviewed to ensure the air system capacity would remain
within its design limits. The inspectors reviewed the voltage and power supply
requirements and verified the minimum required voltage would be available to
the valve under all postulated conditions. The inspectors also verified the
operation of the valve was appropriately modelled in battery sizing calculation.
Reactor Core Isolation Cooling Steam Supply Inboard Containment Isolation
Valve (MO-2075): The inspectors reviewed the motor-operated valve (MOV)
calculations, including required thrust, weak link, degraded voltage, and
maximum differential pressure, to ensure the valve was capable of functioning
under design and licensing bases conditions. Diagnostic, IST, and local leak
rate test results were reviewed to verify acceptance criteria were met and
performance degradation would be identified. The inspectors reviewed the
voltage and power supply requirements and verified the minimum required
voltage will be available to the valve under degraded voltage conditions.
Residual Heat Removal Pump 13 (P-202C): The inspectors reviewed the system
flow and NPSH calculations to verify the pump was capable of performing its
safety-related functions. The IST results were reviewed to assess potential
component degradation and impact on design margins. The IST procedures
were examined to determine whether the acceptance criteria adequately
evaluated pump performance. Pump operation in various modes was reviewed
to evaluate the pumps ability to provide the required flow in each mode. The
inspectors reviewed the periodic testing to ensure the pump interlocks would
function as required. The motors fuse/breaker coordination study was examined
to verify adequate coordination. The inspectors reviewed the environmental
qualification (EQ) evaluation and vendor manuals to verify manufacturers
requirements for cooling the motor upper bearing during a postulated event were
addressed. The motor overhaul/replacement schedule and the specification for
overhauling motors was reviewed to ensure the motors safety-related
qualification was maintained. The inspectors compared the motor nameplate
with information in the emergency diesel generator (EDG) loading calculation to
ensure the correct values were incorporated into the calculation.
7
Residual Heat Removal Service Water Pump 13 (P-109C): The inspectors
reviewed system flow and NPSH calculations to determine whether the pump
would operate at the minimum water level in the intake structure. Further,
calculations and the adequacy of the differential pressure setpoint across the
residual heat removal (RHR) heat exchanger were reviewed to ensure the
service water side was at a higher pressure than the RHR side. The inspectors
reviewed the maintenance documents for the most recent pump overhaul and the
re-baselining of the pump performance curves to determine whether the rebuilt
pump met design basis requirements. In addition, the inspectors reviewed
completed pump surveillances for the rebuilt pump to ensure that actual
performance was acceptable. The inspectors reviewed the EQ evaluation and
vendor manuals to verify manufacturers requirements for cooling the motor
upper bearing during a postulated event were addressed. The motors
fuse/breaker coordination study was reviewed to verify adequate coordination.
The inspectors compared the motor nameplate with information in the EDG
loading calculation to ensure the correct values were incorporated into the
calculation. The motor overhaul/replacement schedule and the specification for
overhauling motors was examined to ensure the motors safety-related
qualification was maintained.
Drywell-to-Torus Vacuum Breaker (AO-2382A): The inspectors reviewed the
calculations to demonstrate the valve would function as designed following a
loss-of-coolant accident (LOCA). Specifically, the inspectors reviewed
calculations establishing the valve capacity (e.g., sizing) and the maximum stress
on valve internal components. Additionally, the inspectors reviewed calculations
establishing the acceptance criteria used in TS related surveillance tests
including; the maximum allowable torque required to fully open the valve, and
the differential pressure decay curve for establishing allowable seat leakage.
The inspectors also reviewed completed surveillance and maintenance records
to verify acceptance criteria were met and performance degradation would be
identified. The inspectors reviewed the solenoid valve voltage and power supply
requirements and verified that minimum required voltage would be available
under the worst case loading conditions. The inspectors also reviewed the
micro switch replacement history and the reasons for replacement.
Safety Relief Valve (RV-2-71E): The inspectors reviewed maintenance and test
procedures to determine if the procedures were adequate to ensure that the
safety relief valve (SRV) would reliably function to relieve an over-pressure
condition. Additionally, the inspectors reviewed the calculation demonstrating the
valve had a sufficient supply of nitrogen from the safety-related alternate nitrogen
(AN2) system to allow manual actuation and operation to support post-accident
mitigation functions. The inspectors also reviewed completed surveillance and
maintenance records to verify acceptance criteria were met and performance
degradation would be identified. The inspectors reviewed the actuation of the
low-low set SRV to ensure response times were within allowable values. A
review of the control circuit, calculations for the setpoints, and solenoid response
times was performed to ensure coordination of the low-low set SRV with the
balance of mechanically operated SRVs.
8
Emergency Diesel Fuel Oil System: The inspectors reviewed the modification
that restored the fuel oil system to within the plants licensing basis. Specifically,
the inspectors reviewed the following system components:
Diesel Fuel Oil Transfer Pumps (P-160A-D): The inspectors reviewed the
calculation to confirm these pumps developed sufficient flowrates to support
the system accident mitigation function. Specifically, the inspectors reviewed
the hydraulic calculation that evaluated eight operating configurations to
ensure the minimum required NPSH was maintained for the limiting pump,
and the pump flow capacity was sufficient to maintain the associated EDG
day tank level and/or support transfer of fuel to other storage tanks.
Additionally, the inspectors reviewed the completed pre-operational pump
acceptance tests and performed a visual inspection of the pumps to assess
configuration and potential vulnerabilities to hazards. The inspectors
reviewed the design of the EDG fuel oil system to determine whether all
applicable standards and the requirements for train separation were met.
The inspectors reviewed the control and motor protection scheme for the
newly installed transfer pumps and the associated calculations. Also
reviewed were the cable sizing, voltage drop to motor terminals and motor
control center starter coil pick-up voltages, and additional loading on the EDG
by the additional transfer pump motors. The method for fire separation of
Division II piping and cabling routed through the Division I EDG room was
reviewed to ensure a fire in one room would not affect both EDGs.
Diesel Fuel Oil Transfer Pump Relief Valves (RV-1523, RV-1524, RV-1525,
RV-1526) and Attached Piping: The inspectors reviewed the safety relief
valve design data sheet and vendor catalog information used to establish the
valve lift setpoint and capacity to ensure that the relief valves provided
adequate overpressure protection for the system to meet the pipe design
Code (1977 Edition, Winter 1978 Addenda, ANSI B31.1 Power Piping). The
inspectors reviewed the completed pre-operational acceptance testing for the
relief valves and performed a visual inspection of these valves to assess
configuration and potential vulnerabilities to hazards. Additionally, the
inspectors reviewed the certified material test reports and certification of
conformance records for the relief valves and select pipe components
replaced during the relief valve installation to confirm the valve and pipe
component materials met the design/fabrication Code and pipe specifications.
250vdc Bus (D311): The inspectors reviewed the fault current calculation and
vendor documents regarding breakers contained within bus D311. The
inspectors reviewed the feeder breaker calculation for sizing and protection
scheme. The inspectors reviewed the environmental conditions in the RCIC
room (location of D311) during a high energy line break (HELB). The inspectors
reviewed the D311 cabinet and reviewed cabinet/equipment specifications for
temperature and humidity to ensure equipment would function as required under
worst case environmental conditions. The inspectors also considered the
qualification testing and calculations regarding the HELB boundary door between
the RCIC room and the torus area to verify the door would maintain an adequate
boundary during a HELB event.
9
250vdc Battery (#13): The inspectors reviewed the battery sizing calculation to
verify the battery has adequate capacity to cope with the most limiting accident
and transient conditions, the load profile modeled was conservative compared to
actual worst case loading scenario in the plant. The inspectors also reviewed the
voltage drop calculation to verify the voltages available at all components, under
worst case loading conditions, were above their minimum voltage requirements.
250vdc Battery Charger (D-52): The inspectors reviewed the battery charger
sizing calculation to verify the battery charger has sufficient capacity to supply
the normal loads and fully charge the battery from a fully discharged state within
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The inspectors also reviewed the scheme to supply the charger from
the non-safety-related DG-13 during an extended SBO.
