ML072050572
| ML072050572 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 07/19/2007 |
| From: | Cain L NRC/RGN-II/DRS/EB1 |
| To: | Tynan T Southern Nuclear Operating Co |
| References | |
| IR-07-006 | |
| Download: ML072050572 (36) | |
See also: IR 05000424/2007006
Text
July 19, 2007
Southern Nuclear Operating Company, Inc.
ATTN: Mr. T. E. Tynan
Vice President - Vogtle
Vogtle Electric Generating Plant
7821 River Road
Waynesboro, GA 30830
SUBJECT: VOGTLE ELECTRIC GENERATING PLANT- NRC COMPONENT DESIGN
BASIS INSPECTION REPORT 05000424/2007006, 05000425/2007006
Dear Mr. Tynan:
On May 25, 2007, the U. S. Nuclear Regulatory Commission (NRC) completed an inspection at
your Vogtle Electric Generating Plant Units 1 and 2. The enclosed inspection report documents
the inspection findings which were discussed with you on May 25, 2007, and on June 4, 2007,
with Mr. J. Williams 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.
Based on the results of this inspection, the inspectors identified two findings of very low safety
significance (Green). These two findings were determined to involve violations of NRC
requirements. However, because of their very low safety significance and because they are
entered into your corrective action program, the NRC is treating these violations as non-cited
violations consistent with Section VI.A.1 of the NRCs Enforcement Policy. If you contest any of
these non-cited violations you should provide a response within 30 days of the date of this
inspection report, with the basis for your denial, to the United States Nuclear Regulatory
Commission, ATTN: Document Control Desk, Washington DC 20555-0001, with copies to the
Regional Administrator, Region II; the Director, Office of Enforcement, U. S. Nuclear Regulatory
Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at Vogtle.
SNC 2
In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its
enclosure, and your response (if any) will be available electronically for public inspection in the
NRC Public Document Room or from the Publicly Available Records (PARS) component of
NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Loyd M. Cain, Acting Chief
Engineering Branch 1
Division of Reactor Safety
Docket Nos.: 50-424,40-425
Enclosure: NRC Inspection Report 05000424/2007006, 05000425/2007006
w/Attachment: Supplemental Information
cc w/encl:
J. T. Gasser
Executive Vice President
Southern Nuclear Operating Company, Inc.
Electronic Mail Distribution
L. M. Stinson, Vice President, Fleet Operations Support
Southern Nuclear Operating Company, Inc.
Electronic Mail Distribution
N. J. Stringfellow
Manager-Licensing
Southern Nuclear Operating Company, Inc.
Electronic Mail Distribution
M. Caston
Southern Nuclear Operating Company, Inc.
Bin B-022
P. O. Box 1295
Birmingham, AL 35201-1295
Director, Consumers' Utility Counsel Division
Governor's Office of Consumer Affairs
2 M. L. King, Jr. Drive
Plaza Level East; Suite 356
Atlanta, GA 30334-4600
(cc w/encl contd - See page 3)
SNC 3
(cc w/encl contd)
Office of the County Commissioner
Burke County Commission
Waynesboro, GA 30830
Director, Department of Natural Resources
205 Butler Street, SE, Suite 1252
Atlanta, GA 30334
Manager, Radioactive Materials Program
Department of Natural Resources
Electronic Mail Distribution
Attorney General
Law Department
132 Judicial Building
Atlanta, GA 30334
Laurence Bergen
Oglethorpe Power Corporation
Electronic Mail Distribution
Resident Manager
Oglethorpe Power Corporation
Alvin W. Vogtle Nuclear Plant
Electronic Mail Distribution
Arthur H. Domby, Esq.