250vdc Battery Room Ventilation Fan (V-EF-40B): The inspectors reviewed
calculations concerning the battery room airflow required for limiting hydrogen
accumulation and the flow necessary to supply outside air across the control
room emergency filtration train (EFT) system inlet radiation monitor to determine
whether the current airflow met design basis requirements. The modification to
the EFT system that blanked off a portion of the EFT inlet duct work was
reviewed to determine whether it would interfere with the fans safety-related
function. The inspectors reviewed periodic system testing and test results to
verify acceptance criteria were met and performance degradation would be
identified. For out of specification flow readings, the inspectors verified causes
were identified and adequate corrective actions were taken. Normal and
abnormal operating procedures were reviewed to ensure they were updated after
the modifications. The inspectors reviewed electrical schematics to ensure
adequate power was available to the fan motor and control room alarms.
4160vac Essential Bus 15 (A5): The inspectors reviewed the sizing and
coordination of the feeder and load breakers. The degraded voltage calculation
was reviewed to verify adequate voltage will be available to safety-related
components during a design basis event concurrent with a degraded voltage
condition. The inspectors also reviewed documents to verify that the feeder
cable to the bus was adequately sized. The 125vdc voltage drop calculation was
reviewed to verify the feeder and load breaker control components will have
sufficient voltage available during the worst case loading conditions. The bus
breaker/relay testing procedures were also reviewed.
b.
Findings
(1) Inadequate Quality Assurance Controls for Nitrogen Supply for the Alternate Nitrogen
System
Introduction: The inspectors identified a finding of very low safety significance (Green),
and an associated Non-Cited Violation (NCV) of Title 10, Code of Federal Regulations
(CFR), Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to
assure the nitrogen supply for the AN2 system was controlled as safety-related in
system specifications, drawings, procedures, and instructions. Specifically, the licensee
had not confirmed effective quality assurance controls were in place to ensure the
bottled nitrogen was acceptable to support the safety-related functions of this system.
10
Description: On July 23, 2015, the inspectors identified the licensee failed to control the
nitrogen supply for the AN2 system as safety-related in system specifications, drawings,
procedures, and instructions. In particular, the inspectors were concerned that the
failure to implement adequate quality controls could result in failure of the AN2 system to
function in support of accident mitigation.
The USAR Section 4.4.2.1, Safety/Relief Valves, stated, the automatic
depressurization system safety/relief valves are designed to withstand a hostile
environment and still perform their function for 100 days following an accident. In
support of this function, a safety-related backup pneumatic supply was provided by the
AN2 system, which automatically supplies pressure to 6 of the 8 SRV actuators upon
loss of the non-safety related instrument nitrogen system. The USAR Section 4.4.4,
stated, The bottled nitrogen supply racks used for the AN2 system are manually
checked for adequate supply and pressure during plant operation at a frequency to
assure minimum design capacity requirements of the system will be met, when required,
assuming worst case leakage rates. To ensure an adequate supply of nitrogen to the
safety-related AN2 system, the licensee determined in Calculation 94-017, Calculation
of Alternate Nitrogen Operability Leakage Criteria, that in addition to the 8 installed
nitrogen bottles, 59 spare nitrogen bottles charged to a minimum of 2283 psig were
required. This quantity of nitrogen represented a 7 day supply, which provided time for
the licensee to procure additional nitrogen from an offsite supplier.
The inspectors observed that the licensee had stored 8 spare bottles of nitrogen in the
turbine building, and in excess of 51 spare bottles within the onsite shipping/receiving
warehouse. These spare nitrogen bottles did not have installed pressure gauges, so the
inspectors could not confirm the pressure (e.g., quantity) of nitrogen stored in the spare
bottles. On August 14, 2014, during installation of spare nitrogen bottles to the AN2
system, the licensee identified two empty nitrogen bottles that prompted an apparent
cause investigation documented in Action Request (AR) 01443013. As a result, the
licensee determined the cause of the empty bottles was the spare nitrogen bottles were
not verified fully charged prior to installation. To correct this issue, the licensee checked
each bottle (with a temporary pressure gage) on a weekly basis to confirm that the spare
bottles stored in the turbine building were fully charged. However, the licensee had
never checked the pressure of the spare bottles in the receiving warehouse, and had not
determined if the empty bottles identified in 2014 were the result of an error in the gas
vendors quality controls or an error in the licensees onsite inventory control process.
The inspectors observed that the nitrogen bottles stored in the receiving warehouse
were not labeled as full or empty and most did not have material stock tags. Because
these bottles were not procured as safety-related, the licensee did not have an inventory
control procedure that required labeling nitrogen bottles as full or empty, or that
prohibited storing empty nitrogen bottles with full bottles of nitrogen, or that required use
of material control stock tags. The inspectors questions on inventory control prompted
the licensee to measure the pressure of the spare nitrogen bottles stored in the receiving
warehouse. As a result of this activity, the licensee identified one bottle with an
unexpectedly low-pressure of 1800 psig. The licensee quarantined this bottle for
subsequent investigation to determine the cause of the unexpected low-pressure.
In addition to the quantity of nitrogen for the AN2 system, the inspectors were concerned
with the quality of the nitrogen because the licensee procured this nitrogen from a
commercial gas supply vendor without performing tests to confirm the type or quality of
the gas received. The inspectors were concerned that if the commercial vendor quality
11
controls were not sufficient, the nitrogen supply may contain high moisture content,
particulates, or be mixed with other gas types. In particular, if moisture levels were
excessive, the water vapor would freeze during expansion of the gas at the AN2 system
pressure reducers and create ice particles that could block AN2 system components
(e.g., pipes or valves), and result in SRVs which could not be manually actuated.
Similarly, a high particulate concentration could block small passages in AN2 system
components (e.g., pressure regulators) and restrict the flow of nitrogen resulting in
SRVs, which could not be manually actuated. If the SRVs could not be operated
manually, it would impair/prevent accident mitigation functions such as reactor pressure
control, reactor depressurization, and alternate shutdown cooling. The inspectors
concerns, prompted the licensee to contact the gas supply vendor to determine what
vendor controls were used to confirm the quantity and quality of the nitrogen delivered.
The commercial vendors controls included evacuation of reused bottles and sampling of
the gas in one bottle from each batch (groups of 24) to confirm gas purity and lack of
contaminants (e.g., moisture content). Additionally, the gas supply vendor reportedly
used a closed process to fill the nitrogen bottle that did not introduce particles. The
licensee concluded that the gas vendor quality controls provided a sufficient basis to
conclude that the AN2 system was operable.
Title 10 CFR 50.2 states, that, safety-related structures, systems and components
(SSCs) means those SSC that are relied upon to remain functional during and following
design basis events to assure: (1) The integrity of the reactor coolant pressure
boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown
condition; or (3) The capability to prevent or mitigate the consequences of accidents
which could result in potential offsite exposures comparable to the applicable guideline
exposures set forth in 10 CFR 50.34(a)(1), or 10 CFR 100.11 of this chapter, as
applicable. The licensee guidance to implement this definition existed in Attachment 2,
Classification Guidance, of procedure FP-E-RTC-02, Equipment Classification, which
stated, in part, Items that are either installed in safety-related systems and relied upon
to provide or support the safety-related functions, or are installed in any system needed
to satisfy safety-related interface requirements (e.g., isolation devices) are identified.
These items are classified as safety-related. Based upon this guidance, the nitrogen
supplied by four bottles installed in each AN2 system train should have been identified
as safety-related because the nitrogen was required to support the safety-related
functions of the AN2 system. On drawing NH-36049-10, Alternate Nitrogen Supply
System, the installed nitrogen bottles were located outside the safety-related portion
of the AN2 system piping boundary and instead were identified as a special concerns
item, which was defined as an item subject to augmented quality controls in
FP-E-RTC-02. The licensee added the special concerns item designation for the
nitrogen bottles in 1988, as a result of an NRC commitment associated with
NUREG 0737, Clarification of Three Mile Island Action Plan Requirements.
However, the licensee had not procured the installed or spare nitrogen bottles under
a safety-related Quality Control Program as described in 10 CFR Part 50, Appendix B.
Instead, the licensee had procured the nitrogen bottles from a commercial vendor
without auditing the gas vendors quality controls and without conducting confirmatory
tests to verify the type, quality or quantity of gas delivered.
The licensee initiated AR 01486991, and contacted the commercial nitrogen gas supplier
to confirm that the vendors quality controls provided a sufficient basis to conclude the
AN2 system was operable. In addition, the licensee identified an action to evaluate the
controls in place to ensure that AN2 system nitrogen supply bottles had adequate
pressure and adequate gas quality.
12
Analysis: The inspectors determined the failure to demonstrate the nitrogen supply for
the AN2 system was controlled as safety-related in system specifications, drawings,
procedures and instructions was contrary to 10 CFR Part 50, Appendix B, Criterion III,
Design Control, and a performance deficiency. The finding was determined to be more
than minor in accordance with Inspection Manual Chapter (IMC) 0612, Appendix B,
Issue Screening, dated September 7, 2012, because the inspectors answered Yes
to the More-than-Minor screening question, If left uncorrected, would the performance
deficiency have the potential to lead to a more significant safety concern? Specifically,
if the commercial (e.g., non-safety) gas supply vendor quality controls were not
adequate to ensure contaminants such as moisture or particulates were excluded from
the nitrogen gas bottles, it could potentially disable the AN2 system capability to support
manual operation of SRVs during post LOCA mitigation.
The inspectors determined the finding could be evaluated using the Significance
Determination Process (SDP) in accordance with IMC 0609, Significance Determination
Process, dated April 29, 2015, Attachment 0609.04, Phase 1 Initial Screening
and Characterization of Findings, dated June 19, 2012, for the Mitigating Systems
cornerstone. The inspectors evaluated the finding using Appendix A, The Significance
Determination Process for Findings At-Power. The finding screened as very low safety
significance (Green) because the inspectors were able to answer Yes to screening
Question A1 in Exhibit 2 because the finding represented a design deficiency confirmed
not to result in loss of operability or functionality.
The inspectors did not identify a cross-cutting aspect associated with this finding as it did
not reflect current performance.
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, required,
in part, Measures shall be established to assure that applicable regulatory requirements
and the design basis, as defined in 10 CFR 50.2, and as specified in the license
application, for those SSC to which this appendix applies are correctly translated into
specifications, drawings, procedures, and instructions. These measures shall include
provisions to assure that appropriate quality standards are specified and included in
design documents and that deviations from such standards are controlled. Measures
shall also be established for the selection and review for suitability of application of
materials, parts, equipment, and processes that are essential to the safety-related
functions of the SSC.
Contrary to the above, as of July 23, 2015, the licensee had not established measures to
assure that the design basis for the nitrogen supply to the AN2 system was correctly
translated (e.g., classified/controlled as safety-related) into specifications, drawings,
procedures, and instructions.
Because this violation was of very-low safety significance, and it was entered into the
licensees Corrective Action Program (CAP) as AR 01486991, where the licensee
contacted the supplier to confirm the vendors quality controls provided a sufficient basis
to conclude the AN2 system was operable, this violation is being treated as an NCV,
consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000263/2015007-
01, Inadequate Quality Assurance Controls for Nitrogen Supply for the AN2 System).
13
(2) Failure to Review for Suitability of Application of Safety-Related Relays Installed Beyond
Their Service Life
Introduction: The inspectors identified a finding of very low safety significance (Green),
and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control,
for the failure to assure measures were established for the selection and review for
suitability of application of materials, parts, equipment and processes that were essential
to the safety-related functions of SSC. Specifically, the licensee failed to review for
suitability of application of safety-related Agastat and General Electric (GE) relays that
exceeded their service life, a condition nonconforming to their design basis, to justify
their continued service considering in-service deterioration.
Description: During the 2012 Problem Identification and Resolution inspection,
Unresolved Item (URI)05000263/2012008-01 was opened related to the qualification
basis for safety-related relays and motor starter contactors. The URI identified concerns
with the licensee not replacing safety-related relays and motor starter contactors that
were beyond the vendors recommended service life without an appropriate evaluation
justifying the extension of their service life. The inspectors in consultation with Nuclear
Reactor Regulation staff issued Task Interface Agreement (TIA) 2014-01, Final Task
Interface Agreement - Regulatory Position on Design Life of Safety-Related Structures,
Systems, and Components Related to Unresolved Items at Donald C. Cook Nuclear
Power Plant, Monticello Nuclear Generating Plant and Palisades Nuclear Plant. The
TIA was issued on May 7, 2015, and concluded when a licensee becomes aware that a
safety-related SSCs service life has been exceeded or information challenges the
presumption that a safety-related SSC can perform its specified function, the licensee
must promptly address and document this non-conforming condition in accordance with
the licensees NRC approved Quality Assurance Program, the licensees
operability/functionality program and the CAP. This includes completing appropriate
corrective actions in a timely manner and documenting licensees evaluations justifying
the service life extensions.
During this inspection, the inspectors noted the licensee previously initiated
AR 01446684, which identified a number of corrective actions. Some actions were
already completed and the remaining were scheduled for completion in a timely
manner. Immediate corrective actions included instituting a Relay Monitoring Program,
performing generic service life evaluations on some of the safety-related Agastat and GE
relays, and identifying and replacing relays that had exceeded vendor recommended
service life. The licensee continued to identify safety-related relays exceeding vendor
recommended service life and had plans to conduct extent of condition reviews.
A separate action item was initiated to evaluate motor starter contactors.
Analysis: The inspectors determined the failure to review for suitability of application of
safety-related relays installed beyond their service life to justify their continued service,
considering in-service deterioration, was contrary to 10 CFR Part 50, Appendix B,
Criterion III, and a performance deficiency. The finding was determined to be more than
minor in accordance with IMC 0612, Appendix B Issue Screening, because the
inspectors answered Yes to the More-than-Minor screening question, If left
uncorrected, would the performance deficiency have the potential to lead to a more
significant safety concern? Specifically, these safety-related relays were installed in
protective circuits such as reactor protection system, etc., and their failure could impact
the proper operation of these protective schemes.
14
The inspectors determined the finding could be evaluated using the SDP in accordance
with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1
Initial Screening and Characterization of Findings, for the Mitigating Systems
cornerstone. The inspectors evaluated the finding using Appendix A, The Significance
Determination Process for Findings at Power. The finding screened as very low safety
significance (Green) because the inspectors were able to answer Yes to screening
Question A1 in Exhibit 2, because the finding represented a qualification deficiency of a
mitigating SSC confirmed not to result in loss of operability or functionality.
The inspectors did not identify a cross-cutting aspect associated with this finding as it did
not reflect licensees current performance.
Enforcement: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, required,
in part, Measures shall be established to assure that the selection and review for
suitability of application of materials, parts, equipment, and processes that are essential
to the safety-related functions of SSC.
Contrary to the above, as of July 24, 2015, the licensee failed to establish measures
to ensure the selection and review for suitability of application of materials, parts,
equipment, and processes that were essential to the safety-related functions of SSC.
Specifically, the licensee failed to review for suitability of application of safety-related
Agastat and GE relays that exceeded their service life, a condition nonconforming to
their design basis, to justify their continued service considering in-service deterioration.
Because this violation was of very-low safety significance, and it was entered into the
CAP as AR 01446684, where corrective actions to replace or evaluate relays were
either already completed or scheduled for completion in a timely manner, this violation is
being treated as an NCV, consistent with Section 2.3.2, of the NRC Enforcement Policy.
(NCV 05000263/2015007-02, Failure to Review for Suitability of Application
Safety-Related Relays Installed Beyond Their Service Life.)
.4
Operating Experience
a.
Inspection Scope
The inspectors reviewed five operating experience issues (samples) to ensure that NRC
generic concerns had been adequately evaluated and addressed by the licensee. The
operating experience issues listed below were reviewed as part of this inspection:
IN 2012-14, Motor-Operated Valve Inoperable Due to Stem-Disc Separation;
IN 2013-05, Battery Expected Life and Its Potential Impact on Surveillance
Requirements;
RIS 2000-012, Resolution of Generic Safety Issue B-55, Improved Reliability of
Target Rock Safety Relief Valves;
GE Service Information Letter (SIL) 44, GE HFA Relay Coil Life; and
GE SIL 196 - Original thru Supplement 17, Recommendations for Target Rock
Main Steam Safety/Relief Valves.
15
b.
Findings
No findings were identified.
.5
Modifications
a.