Troutman Sanders
Electronic Mail Distributioin
Senior Engineer - Power Supply
Municipal Electric Authority
of Georgia
Electronic Mail Distribution
Reece McAlister
Executive Secretary
Georgia Public Service Commission
244 Washington Street, SW
Atlanta, GA 30334
Distribution w/encl: (See page 4)
SNC 4
Distribution w/encl:
B. Singal, NRR
C. Evans (Part 72 Only)
L. Slack, RII EICS
RIDSNRRDIRS
OE Mail (email address if applicable)
PUBLIC
___
OFFICE RII:DRS RII:DRS RII:DRS RII:DRS contractor contractor RII: DRP
SIGNATURE RA RA RA RA RA RA RA
NAME RBerryman SRose DMas-Penaranda LCain MShlyamberg JChiloyan SShaeffer
DATE 06/22/2007 06/22/2007 06/21/2007 07/19/2007 06/22/2007 06/22/2007 07/18/2007
E-MAIL COPY? NO NO NO YES NO NO YES
U.S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos.: 50-424, 50-425
Report Nos.: 05000424/2007006, 05000425/2007006
Licensee: Southern Nuclear Operating Company
Facility: Vogtle Electric Generating Plant, Units 1 and 2
Location: Waynesboro, GA 30830
Dates: April 23 - May 25, 2007
Inspectors: R. Berryman, P.E., Senior Reactor Inspector (Lead)
S. Rose, Senior Reactor Inspector
D. Mas-Penaranda, Reactor Inspector
M. Shlyamberg, Contractor
J. Chiloyan, Contractor
Accompanying
personnel: A. Issa, P.E., Reactor Inspector (trainee)
Approved by: Loyd M. Cain, Acting Chief
Engineering Branch 1
Division of Reactor Safety
Enclosure
2
SUMMARY OF FINDINGS
IR 05000424/2007006, 05000425/2007006; 4/23/2007 - 4/27/2007, 5/7/2007 - 5/11/2007,
5/21/2007 - 5/25/2007; Vogtle Electric Generating Plant Units 1 and 2; Component Design
Bases Inspection.
This inspection was conducted by a team of three NRC inspectors and two NRC contractors.
Two Green non-cited violations were identified during this inspection. The significance of most
findings is indicated by their color (Green, White, Yellow, Red) using IMC 0609, Significance
Determination Process (SDP). Findings for which the SDP does not apply may be Green or be
assigned a severity level after NRC management review. The NRCs program for overseeing
the safe operation of commercial nuclear power reactors is described in NUREG-1649,
Reactor Oversight Process, Revision 4, dated December 2006.
A. NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green. The inspectors identified a violation of 10 CFR 50, Appendix B, Criterion
XI, Test Control, for failure to implement a test program to assure that all
installed safety related molded case circuit breakers (MCCBs) will perform
satisfactorily in service. Under postulated electrical fault conditions, failure of
one of these circuit breakers to operate properly would lead to either a loss of
power to safety-related components or lead to a potential for compromising other
equipment on a single fault that the MCCB was designed to isolate. The 1A, 1B,
2A, and 2B motor driven auxiliary feedwater pump (MDAFWP) recirculation valve
motor operators and the MDAFWP room cooler fans for both trains at both units
receive their power through MCCBs.
This finding is more than minor because it is associated with the Mitigating
Systems Cornerstone attribute of Procedure Quality. It impacts the cornerstone
objective of ensuring the availability, reliability, and operability of the MDAFW
pumps to perform their intended safety function during a design basis event.
The inspectors assessed the finding using the SDP and determined the finding
was of very low safety significance (Green) because the inspectors found no
documented history of in-service failures of MCCBs rendering safety-related
equipment inoperative. This issue is documented in the corrective action
program as condition report (CR) 2007105855. (Section 1R21.2.9)
- Green. The inspectors identified a violation of 10 CFR 50, Appendix B, Criterion
III, Design Control, for failure to evaluate the impact of an increase of the
residual heat removal (RHR) system pressure during the RHR pump operation in
a minimum flow alignment in determining the maximum dP across the
containment emergency sump isolation valves 1/2-HV-8811A/B, which could
have challenged the capability of these motor operated valves (MOVs) to open
following a small break loss of coolant accident (SBLOCA). Maximum system
pressure would occur following a SBLOCA.
3
This finding is more than minor because it is associated with the Mitigating
Systems Cornerstone attribute of Design Control. It impacts the cornerstone
objective of ensuring the availability, reliability, and operability of the containment
emergency sump isolation valves to perform their safety function during a
SBLOCA event. The calculation deficiencies represented reasonable doubt
regarding the operability of MOVs 1/2-HV-8811A/B pending the outcome of
additional calculations initiated after the inspectors questioned the condition.
The lack of an accurately calculated maximum dP across these MOVs created
the possibility for repairs or modifications to be performed while using an
incorrect dP value as a design input. The inspectors assessed the finding using
the SDP and determined the finding was of very low safety significance (Green)
because there was not a loss of safety system function based upon the
inspectors verification of the SNC analysis that the containment emergency
sump isolation valves 1/2-HV-8811A/B were currently operable. This issue is
documented in the corrective action program as CR 2007100247 and CR
2007105848. (Section 1R21.2.17)
B. Licensee-Identified Violations
None
4
REPORT DETAILS
1. REACTOR SAFETY
Cornerstones: Mitigating Systems and Barrier Integrity
1R21 Component Design Bases Inspection (71111.21)
.1 Inspection Sample Selection Process
The inspectors selected risk significant components and operator actions for review
using information contained in the licensees Probabilistic Risk Assessment (PRA). In
general, this included components and operator actions that had a risk achievement
worth factor greater than two or Birnbaum value greater than 1 X10E-6. The
components selected were located within the auxiliary feedwater (AFW) system, nuclear
service cooling water system (NSCW), emergency diesel generator (EDG), 480 VAC
electrical system, 4160 VAC electrical system, 125 VDC electrical system, and the
offsite power system. The sample selection included 19 components, five operator
actions, and six operating experience items. Additionally, the inspectors reviewed two
modifications by performing activities identified in IP 71111.17, Permanent Plant
Modifications, Section 02.02.a. and IP 71111.02, Evaluations of Changes, Tests, or
Experiments.