Inspection Scope
The inspectors reviewed four permanent plant modifications related to selected
risk-significant components to verify that the design bases, licensing bases, and
performance capability of the components had not been degraded through modifications.
The modifications listed below were reviewed as part of this inspection effort:
DC79M070, Modify Drywell to Torus Vacuum Breakers;
EC23805, EDG Fuel Oil Train Separation; and
EC25733, Alternate Nitrogen Bottle Change-out Check Valves.
b.
Findings
No findings were identified.
.6
Operating Procedure Accident Scenarios
a.
Inspection Scope
The inspectors performed a margin assessment and a detailed review of two
risk-significant, time critical operator actions and an alternate method to provide power to
battery chargers during a prolonged SBO. These actions were selected from the
licensees PRA rankings of human action importance based on risk achievement worth
values. Where possible, margins were determined by the review of the assumed design
basis and USAR response times and performance times documented by job
performance measures results. For the selected operator actions, the inspectors
performed a detailed review and walk through of associated procedures, including
observing the performance of some actions in the plant, with an appropriate plant
operator to assess operator knowledge level, adequacy of procedures, and availability of
special equipment where required.
The following operator actions were reviewed:
Actions to isolate flooding from plant administration building fire header;
Actions to isolate Service Water line to 12 Main Feedwater cooler line break; and
Actions to use the non-safety-related DG13 to provide power to the Division II
250 vdc Battery Chargers in the event of an SBO.
b.
Findings
No findings were identified.
16
4.
OTHER ACTIVITIES
4OA2 Identification and Resolution of Problems
.1
Review of Items Entered Into the Corrective Action Program
a.
Inspection Scope
The inspectors reviewed a sample of the selected component problems identified by
the licensee and entered into the CAP. The inspectors reviewed these issues to verify
an appropriate threshold for identifying issues and to evaluate the effectiveness of
corrective actions related to design issues. In addition, corrective action documents
written on issues identified during the inspection were reviewed to verify adequate
problem identification and incorporation of the problem into the CAP. The specific
corrective action documents sampled and reviewed by the inspectors are listed in the
attachment to this report.
The inspectors also selected two issues identified during previous CDBIs to verify that
the concern was adequately evaluated and corrective actions were identified and
implemented to resolve the concern, as necessary. The following issues were reviewed:
NCV 05000263/2012007-03; Failure to Maintain the Degraded Voltage Function
Time Delay Design: The inspectors reviewed the licensees design change that
removed the 1AR transformers additional 5 second time delay and restored
compliance to the TSs.
NCV 05000263/2012007-04; Failure to Analyze Effect of Degraded Voltage on
Proper Operation of Thermal Overload Relays: The inspectors reviewed three of
four corrective actions completed associated with this issue. The completed
issues included: 1) EC19903 increased the margins for the subject thermal
overload relay (TOL) settings; 2) EC25687 analyzed TOL performance for MOVs
during a degraded voltage with LOCA scenario; and 3) EC25688 analyzed TOL
performance for all continuous duty motors during a degraded voltage with LOCA
scenario. The fourth issue to formalize the analysis was included in the
Monticello Calculation Reconstitution Project with completion planned by
July 2016. This was being tracked by AR 01197202 and OBN01479704-04.
b.
Findings
No findings were identified.
4OA5 Other Activities
.1
(Closed) URI 05000263/2012008-01; Qualification Basis for Safety-Related Relays and
Motor Starter Contactors: This URI is closed to NCV 05000263/2015007-01, Failure to
Review for Suitability of Application of Safety-Related Relays Installed Beyond Their
Service Life. See Section 1R21.3.b.(2).
.2
(Closed) URI 05000263/2012008-02; Concern with Periodic Design Basis Testing of
Installed Relays and Motor Starter Contactors: During the 2012 Problem Identification &
Resolution inspection, the inspectors were concerned the licensee was not testing
installed relays and motor starter contactors to verify their design basis capacity in
17
accordance with IEEE Standard 336-1971 and Regulatory Guides 1.30 and 1.33. The
inspectors noted that the Regulatory Guides did not contain detailed or specific testing
instructions and only had general guidelines. The IEEE-336 did have detailed
instructions for installation, inspection, and testing for class 1E power, instrumentation
and control equipment at nuclear facilities. While reviewing the applicability section of
the IEEE-336, inspectors noted the standard did not apply to periodic testing and
maintenance following initial installation. The standard only applied to initial installation
of new equipment or equipment modifications, or modification of power, instrumentation
and control equipment and systems in a nuclear facility from the time the equipment was
turned over for installation until it was declared operable for service. Therefore, the
inspectors concluded the existing periodic testing and maintenance activities performed
by the licensee on installed relays and motor starter contactors were adequate. No
violations of NRC requirements were identified by the inspectors. Therefore, this URI is
closed.
4OA6 Management Meeting
.1
Exit Meeting Summary
On July 24, 2015, the inspectors presented the inspection results to Mr. P. Gardner, and
other members of the licensee staff. The licensee acknowledged the issues presented.
The inspectors asked the licensee whether any materials examined during the
inspection should be considered proprietary. Several documents reviewed by the
inspectors were considered proprietary information and were either returned to the
licensee or handled in accordance with NRC policy on proprietary information.
ATTACHMENT: SUPPLEMENTAL INFORMATION
Attachment
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
P. Gardner, Site Vice President
S. Northavol, Vice President Nuclear Fleet Operations
T. Talyor, Vice President Nuclear Oversight
H. Hanson, Jr., Plant Manager
A. Gonnering, Configuration Management Supervisor
M. Kelly, Performance Assurance Manager
M. Lingenfelter, Director of Engineering
K. Scott, Director Site Operations
A. Ward, Regulatory Affairs Manager
R. Zyduck, Design Manager
B. Halvorson, Engineering Supervisor
A. Kouba, Regulatory Affairs Manager
C. Fosaaen, Regulatory Affairs
N. Friebel, Design Engineer
D. Alstad, Design Engineer
E. Watzel, Electrical Design Engineering Supervisor
P. Young, Program Engineering Supervisor
U.S. Nuclear Regulatory Commission
K. OBrien, Director, Division of Reactor Safety
P. Zurawski, Senior Resident Inspector
P. Voss, Resident Inspector
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened 05000263/2015007-01
Inadequate Quality Assurance Controls for Nitrogen Supply
for the AN2 System (Section 1R21.3.b.(1))05000263/2015007-02
Failure to Review for Suitability of Application of
Safety-Related Relays Installed Beyond Their Service Life
(Section 1R21.3.b.(2))
Closed 05000263/2015007-01
Inadequate Quality Assurance Controls for Nitrogen Supply
for the AN2 System (Section 1R21.3.b.(1))05000263/2015007-02
Failure to Review for Suitability of Application of
Safety-Related Relays Installed Beyond Their Service Life
(Section 1R21.3.b.(2))05000263/2012008-01
Qualification Basis for Safety-Related Relays and Motor
Starter Contactors (Section 4OA5)05000263/2012008-02
Concern with Periodic Design Basis Testing of Installed
Relays and Motor Starter Contactors (Section 4OA5)
2
LIST OF DOCUMENTS REVIEWED
The following is a list of documents reviewed during the inspection. Inclusion on this list does
not imply that the NRC inspectors reviewed the documents in their entirety, but rather, that
selected sections of portions of the documents were evaluated as part of the overall inspection
effort. Inclusion of a document on this list does not imply NRC acceptance of the document or
any part of it, unless this is stated in the body of the inspection report.