The inspectors performed a margin assessment and detailed review of the selected risk-
significant components to verify that the design bases have been correctly implemented
and maintained. This design margin assessment considered original design issues,
margin reductions due to modification, or margin reductions identified as a result of
material condition issues. Equipment reliability issues were also considered in the
selection of components for detailed review. These included items such as failed
performance test results, significant corrective action, repeated maintenance,
maintenance rule (a)1 status, Regulatory Issue Summary 05-020 (formerly Generic
Letter 91-18) conditions, NRC resident inspector input of problem equipment, system
health reports, industry operating experience and licensee problem equipment lists.
Consideration was also given to the uniqueness and complexity of the design, operating
experience, and the available defense in depth margins. An overall summary of the
reviews performed and the specific inspection findings identified are included in the
following sections of the report.
5
.2 Results of Detailed Reviews
.2.1 Plant Wilson
a. Inspection Scope
The inspectors reviewed design basis documentation, electrical one line diagrams, and
the Updated Final Safety Analysis Report (UFSAR) to identify the design and licensing
basis requirements for the Plant Wilson power generation and transmission facilities to
supply offsite power to Vogtle Units 1 and 2 engineered safety features (ESF) loads
when one of the reserve auxiliary transformers (RATs) was taken out of service for
maintenance. The review focused on evaluating the capability of the 13.8 KV distribution
cable, the distribution substation switchyard equipment, the black-start diesel generator,
the combustion turbines, and associated auxiliaries. The inspectors reviewed station
blackout (SBO) operating procedures, station drawings, system description documents,
equipment specifications, short circuit and voltage profile calculations, and protective
relay settings to determine the individual system and component operating requirements
to supply offsite power to the standby auxiliary transformer (SAT). The inspectors also
reviewed circuit breaker, battery, and relay maintenance and calibration test results to
verify that the functional test results satisfied the design basis performance
requirements. Completed surveillance test results were also reviewed to verify the
capability of the Plant Wilson to supply offsite power to the plant through the SAT during
SBO events. The inspectors interviewed system engineers and control room operators
to verify that the equipment configuration was consistent with station drawings and to
verify that the protocols between the control room operators, Plant Wilson operators,
and the transmission system operators were consistent with the operating procedures.
Field walkdowns were conducted of the Plant Wilson combustion turbines, associated
battery and control rooms, the black-start diesel generator and associated battery and
control rooms, switchyard equipment including the 13.8 KV distribution cable
terminations, and the associated control house to observe general material conditions.
A field walkdown of the Vogtle 13.8 KV/4160 VAC switch yard was also performed to
observe the material condition of the 13.8 KV cable terminations.
b. Findings
No findings of significance were identified.
.2.2 Standby Auxiliary Transformer - ANXRA
a. Inspection Scope
The inspectors reviewed drawings, design basis documents and the UFSAR to identify
the design and licensing basis requirements for the standby auxiliary transformer (SAT).
The inspectors reviewed vendor specifications, nameplate data, one-line diagrams,
protective relay settings, short circuit and voltage drop calculations, and ESF loading
requirements to evaluate the capability of the SAT to supply the voltage and current
requirements to one train of ESF loads. The inspectors performed independent
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transformer protective relay trip setpoint calculations to verify that the applied protective
relay setting calculations had adequately accommodated for the transformer being
energized, through-faults, and maximum loading conditions. The relay settings review
included the transformer overall differentials, phase and ground over-currents, and
distance relays associated with the 13.8 KV distribution cable from Plant Wilson. The
inspectors also reviewed the rating of the transformer neutral grounding resistor to verify
that it was adequate. The inspectors reviewed the results of completed transformer
preventive maintenance and relay setpoint calibrations to verify that the test results were
within their allowed limits. The inspectors performed a visual inspection of the
observable portions of the SAT and the associated neutral grounding resistor bank to
assess material condition.
b. Findings
No findings of significance were identified.
.2.3 Reserve Auxiliary Transformers - 1NXRA and 1NXRB
a. Inspection Scope
The inspectors reviewed the design basis descriptions, equipment specifications,
drawings, equipment name plate data, voltage drop calculations, and short circuit and
load flow calculations to determine if the voltage and current requirements of the ESF
loads could be powered by Reserve Auxiliary Transformers (RATs) 1NXRA and 1NXRB.
The review was also conducted to verify that circuit switchers 1IM1A and 1IM1B, low
voltage side feeder cables, and circuit breakers 1AA02 and 1BA03 were adequately
sized. Protective relay trip setting calculations were reviewed to verify whether
adequate protection coordination margins were provided. The relay settings review
included the transformer overall differentials and the ground overcurrent relays. The
inspectors also reviewed the ratings of the transformer neutral grounding resistors and
verified that the ground relay trip settings were coordinated with the offsite 230KV
transmission system relay schemes to prevent inadvertent tripping of circuit breakers or
circuit switchers. The inspectors reviewed the results of completed transformer
preventive maintenance and relay calibrations to verify that the test results were
satisfactory. The loss of voltage and degraded voltage relay settings were also reviewed
to verify that they satisfied the requirements of Technical Specifications (TS) 3.8.1.
Records of 230KV and 4160V system voltage profiles were reviewed to verify that they
were consistent with the design basis assumptions. The inspectors performed a visual
inspection of the observable portions of Unit 1 RATs and the associated neutral
grounding resistors to assess the installation configuration and material condition. The
inspectors also performed walkdowns of the 4160 V switchgear 1AA02 and 1BA03 to
verify that the installed local and remote circuit breaker control switches, breaker
position indicating lights, and the operator actions for the resetting of the lockout relays
were consistent with design drawings.
7
b. Findings
No findings of significance were identified.
.2.4 Emergency Diesel Generators - DG1A & DG1B (Electrical)
a. Inspection Scope
The inspectors reviewed drawings, design basis descriptions and the UFSAR to identify
the design and licensing basis requirements for the EDGs. The inspectors reviewed the
EDG loading calculations including voltage, frequency, current, and loading sequences
during postulated loss of offsite power and loss of coolant accidents to verify the
capability of the EDGs to perform their intended safety function. Short circuit
calculations were reviewed to ensure that the ratings of the generator output breakers
were adequate. The inspectors also performed independent calculations of available
phase and ground short circuit currents to ensure that the maximum system short circuit
duty was within equipment rating. Protective relay setpoint calculations and setpoint
calibration test results were reviewed to assess the adequacy of protection during
testing and emergency operations and to ensure that excessive setpoint drift had not
taken place. The generator grounding scheme was also reviewed to determine the
adequacy of ground overcurrent relay coordination. The electrical drawings and
calculations that describe the generator output breaker control logic and interlocks were
reviewed to determine whether the breaker opening and closing control circuits were
consistent with design basis documents. The inspectors also reviewed surveillance test
results to verify that applicable test acceptance criteria and test frequency requirements
for the EDGs were satisfied. The inspectors conducted a field walkdown of the electrical
relay cabinets, output breaker control switches and breaker position indicating lights to
assess material conditions and to verify that the installed configuration was consistent
with system drawings. The inspectors also visited the control room to observe meter
readings, switch positions, indicating lights and annunciator alarm panels associated
with the 1A and 1B EDG to verify that they were consistent with design basis documents
and operating procedures. The inspectors also interviewed system engineers regarding
design aspects and operating history for the EDGs.
b. Findings
No findings of significance were identified.
.2.5 4160/480 V Load Center Transformer - 1AB05X
a. Inspection Scope
The inspectors reviewed calculations, design basis descriptions, and drawings to verify
that the loading of 4160/480 V transformer 1AB05X, 4160 V breaker 1AA0221 and 480
V breaker 1AB0501 was within the corresponding transformer and switchgear ratings.
The inspectors reviewed design assumptions and calculations related to short circuit
currents, voltage drop and protective relay settings associated with transformer 1AB05X
8
and the feeder cables to verify that they were appropriate. The inspectors reviewed
operating procedures and design drawings to assess the adequacy of the ground
detection design of 480 V switchgear 1AB05 and to verify that the grounding equipment
ratings would not challenge the control room operator response time to ground alarms.
The inspectors reviewed a sample of completed maintenance and functional verification
testing results to verify that the high and low voltage cable feeders associated with
transformer 1AB05X were capable of supplying the power requirement of the 480 V bus
1AB05 during normal and postulated accident conditions. The inspectors performed a
sample of independent short circuit and voltage drop calculations to verify that the
values stated in design bases documents were appropriate. The inspectors reviewed
system health reports, interviewed system engineers, and conducted a field walkdown of
load center transformer 1AB05X, 4160 V circuit breaker 1AA0221, and 480 V circuit
breaker 1AB0501 to verify that equipment alignment and nameplate data was consistent
with design drawings and to assess the material condition of 1AB05X.
b. Findings
No findings of significance were identified.