CALCULATIONS
Number
Description or Title
Revision 01-036
lnservice Testing Pump and Valve Acceptance Criteria Rounding
Evaluation
48 01-043
Verification of Torus to Drywell Vacuum Breaker Sizing Parameters
0 02-179
MNGP 125 Volt Division. I Battery Calculation
3 04-048
MNGP 250 Volt Division I Battery Calculation
2 05-128
- 13 and #16 Battery Charger Sizing
0 06-104
480V MCC to Terminal Voltage Drop
3E 08-077
0A 09-192
Reactor Building Composite Profiles for Environmental Qualification
0 10-168
RHR and Core Spray Motor Feeder Cable Sizing
0 10-118
0 11-295
MO-2075 Component Calculation
0 11-326
0 14-001
Monticello Stem Lubrication Study
0 14-057
Evaluation Buried Diesel Oil Overflow Line for Day Tank T-45B
1 14-073
EDG Diesel Oil Hydraulic Model
0
15-014/12
EDG Fuel Oil Piping/Cable Fire Barrier
0 92-220
Instrument Setpoint Calculation, 4.16 Kv Degraded Voltage
2 92-224
Emergency Diesel Generator Loading
006A 93-066
AC Loads Study, Degraded Voltage Setpoint, 1R Transformer,
LOCA Load
6 94-094
MCC Starter Coil Pick-Up Voltages & Maximum Cable Lengths
1, 1B 95-049
Monticello Apparent Disc Coefficient of Friction Determination
3
CA 08-157
Combined AC Model Database
000-B
CA-00-003
Response Time Increase of SRV Solenoids
0
CA-00-057
Drywell to Suppression Chamber Differential Pressure Decay Curve
for a 1 Inch Diameter Orifice
0
CA-00-104
Intake Structure Minimum Water Level
0A
CA-01-037
Determination of the Maximum Allowable Torque to Open Torus to
Drywell Vacuum Breakers
0A
CA-01-053
Evaluation of the Pressure Capacity of a Door
7
CA-01-137
Evaluation of Drywell/Wetwell Vacuum Breakers
0
CA-01-155
Maximum Allowable Leak Rate for the RCIC Minimum Flow Valve
Air Accumulator System
1
CA-01-174
Minimum Required RHRSW Pressure at RHR Heat Exchanger
3
CA-01-188
RHR Motor Start Time Evaluation
0
CA-02-002
RCIC Min Flow Line Flow Rate Analysis
0
CA-02-145
HPIC and RCIC NPSH Calculations for Use in EOPs
0
CA-02-197
EQ of Dow Corning Silicone RTV Foam
1
3
CALCULATIONS
Number
Description or Title
Revision
CA-03-039
SRV Low-Low Setpoint
1
CA-03-041
Setpoint Calc SRV Low-Low Set Inhibit Timer
0
CA-03-06
AOV Component Calculation, CV-2104
3
CA-03-097
HPCI/RCIC Suction Head Height Difference
0
CA-03-199
Sensitivity of EOP Calculations to ECCS Pump Curve Data
0
CA-05-019
NPSH Requirements for Operating ECCS Pumps from the CST
0C
CA-05-124
Hydrogen Generation of #13 & #16 Battery Rooms
15
CA-13-055
Core Spray and LPCI Flow Delivered to Reactor Vessel for
Safety Analyses
0
CA-80-020
NPSH Requirements for RHR Pumps
0
CA-91-009
250VDC Fault Current
1
CA-92-224
Emergency Diesel Generator Loading
6
CA-94-017
Calculation of Alternate Nitrogen Operability Leakage Criteria
10
CA-95-099
Determine the Minimum RHR Pump Flow Required During
Testing
0
CA-95-116
Stem Thrust Assessment of 3 A/D Gate Valves: MO-2075 &
MO-2076
1
CA-96-079
High Energy Line RCIC HELB in the RCIC Room
1
CA-96-169
3B
CA-97-194
LLRT Test Volumes for the RCIC Air Accumulator Check Valve
AI-612
0
CA-99-011
Outlier (Seismic) Evaluation of Service Water Pumps
1
RX Bldg Envir for EPU HELB, SBA & Post LOCA
1
Motor Control Center Thermal Lag Analysis
0
Review of TOL Performance for Auto Initiated MOVs During a
Degraded Voltage Condition
0
Review of Protective Device Performance for Safety Related
Continuous Duty 480V Loads During a Degraded Voltage
Condition
0
FBS-0503-1
Fuse and Breaker Coordination Study
2
PRA-CALC-
II.SMR.02.001 Makeup Requirements After Scram from 1775 MWth
0
PRA-CALC-
II.SPA.02.001
RCIC Min Flow Valve
0
Evaluation of Agastat Service Life
0
Service Life Evaluation for Select GE HFA Century Series
Relays
0
Engineering Evaluation Supporting 2000 Cycle Test Basis for
0
Service Life Evaluation for GE HFA Century Series Relays
0
Service Life Evaluation for Select GE HGA Century Series
Relays
0
Service Life Evaluation for Reasonable Assurance of Agastat
Function
0
4
CORRECTIVE ACTION DOCUMENTS GENERATED DUE TO THE INSPECTION
Number
Description or Title
Date
01483209
2nd Transmittal did not have All Requested Info
06/17/15
01483311
Calcs for EDG Fuel Oil Mod Needed to be Re-sent
06/22/15
01483808
Material Storage in RCIC Cable Closet
06/23/15
01483828
Bent Rod Hanger on Conduit in RCIC Room
06/23/15
01483833
Typo Corrected on the Receipt Inspection Report from 2012
06/23/15
01484025
No Formal Testing of ECCS Corner Room Sump Capacity
06/24/15
01484043
Inspector Question Response Delay
06/25/15
01484051
Unexpected Absence of Inspector Shadow
06/25/15
01484170
CMTRs not Located- E-SRV & Valve Body & Disc of AO-2382A
06/25/15
01484177
13 Diesel Fuel Cooler Fan Power Cable Potential Damage
06/25/15
01484180
Internal Flood TCOA Scenario Insufficient
06/25/15
01484193
Incorrect Calc Given to NRC
06/25/15
01484193
Calc 02-197 not Taken to Inactive Correctly
06/25/15
01484210
CMTR for Weld Material Q12 not Provided to NRC
06/25/15
01484265
Walkdown Forms not Incorporated into Plant Records
06/26/15
01484265
DOL Walkdown Forms not Incorporated into Plant Records
06/26/15
01484364
Internal Flooding DBD not Consistent with Licensing Basis
06/26/15
01484365 03-006 Apparent Typo Error in Section 6.0
06/26/15
01484534
MOV TOL Calculations at Incorrect Status
06/29/15
01484697
Extraneous Information on NX-8685-4
06/30/15
01484859
Listed Horsepower Wrong on RHRSW Control Drawing
07/01/15
01484919
Procedure 0214 Temperature Controls Needs Enhancement
07/02/15
01485196
RCIC 4120-PM Documentation Enhancement
07/06/15
01485387
8153 Procedure Improvement Opportunities
07/07/15
01485410
Unable to Locate GE SIL 196 Evaluation Supplement 5, 11, 17
07/07/15
01485425
ECCS Corner Room Sump Pump Capacity
07/07/15
01485467
NRC Insp Question Response Delay
07/08/15
01485508
FLEX Charger Mod did not Update Procedure 8153
07/08/15
01485509
ECCS Corner Room Sump Pump Information Differences
07/08/15
01485551
Motor Program Documents not Maintained
07/08/15
01485554
Evaluate Recommendation of GE SIL 196 Supplement 1
07/08/15
01485569
RV-1524 Outlet Pipe Size not in Compliance with B31.1 Code
07/09/15
01485668
Motor Refurb Spec does not Exist for Safety-Related
07/09/15
01485693
Diesel Oil System RV Reaction Loads not In Pipe Analysis
07/09/15
01485697
AR01322841-10 Completed at Incorrect Status
07/09/15
01485786
Delay in Issuing Contract to Support Question Response
07/10/15
01485799
SRV PM Requirements not Transferred to Vendor PM
07/10/15
01486266
Drawing does not Reflect Specific Valves with Shims
07/15/15
01486343
Torus/DW Vacuum Breaker Test Report
07/16/15
01486500
VTMs not Updated to Reflect SLA Results