.2.6 125 VDC Battery - 1CD1B and Output Breaker - 1CD101
a. Inspection Scope
The inspectors reviewed design calculations, system drawings, the UFSAR and TS
requirements for battery 1CD1B to verify that the battery size would satisfy the
functional and design basis requirements. These requirements included the safety
related and risk significant loads and the worst case minimum voltage. The review also
focused on verifying that the battery was adequately sized to supply the design duty
cycle of the 125 VDC system for a loss of coolant accident (LOCA) concurrent with a
loss of offsite power (LOOP) and SBO scenarios. The battery charger sizing calculation
was reviewed to verify whether it was consistent with the design and licensing bases.
The inspectors reviewed the short circuit duty and breaker trip settings calculations to
verify the calculated fault interrupting duties were within the interrupting current rating of
battery output breaker 1CD101 and that adequate protection coordination was provided.
The inspectors reviewed completed maintenance and trip setting test results on breaker
1CD101 to verify that the results were within acceptable limits. The inspectors reviewed
completed battery capacity test results with particular focus on the applied methodology
and acceptance criteria for battery capacity testing and whether the test was terminated
only after the battery terminal voltage reached the minimum voltage as specified in IEEE 450-1995, Recommended Practice for Maintenance, Testing, and Replacement of
Vented Lead-Acid Batteries for Stationary Applications. The inspectors interviewed
system and design engineers regarding design aspects and operating history and
reviewed a sample of condition reports to verify that material deficiencies were
adequately addressed. The inspectors performed a walkdown to evaluate the material
condition of the battery and battery chargers.
9
b. Findings
No findings of significance were identified.
.2.7 125 VDC Switchgear Bus - 1CD1
a. Inspection Scope
The inspectors reviewed drawings, system descriptions, nameplate data, circuit beaker
ratings, short circuit and voltage drop calculations to determine the capability of 125
VDC bus 1CD1 to supply required voltage and current. The inspectors reviewed a
sample of completed preventive maintenance testing results to verify that the test results
were within the allowed limits. The inspectors interviewed system engineers and
performed a visual inspection of the observable portions of 125 VDC bus 1CD1 to verify
that the installed configuration and instrument readings were consistent with design
drawings and procedures. A walkdown was also performed to evaluate the observable
material condition.
b. Findings
No findings of significance were identified.
.2.8 125 VDC Motor Control Center - CD1M
a. Inspection Scope
The inspectors reviewed system descriptions, drawings, circuit breaker nameplate data,
and calculations of short circuit and voltage drop to verify that 125 VDC MCC CD1M
could supply the voltage and current requirements of the safety related loads. The
inspectors reviewed a sample of completed preventive maintenance testing results to
verify that the test results were within the allowed limits. The inspectors interviewed
system engineers to review performance history and conducted a visual inspection of
MCC CD1M to verify that the installed configuration was consistent with design drawings
and to observe the material condition.
b. Findings
No findings of significance were identified.
.2.9 480 VAC Motor Control Center - 1ABF
a. Inspection Scope
The inspectors reviewed schematic diagrams and associated calculations to verify that
bus current, voltage, short circuit capability, and breaker coordination were adequate.
The review included verification that the voltage calculations would support equipment
operation under worst case conditions, that calculated bus loading did not exceed bus
10
ratings, and that calculated short circuit conditions did not exist which could exceed the
switchgear or circuit breaker ratings. The history of corrective actions and maintenance
was reviewed to assess the potential for component degradation and any corresponding
impact on design margins or performance. Field walkdowns were also performed to
determine the material condition of the switchgear
b. Findings
Introduction. The inspectors identified a finding of very low safety significance involving
a violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, for failure to implement
a test program to assure that all installed safety related MCCBs will perform
satisfactorily in service. Under postulated electrical fault conditions, failure of one of
these circuit breakers to operate properly would lead to either a loss of power to safety-
related components or lead to a potential for compromising other equipment on a single
fault that the MCCB was designed to isolate.
Description. During a walkdown of MCC 1ABF in the 1A EDG building, the inspectors
observed rust areas on MCC cubicle doors as well as on and around the operating
mechanism of several installed MCCBs. This created questions regarding potential
adverse effects that this would have on breaker functional performance and its effect on
the electrical fault interrupting capability and assurance of electrical protection
coordination. The inspectors reviewed several CRs concerning circuit breakers
installed in MCC 1ABF. Two CRs were generated concerning MCCBs which were
removed from MCC 1ABF for current trip setpoint changes. In both of these instances,
following the trip setting change, the MCCBs failed the post modification functional
verification tests and were discarded because the current trip setpoints could not be
adjusted. During the walkdown, the inspectors also observed that MCCs 1ABF,1BBF,
2ABF and 2BBF were located in close proximity to a source of outdoor moisture from
large dampers located within a few feet of the MCCs. The inspectors determined,
based on the history of MCCB failures throughout the industry, that the lack of
performance of periodic maintenance and functional verification tests on safety-related
MCCBs in a moisture-prone environment subjected these MCCBs to a potentially
increased vulnerability to fail to isolate faulted equipment, cause loss of protection
coordination, and the potential loss of power to all loads powered from the affected
MCC.
The 1A, 1B, 2A, and 2B motor driven auxiliary feedwater pump (MDAFWP) recirculation
valve motor operators and the MDAFWP room cooler fans for both trains at both units
are powered from MCCs 1ABF, 1BBF, 2ABF, and 2BBF respectively. The MDAFWP
recirculation valves are required to be operable to protect the MDAFWP from damage
during low flow conditions. The MDAFWP room cooler fans are required to be operable
to ensure adequate cooling to the pumps.
Analysis. Southern Nuclear Operating Companys (SNCs) failure to perform periodic
maintenance and functional verification tests on MCCBs used in safety-related
applications on MCCs 1ABF, 1BBF, 2ABF, and 2BBF was a performance deficiency.
This finding is more than minor because it is associated with the Mitigating Systems
11
Cornerstone attribute of Procedure Quality. It impacts the cornerstone objective of
ensuring the availability, reliability, and operability of the MDAFW pumps to perform their
intended safety function during a design basis event. The inspectors assessed the
finding using the SDP and determined the finding was of very low safety significance
(Green) because the inspectors found no documented history of in-service failures of
MCCBs rendering safety-related equipment inoperative. This finding was reviewed for
cross-cutting aspects and none were identified.
Enforcement. 10 CFR 50, Appendix B, Criterion XI, Test Control stated, in part, that test
programs shall be established to assure that all testing required to demonstrate that
structures, systems and components will perform satisfactorily in services. UFSAR
Section 8.1.4.3 C stated that the onsite electrical system was designed in accordance
with IEEE 308-1974, Criteria for Class IE Power Systems at Nuclear Generating
Stations. This standard recommended periodic tests be performed at scheduled
intervals to detect deterioration of equipment and to demonstrate operability of the
components that are not exercised during normal operation. Contrary to the above,
SNC did not establish adequate test control measures to assure that the protective
function of the 480 VAC safety-related MCCBs are periodically verified. This condition
has existed since plant initial operation. Because this finding is of very low safety
significance and was entered into SNCs corrective action program as CRs 2007105855,
this violation is being treated as a non-cited violation (NCV), consistent with Section
VI.A.1 of the NRC Enforcement Policy. (NCV 05000424,05000425/2007006-01,
Safety-Related 480 VAC Molded Case Circuit Breakers On MCC 1ABF, 1BBF, 2ABF,
and 2BBF Not Tested)
.2.10 NSCW Return Water Valves - 1668A, 1669A and High Temperature Switches
a. Inspection Scope
The inspectors reviewed the MOV calculations for the NSCW return water valves to
verify that appropriate design basis event conditions and degraded voltage conditions
were used as inputs into the determination of MOV actuator setpoints and actuator
sizing. Test results were reviewed to verify valve performance was monitored and that
performance degradation could be identified in a timely manner. The inspectors also
reviewed the elementary and schematic diagrams of the temperature switches
associated with the MOVs to verify that the circuitry satisfied the logic presented in the
design basis documentation. The instrument loop uncertainty calculations for the
temperature switches were reviewed to verify that plant instrument calibration
procedures properly incorporated the set point values delineated in the calculations of
record. The inspectors reviewed surveillance procedures and calibration test records for
the temperature switches to assess any adverse trends in instrument performance.
b. Findings
No findings of significance were identified.
12
.2.11 120 VAC Instrument Panel - 1AY2A
a. Inspection Scope
The inspectors reviewed schematic diagrams, design calculations, and vendor
specifications of connected loads to verify that panel loading limits, voltage adequacy,
short circuit capability, and protection coordination would support satisfactory operation
of connected loads. The review included verifying that AC voltage calculations assured
satisfactory voltage to the panel under worst case conditions, verifying that panel
loading did not exceed the panel rating, and reviewing short circuit calculations to verify
that a condition did not exist which could exceed the instrument panel and breaker
ratings. The inspectors also reviewed transformer loading calculations and schematic
diagrams to verify that the protection scheme would allow manual transfer to regulating
transformers and that the regulating transformers could carry the panel loads in the
event of the loss of the inverter. A field walkdown was also performed to determine the
material condition of the panel.
b. Findings
No findings of significance were identified.