07/17/15
01486699
Safety Related Relays not Part of the Vendor Contact Program
07/20/15
01486828
16A-K37 Beyond Service Life
07/21/15
01486951
Some Alt N2 Bottles Stored in REC WH not Tagged
07/22/15
01486991
Question on AN2 Bottle Pressure and Gas Quality
07/22/15
01487027
Lack of Procedure Controls for AN2 Bottle Storage
07/22/15
01487139
NRC Question on Qualification of Motor Bearing
07/23/15
01487272
Component Storage in Recv Whse Questioned
07/24/15
5
CORRECTIVE ACTION DOCUMENTS REVIEWED DURING THE INSPECTION
Number
Description or Title
Date
00621181
RIS-2000-12-Resolution of Generic Safety Issue B-5
08/28/00
00630310
Lack of Ventilation to Diesel Pump Room
08/14/02
01044201
Fuel Oil House has Potential for Hazardous Environment
08/14/06
01169547
RSW Pump Flow Ref Value Discrepancy Exists Between
Calculation & Procedure
02/16/09
01182779
- 13 DG Small Coolant Leak Getting Worse
05/21/09
01192708
Question Basis for Min RHRSW DP per IST Program
08/07/09
01209786
13 RHRSW Pump September Trends
12/08/09
01229823
P-109C IST Reference Value Change
04/28/10
01233820
Basis Change: Add LR Text and Attribute for OCCW PMRQ's
05/21/10
01238600
V-EF-40B Discharge Flow Out of Specification
06/2510
01242365
V-EF-40B Discharge Flow Out of Specification
07/23/10
01249264
Failed PMT on V-EF-40B
10/07/10
01278466
MO-2075 Excessive Thrust on As-Found VIPER Test
04/01/11
01289417
Possible Deficiency in EQ Testing of Limitorque Part 21
06/06/11
01289887
RV-2-71E Elevated Tailpipe Temps
06/09/11
01291640
08/08/11
01291959
Foreign Material Found in Spare SRV (at NWS)
06/24/11
01293850
Allowable Leakage for HPCI/RCIC Minimum Flow Valves
07/08/11
01312421
Untimely Resolution of CAP 1196513
11/09/11
01332373
2012 CDBI Motor TOLs may Trip w/ Degraded Volt
04/05/12
01332567
2012 CDBI TOL Coordination w DVR
04/06/12
01334146
ACE 2012 CDBI TS Degraded Voltage Time Value
04/17/12
01334248
Potential Margin Reduction from Degraded ECCS Pump Head
11/26/13
01338565
MOV Limiting Stroke Time Margin Issue
05/21/12
01338566
RHRSW Reduced Flow/Head Margin Issue
05/21/12
01338567
RHR Reduced Flow/Head & NSPH/Vortex Concerns
07/15/14
01345964
NRC IN 2012-14 Motor-Operated Valve Inoperable due to Stem-
Disc Separation
07/25/12
01350679
P-109C, Reference Value Change
09/06/12
01356651
Discrepancy with Sulzer Info and 4214-PM
10/26/12
01375387
SRV LLS TS Allow Values Conflict w/ SRV Mech Allow Setpoint
03/20/13
01375742
OE: NRC IN 2013-05
03/22/13
01378744
06/28/13
01379613
SRV Actuator Testing may be Non-Conservative
04/19/13
01389246
NRC has Question on Alignment Data for HPCI and RCIC
08/05/13
01390472
OE 248697 Both Divs of RHR Inop From Leak
07/18/13
01411214
12/30/13
01417977
Failure of Drywell Vacuum Breaker to Close
02/07/14
01418471
AO-2382A Torus-to-DW Vac Brk Closed Indication Anomaly
02/11/14
01420318
Torus Vacuum Breaker Inadequate PMT
02/25/14
01420700
Small Coolant Leak on #13 DG Radiator
02/28/14
01423951
13 RHRSW Pump Exceeds MR Reliability Criteria
03/24/14
01424260
Future Preconditioning of Vacuum Brkrs Found Unacceptable
03/26/14
01426064
TS 3.6.1.7 has no Actions for Closed Vlv Brkr that Failed STP
04/09/14
01431529
Internal Flooding PAB F.P. Break Control of TCOA Inadequate
06/06/15
6
CORRECTIVE ACTION DOCUMENTS REVIEWED DURING THE INSPECTION
Number
Description or Title
Date
01434290
Coolant Leak Observed on #13 DG
06/12/14
01438672
Oil Leak Detected on 13 Diesel Generator
07/17/14
01439686
Undefined Term Used in the USAR with Regard to RCIC
11/05/14
01443013
Replacement Alternate N2 System Found Empty
08/14/14
01443073
13 Diesel Generator has a Minor Leak from Header
08/15/14
01443510
OE: 312166 Question Concerning RCIC Cooling Test
08/19/14
01448769
C&D Tech Identifies Issue with Battery Separator Plates
10/01/14
01453481
OE: NRC Part 21 C&D Technologies Batteries
10/29/14
01459466
13 Diesel (G-90) Engine Coolant Leak
12/15/14
01459539
OE: NRC PEN 50675 LaSalle RCIC Unanalyzed Condition
12/15/14
01471379
T-44 Level with 2 Pump Operation after Mod Implementation
03/24/15
01474704
Design Issue Discovered in DOL Separation Modification
04/15/15
01475109
Design Input not Considered for DOL Hydraulic Model
04/17/15
01475179
MO-2075 Exceeded App J Admin Limit
04/18/15
01475653
Leakage Found on RV-2-71E Actuator
04/22/15
01476203
Air Leaks on SRV Bellows Leak Alarm Pressure Switches
04/24/15
01476257
P-160A/B Pipe Unions Found Hand Tight
04/25/15
01477101
MOV Transient Analysis did not Consider TOL Size
04/30/15
01477714
No Formal Calc to Support MOV TOLs
05/05/15
01477916
Invalid AO-2382A Full Open Torque
05/06/15
01477935
CAs for 2012 Violation Inadequate
05/06/15
01478212
Interference on AO-2382A Vacuum Breaker Actuator
05/08/15
01479704
Circuit Protective Device Operation-Sustained Degraded Volt
05/18/15
DRAWINGS
Number
Description or Title
Revision
ES1506100
Fuel Transfer Pump Assembly
A
M-288
Reactor BLDG.-Plan at EL. 8963
C
NE-36347-10
- 142-480V MCC B42
81
NE-36394-10B
RHRSW Pump P109-C ACB No. 152-507
76
NE-36399-9
Essential Bus Transfer Circuits-Division I
77
NE-36404-4B
RHR Pump P-202C ACB 152-503 Control
76
NE-36438-9
11EDG Diesel Oil Pumps A and C, P-160A and C Control
82
NE-36640-5
250VDC MCC Schedule D311
76
NE-93503-3
HVAC Controls & Interlocks Scheme V201
F
NE-93504-20
EFT System HVAC Annunciator
C
NE-93545
Loop Diagram Exhaust Fans V-EF-40B
4
NF-119034-2-C
- 12 Diesel Generator Fuel Oil System
0
NF-36298-1
Electrical Load Flow One Line Diagram
111
NF-36298-2
DC Electrical Load Distribution One Line Diagram
90
NF-36672
Standby Diesel Generators Arrangement & Piping
78
NF-95915-3
Blowdown Control System Division I Elem Diagram
76
NH-170037
Main Control Room CRV/EFT System
81
NH-36049-10
Alternate Nitrogen Supply System
78
NH-36051
Diesel Oil System
84
NH-36241-1
Reactor Pressure Relief
78
NH-36246
Residual Heat Removal System
84
7
DRAWINGS
Number
Description or Title
Revision
NH-36247
Residual Heat Removal System
85
NH-36251
RCIC (Steam Side)
80
NH-36252
RCIC (Water Side)
79
NH-36664
RHR Service Water & Emergency Service Water Systems
87
NH-36665
Service Water System & Make-up Intake Structure
97
NH-91177
Disc and Post for Vacuum Breaker Valve
C
NQ-74976
Three Hour Fire Barrier 11 EDG Room, EDG Trench
0
NX-15111-1
MNGP Main Steam Safety/Relief Valve, Target Rock Model
7467F, 6X10, Outline
E
NX-17496-3
MNGP Protective Relay Cards-4kV
11
NX-7822-22-5A
RCIC Steam Supply Isolation MO-2075 Scheme
A
NX-7831-439
Main Steam Safety/Relief Valves, Target Rock Model 7367F,
6X10, Parts List
77
NX-7831-539