.2.12 Sequencer 48 VDC Power Supply and Input/Output devices
a. Inspection Scope
The inspectors reviewed calculations to verify that the input voltage to the sequencer
under degraded voltage conditions would satisfy vendor recommendations. The
inspectors reviewed service and performance testing, preventive maintenance
procedures, and elementary diagrams for the sequencer. The review was performed in
order to verify that the specified acceptance criteria were met and that the equipment
capabilities were consistent with the licensing and design bases. The inspectors
reviewed uncertainty and setpoint calculations to verify that setpoint values were
consistent with TS requirements. A field walkdown was also performed to determine the
material condition of the components.
b. Findings
No findings of significance were identified.
.2.13 120/240 VAC Panel - 1AYB1
a. Inspection Scope
The inspectors reviewed schematic diagrams and design calculations to verify that panel
loading limits, voltage, short circuit capability, and protection coordination were
adequate. The review included verifying that AC voltage calculations would assure
adequate voltage to the panel under worst case conditions, verifying that panel loading
13
did not exceed panel rating, and reviewing short circuit calculations to verify that a
condition did not exist which could exceed the instrument panel or circuit breaker. The
inspectors reviewed vendor recommendations, maintenance records, CRs,
environmental qualification records, and a selected sample of work orders related to the
panel. A field walkdown was also performed to determine the material condition of the
panel.
b. Findings
No findings of significance were identified.
.2.14 EDG 1A/1B/2A/2B Building Ventilation
a. Inspection Scope
The inspectors reviewed the design basis information and supporting calculations
related to heat removal requirements for the diesel generator building ventilation and
combustion air supply.
b. Findings
No findings of significance were identified.
.2.15 Diesel Generator Fuel Oil Transfer Pumps - 1/2-2403-P4-001/2
a. Inspection Scope
The inspectors reviewed the design basis documentation to identify design requirements
related to flow, developed head, net positive suction head (NPSH), vortex formation,
minimum flow and runout protection for all operating conditions. The flow assumptions
in the UFSAR accident analysis were verified. Design calculations were reviewed to
verify that the design and licensing performance requirements were satisfied.
Calculations were reviewed to verify that available NPSH and measures taken to
prevent vortexing when the pumps are operating at a minimum diesel fuel storage tank
level were adequate.
b. Findings
No findings of significance were identified.
.2.16 NSCW Return Water Valves - 1/2-HV-1668A/B and 1/2-HV-1669A/B
a. Inspection Scope
The inspectors reviewed MOV calculations to verify that the design bases, system
conditions, and allowable degraded voltage conditions were used as design inputs to
determine adequate motor size for the actuators and establish set point values.
14
Additionally, the translation of design information into MOV testing acceptance criteria
was reviewed. MOV calculations and related testing documentation were reviewed to
assure that valve performance criteria allowed for anticipated maximum operating
pressure conditions and that appropriate torque switch settings were maintained.
Maintenance documentation was reviewed to determine that MOVs were periodically
tested and inspected to ensure that the design function was maintained.
b. Findings
No findings of significance were identified.
.2.17 Containment Emergency Sump Isolation Valves - 1/2-HV-8811A/B
a. Inspection Scope
The inspectors reviewed the calculations that analyzed the maximum differential
pressure (dP) across the containment emergency sump isolation valves that could still
support adequate MOV operation. This review identified that the containment sump
suction isolation valves, MOV 1/2-HV-8811A/B, were sized to open against a dP of 57
psid. The 57 psid value was postulated for post large break loss of coolant accident
(LBLOCA) conditions.
b. Findings
Introduction: The inspectors identified a violation of 10 CFR 50, Appendix B, Criterion III,
Design Control, for failure to evaluate the impact of an increase in the residual heat
removal (RHR) system pressure during RHR pump operation in minimum flow alignment
to determine the maximum differential pressure (dP) across containment emergency
sump isolation valves 1/2-HV-8811A/B, which could have challenged the capability of
these motor operated valves (MOVs) to open following a small break loss of coolant
accident (SBLOCA). However, maximum system pressure would occur following a
Description: During the review of calculation X4C1000U01, Differential Pressure
Calculations, Rev. 15, the inspectors found that the containment emergency sump
isolation valves were sized to open against a dP of 57 psid. The inspectors reviewed
the plant computer data of the last RHR pump surveillance test using procedure
14805-2, Residual Heat Removal and Check Valve IST and Response Time Tests, Rev.