SRV Air Actuator Model 7467F
77
NX-7905-77
76
NX-8685-4
Funbore Vacuum Breaker Valves
E
NX-9068-37
Outline Induction Motor
F
NX-9235-32
3 600# Globe Valve Motor Operated
A
NX-9235-43
3 900# Gate Valve MO-2075 & MO-2076 Carbon Steel Bolted
L
NX-9285-5
Fuel Transfer Pump Assembly
0
NX-9525-1
RHRSW Pump Assembly
76
NX-9525-8
RHRSW Pump Open Flange Column Details
76
MISCELLANEOUS
Number
Description or Title
Date or
Revision
Plant Health Report - RHRSW System
06/10/15
Plant Health Report - Emergency Filtration Train
06/18/15
Plant Health Report - Reactor Core Isolation Cooling System
06/10/15
System Walkdown Observation- EDG Fuel Oil
11/26/13
System Walkdown Observation- EDG Fuel Oil
08/28/14
System Walkdown Observation- EDG Fuel Oil
12/18/14
System Walkdown Observation- EDG Fuel Oil
03/30/15
Safety Relief Valve Data Sheets-RV 1523, 1524, 1525, 1526
12/11/14
System Health Report - Auto Pressure Relief
06/10/15
System Health Report - Diesel Oil System
06/10/15
System Health Report - Primary Containment
06/10/15
10040.D5.7
Design Criteria Document - Heating, Ventilation, and Air
Conditioning System for the Main Control Room, Emergency
Filter Train and Technical Support Center at Monticello Nuclear
Generating Plant, Northern States Power
5
2015-01-030
Component Design Basis Inspection (CDBI) Readiness
0
22A1121
Design Specification - Drywell to Suppression Chamber
Vacuum Breakers
0
79M070
Design Change to Torus to Drywell Vacuum Breaker
0 98-018
EQ, General Electric Motors (50.49)
1
8
MISCELLANEOUS
Number
Description or Title
Date or
Revision 98-026
Limitorque Motor Operators (50.49)
0
A.3-15-E
Fire Zone 15-E Strategy
7
Contract
940015040
COC- 4(1) Gate CS 800#
01/16/15
DBD T.08
Design Basis for Internal Flooding
3
DBD-B.02.03
Reactor Core Isolation Cooling System
77
DBD-B.03.04
Residual Heat Removal System
7
DBD-B.08.01.03
Residual Heat Removal Service Water System
6
DBD-B.08.13
Control Room Heating, Ventilation and Emergency Filtration
System
3
DRF T23-00789-
00
GE Letter - Monticello Nuclear Power Station - Response to
NMC Question Regarding Impact of Power Rerate on
Drywell-to-Suppression Chamber Steam Bypass Leakage
03/25/01
EM7114T
Baldor 1//.75,1760//1460RPM,3PH,60//50Hz
08/08/14
EQ 98-022
0
FBS-0507-1
Fuse/Breaker Coordination Study, P-109C
1
FBS-4030-02-1
Fuse/Breaker Coordination Study, P-160B
0
FBS-4080-51-1
Fuse/Breaker Coordination Study, P-160-D
0
GE-NE-0000-
0060-9229-TR-R3
Nuclear Management Company, LLC Monticello Nuclear
Generating Plant Extended Power Uprate Task T0400
Containment System Response
3
Heat 001M64068
COC- 1 (217) 2 A106 Schedule 80 SMLS Pipe
03/02/15
Heat 00A132529
COC- 2 (4510) 1A106 Schedule 80 SMLS Pipe
01/02/15
LOT 59464
CMTR- 100- A105 Nuclear 90 Elbow
07/22/14
MPS-0522
Vacuum Breaker Valves
01/14/86
MPS-0567
Specification Hollow Metal Doors, Frames, Hardware
0
MPS-1010
Piping Materials, Classification and Standards for the MNGP
30
MPS-1100
Specification for the Analysis Piping and Piping Support
Systems
11
MPS-2172
Specification for the Procurement of Emergency Diesel
System Diesel Oil Transfer Pumps
3
SAFER/GESTR-LOCA Loss of Coolant Accident Analysis
1
National Electric Code
2011
NSP-43-103
Specification for Vacuum Breaker Replacement Parts
0
NSP-53-103
Wetwell to Drywell Vacuum Breaker Replacement Parts
1
OE Eval
Part 21 on C&D Technologies Battery Cells - Misaligned
Separators
01/15/15
P.O.49546
Schulz Certificate of Conformance, RHR Motor Overhaul
04/13/15
P.O. 205-AB841
COC- Main Steam Safety Relief Valve
11/02/74
P.O. 56112
COC-Pressure Relief Valve
02/27/15
SCR 02-0324
USAR 5.2.1.2.3, Vent and Vacuum Relief System, Rev. 19
0
SCR 14-0541
EC 23085 EDG Fuel Oil Train Separation Screening
4
SRI 96-003
Locked Valve Program Improvements and Associated USAR
Changes
0
TC-15991
RHRSW 13 Motor/Pump Curves
12/08/11
TP-ESI506100
Functional Test Procedure Fuel Transfer Pump/Motor Asm
1
9
MODIFICATIONS
Number
Description or Title
Date or
Revision
DC79M070
Modify Drywell to Torus Vacuum Breakers, Add 1 and Add 2
0
Column Gaskets not Required on RHRSW Pumps
0
RHRSW Pump Impeller Material Change
0
Restoration of Motor Overload Margins in MCC-134/144
0
EDG Fuel Oil Transfer System Modification Support
7/10/14
EDG Fuel Oil Train
0
EDG Fuel Oil Train Separation
0
Drywell to Torus Vacuum Breakers- Remove Upper Portion of
Test Actuator Piston Rod
0
Alternate Nitrogen Bottle Change-out Check Valves
0
ECN25569
EDG Fuel Oil Train Separation
0
OPERABILITY EVALUATIONS
Number
Description or Title
Date
01430505-01
No Analysis Found for HELB at MO-2078 RCIC Steam
Supply and its Effect on MCC-311
05/29/14
01431915-01
Do the ESW and DGN (FSW & RSW also) Systems Remain
Operable While Bypassing the Basket Strainers for Periodic
Cleaning
06/14/14
01442471-01
RCIC High Steam Flow dp Switch Found Out of Tolerance
08/15/14
01478212
Past Operability AO-2382A Vacuum Breaker
06/25/15
PROCEDURES
Number
Description or Title
Revision
0036-01
ECCS Emergency Bus Undervoltage Test and ECCS Loss of
Normal Auxiliary Power Test
30
0137-A
LLRT-LRM-Makeup Flow Method
6
0197-01
- 13 250 Vdc Battery Capacity Test
24
0255-08-ID-03
RCIC CV-2104 Air Accumulator Check Valve (AI-612) Leak
Rate Test
20
0294
SRV Position Indication and Low Set System Instrumentation
Checks
29
1136
RHR Heat Exchanger Efficiency Test
33
1374
Monthly Operability Test of No.13 Diesel Generator
19
1388
13 DG Auto Start/Loading Test
13
1401-01
Locked Valve Alignment
23
1444
Pre and Post Severe Weather Inspection Checklist
10
4050-PM
Torus to Drywell Vacuum Breaker Seal Replacement
8
4280-03-PM
SRV Refurbishment and As-Left Steam Testing
40
4525-PM
NO. 13 & 16 Battery Charger Preventive Maintenance
12
8153
Powering Div. II 250VDC Battery Chargers from #13 Diesel
5
A.6
Acts of Nature
52
B.02.03-01
Reactor Core Isolation Cooling
5
B.03.04-01
Residual Heat Removal System
12
10
PROCEDURES
Number
Description or Title
Revision
B.07.01-02
Operations Manual
21
B.08.01.03-01
RHR Service Water System
10
B.08.01.03-05
RHR Service Water System - System Operation
46
B.08.07-05
Extreme Cold Weather Procedure
45
B.08.08-01
Plant Communications System
6
B.08.08-02
Plant Communications System
4
B.08.11-05
Diesel Oil System
37
B.08.13-05
Control Room H&V and EFT - System Operation
29
B.09.15-01
Non-Essential Diesel Generator
5
B.09.15-05
Non-Essential Diesel Generator
15
C.4-B.08.07.A
Ventilation System Failure - Abnormal Procedures
28
C.4-B.09.02.A
Abnormal Procedure, Station Blackout
46
C.4-I
Plant Flooding
14
EWI-08.13.02
Motor Program
10
FP-E-MOD-02
Engineering Change Control
16
FP-E-MOD-08
Engineering Change Notices
8
FP-E-MOD-10
Modification Turnover and Closeout
13
FP-E-RTC-02
Equipment Classification
11
FP-E-SE-05
System Engineering Walkdowns
0
FP-PA-OE-01
Operating Experience Program
21
MPS-1124
Common Motor Repair and Refurbishment Specification
1
MWI-3-M-2.01
AC Electrical Load Study
14
NWS-R-26
NWS Technologies Repair of Target Rock 3 Stage Main Steam
1
NWS-T-15
NWS Safety Valve Test Procedure for Monticello Nuclear Plant
Target Rock 67F Main Steam Safety Relief Valves
7
Monitor ADS Pneumatic Supply
7
OWI-02.03
Operator Rounds
63
OWI-03.07
Time Critical Operator Actions
10
WORK DOCUMENTS
Number
Description or Title
Date
00049081
0114 RCIC System Test RX Press <165 psig Cycle
05/29/15
00061207
V-EF-40B, Clean, Repair or Replace Flow Element
11/09/10
00106906
Preoperational Testing Drywell to Torus Vacuum Breakers
04/01/01
00106908
Preoperational Testing Drywell to Torus Vacuum Breakers
04/01/01
00113519
SRV Pilot and 2nd Stage Pilot Valve Assembly Inspections,
Refurbishment, and Steam Testing
03/15/02
00123486
PM 4280-1, RV-2-71E
05/21/03
00123487
PM 4280-2, RV-2-71E
05/21/03
00143657
SRV Pilot Valve Assembly (Pilot & 2nd Stage) Change Out
05/25/00
00143658
SRV Main Stage Valve Assembly
02/24/00
00387708
Test Data Evaluation for AOV CV-2104 Rising Stem Valve
04/10/13
00390096
PI-1982, Install Remaining 3 Anchor Bolts in Stanchion
09/09/09
00394602
MO-2075 Disassembly/Inspection
04/12/11
11
WORK DOCUMENTS
Number
Description or Title
Date
00402785
P-109C, Rebuild Spare per 4214-PM
02/16/11
00406241
V-EF-40B Discharge Flow Out of Specification
09/10/10
00414155
TD-152-503, Perform Relay PM
07/15/11
00414333
FI-9195 has Low Flow for V-EF-40B
02/02/11
00415157
TD-4KVB-21, Perform Breaker PM
12/21/12
00416774
Replace 13RHR Pump Cables
06/28/13
00422669
PMT Failure for EF-40B
07/20/11
00423219
V-EF-40B Discharge Flow is Below Spec.
06/04/11
00430036
Replace Vacuum Breaker Air Lines per EC 20501
05/31/13
00440830
MECH - Rd 1 4120-PM on S-200 (Terry Turbine) - All Steps
05/24/15
00441556
TD-Bus 15 Relays, Perform 4850-915 PM
05/21/13
00441564
TD-4KV Bus-15, Perform PM 4858-15
05/15/13
00458886
0255-08-111-1 RCIC Comprehensive PMP & VLV Tests
11/11/13
00462919
ELEC-D-52 Charger, Perform Charger 4525-PM
06/29/15
00463771
Re-Build P-202C
05/12/14
00476238
Comprehensive 13 RHR PMP & VLV Tests
12/09/14
00490342
0255-05-III-3A Comp 13 RHRSW Pump & Valve Test
09/10/14
00490607
Check Stroke Capabilities of Actuators
05/24/15
00490645
OSP-APR-0568 SRV Functional Tests
05/23/15
00490709
0255-07-IB-4 MS SRV Pilot Valve Assembly As-Found Check
05/05/15
00490762
Reactor Coolant Pressure Boundary Leak Test
05/24/15
00490806
0255-08-ID-3 RCIC CV-2104 Air Accum Check Valve LRT
05/11/15
00490905
0214 Drywell to Torus Vacuum Breaker Cycle Leakage Check
05/27/15
00490910
0127 Drywell to Torus Vacuum Breaker
05/06/15
00490910
Drywell-Torus Vacuum Breaker Inspection, Functional Tests, &
Calibration Maintenance of Position Indication & Alarm System
05/14/15
00490947
0137-07-A RX STM SUP VLV LLRT W/RX Press By Air
04/23/15
00490969
PM 4900-1 for MO-2075
02/08/04
0491182
SRV Position Indication and Low Set System Instrumentation
Checks
05/11/15
00491185
0255-07-IB-1 Main Steam AM SRV Bench Checks & Inspection
05/05/15
00491186
0131 Safety Relief Valve Bellows Monitor Check
04/27/15
00491211
- 13 250 Vdc Battery Capacity Test
04/14/15
00491333
0255-03-IA-2B Core Spray Valve Position Indication Test
05/24/15
00497230
Investigate Repair as Required AO-2382A
05/13/15
00504345
0255-08-IA-1 RCIC Quarterly Pump and Valve Tests
02/14/15
00504629
0269 Fire Protection System Valve Check
06/17/15
00505386
EC 23085 EDG Fuel Oil Train Separation and Pre-Op Testing
05/09/15
00505603
RHR Loop A Quarterly Pump and Valve Tests
03/02/15
00505605
02551-05-IA-1-1A RHRSW QRTLY PMP & Valve Test
06/06/15
00508701
Inspect and Rebuild DW to Torus Vacuum Bkr Air Actuators
05/14/15
00509926
0465-01 DIV 1 and 2 EFT Monthly Operation
05/24/15
12
WORK DOCUMENTS
Number
Description or Title
Date
00510950
0143 Drywell to Torus Monthly Vacuum Breaker Check
06/12/15
00510958
0255-04-IA-1-1 RHR Loop A Quarterly Pump and Valve Tests
06/08/15
00511299
1374 Monthly Oper Test of No.13 Diesel
06/15/15
00511829
0465-01 Div 1 and 2 EFT Monthly Operation
06/24/15
00513283
Operations TRB Side CK List Weekly Procedure
07/16/15
13
LIST OF ACRONYMS USED
Agencywide Document Access Management System
AN2
Alternate Nitrogen
Air-Operated Valve
ANSI
American National Standards Institute
Action Request
Corrective Action Program
Component Design Bases Inspection
CFR
Code of Federal Regulations
Emergency Filtration Train
Environmental Qualifications
IEEE
Institute of Electrical & Electronics Engineers
IMC
Inspection Manual Chapter
IN
Information Notice
Inservice Testing
Loss of Coolant Accident
Motor-Operated Valve
Non-Cited Violation
Net Positive Suction Head
NRC
U.S. Nuclear Regulatory Commission
Publicly Available Records System
psig
Pounds Per Square Inch Gauge
Reactor Core Isolation Cooling
Regulatory Issue Summary
Station Blackout
Significance Determination Process
Service Information Letter
Systems, Structures, and Components
Task Interface Agreement
Thermal Overload
TS
Technical Specification
Updated Safety Analysis Report
Unresolved Item
Vac
Volts Alternating Current
Vdc
Volts Direct Current
P. Gardner
-2-
In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public
Inspections, Exemptions, Requests for Withholding, of the NRC's "Rules of Practice," a copy
of this letter, its enclosure, and your response (if any) will be available electronically for public
inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS)
component of the NRC's Agencywide Documents Access and Management System (ADAMS).
ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html
(the Public Electronic Reading Room).
Sincerely,
/RA/
Christine A. Lipa, Chief
Engineering Branch 2
Division of Reactor Safety
Docket No. 50-263
License No. DPR-22
Enclosure:
Inspection Report 05000263/2015007;
w/Attachment: Supplemental Information
cc w/encl: Distribution via LISTSERV
DISTRIBUTION w/encl:
RidsNrrDorlLpl3-1 Resource
RidsNrrPMMonticello
RidsNrrDirsIrib Resource
Cynthia Pederson
DRPIII
DRSIII
ROPreports.Resource@nrc.gov
ADAMS Accession Number ML15245A785
Publicly Available
Non-Publicly Available
Sensitive
Non-Sensitive
To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copy
OFFICE
RIII
RIII
RIII
RIII
NAME
BJose for ADunlop:cl
ADunlop
CLipa
DATE
08/27/15
09/01/15
09/02/15
OFFICIAL RECORD COPY