31. The inspectors found that following a unit 2 RHR pump A train test on May 6, 2007,
the RHR system remained pressurized at a pressure of approximately 293 psig until
pressure was reduced to refueling water storage tank static pressure by bleeding off the
system in accordance with the step 5.1.22 of the test procedure.
The RHR pump test flow path was identical to what would occur following a Safety
Injection (SI) signal starting the RHR pumps during a SBLOCA with the pump operating
in a minimum flow alignment. Emergency operating procedure (EOP) 19000-C, E-0
Reactor Trip or Safety Injection, Rev. 32, directs operators to stop the RHR pumps if the
15
reactor coolant system (RCS) pressure is greater than 300 psig since the RHR pumps
would be operating at or near shutoff head and recirculating in a minimum flow
alignment. The EOP 19013-C, ES-1.3, Transfer To Cold Leg Recirculation, Rev. 26
directs the operators to open the containment emergency sump isolation valves
1/2-HV-8811A/B to transfer the RHR pump line up from injecting water into the RCS
from the RWST to recirculation cooling by taking a suction from the containment sump
and injecting this water into the suction of the SI pumps or directly into the RCS. Since
the dP across MOVs 1/2-HV-8811A/B would be significantly greater than that
determined in calculation X4C1000U01, the valves may not open, thus potentially
disabling the ability to take a suction from the containment sump. Plant EOPs do not
direct operators to bleed off pressure from the RHR system following the securing of the
RHR pumps during accident conditions.
The licensee initiated condition reports CR 2007100247 and CR 2007105848 to address
the questions raised by the inspectors. The licensee conducted an operability
evaluation documented in DOEJ-SM-C070401401-001, Maximum Opening Differential
Pressure for the Vogtle RHR Sump Valves (1/2HV8811A & B) at Steady State Voltage
of 88.5%, dated May 22, 2007, that demonstrated that the maximum available thrust
during postulated worst-case degraded voltage conditions would allow these valves to
open against a dP of 379 psid.
Analysis: Southern Nuclear Operating Companys failure to include the effect of RHR
operation at minimum flow in determining the maximum dP across the containment
emergency sump isolation valves 1/2-HV-8811A/B was a performance deficiency. This
finding is more than minor because it is associated with the Mitigating Systems
Cornerstone attribute of Design Control. It impacts the cornerstone objective of
ensuring the availability, reliability, and operability of the containment emergency sump
isolation valves to perform their safety function during a SBLOCA event. The calculation
deficiencies represented reasonable doubt regarding the operability of MOVs
1/2-HV-8811A/B. Additional calculations, initiated after the inspectors questioned the
condition, confirmed that the valves were able to perform their safety function at the
higher dP. The lack of an accurately calculated maximum dP across these MOVs
created the possibility for repairs or modifications to be performed while using an
incorrect dP value as a design input. The inspectors assessed the finding using the
SDP and determined the finding was of very low safety significance (Green) because
there was not a loss of safety system function based upon the inspectors verification of
the SNC analysis that the containment emergency sump isolation valves
1/2-HV-8811A/B were currently operable. This finding was reviewed for cross-cutting
aspects and none were identified.
Enforcement: 10 CFR 50, Appendix B, Criterion III, Design Control, states, in part, that
design control measures be established and implemented to assure that applicable
regulatory requirements and the design basis for structures, systems, and components
are correctly translated into specifications, drawings, procedures, and instructions.
Contrary to the above, SNC did not include the effects of RHR operation at minimum
flow in determining the maximum dP across the containment emergency sump isolation
valves 1/2-HV-8811A/B. Because this finding is of very low safety significance and was
16
entered into SNCs corrective action program as CR 2007100247 and CR 2007105848,
this violation is being treated as a NCV, consistent with Section VI.A.1 of the NRC
Enforcement Policy. (NCV 05000424,05000425/2007006-02, Maximum Differential
Pressure for Containment Emergency Sump Isolation Valves Not Calculated)
.2.18 Turbine Driven Auxiliary Feedwater (TDAFW) Pumps - 1/2-1302-P4-001 and Motor
Driven Auxiliary Feedwater (MDAFW) Pumps - 1/2-1302-P4-002/003
a. Inspection Scope
The inspectors reviewed design calculations and completed pump surveillance tests
from March 2005 through April 2007 to assess the adequacy of American Society of
Mechanical Engineers (ASME)Section XI testing for the TDAFW pumps and MDAFW
pumps and to ensure compliance with TS SR 3.7.5.2. TS SR 3.7.5.2 requires a
demonstration that the TDAFW and MDAFW pumps develop head at the flow test point
greater than or equal to required developed head every 31 days on a staggered basis.
b. Findings
Introduction. The inspectors identified an unresolved item (URI) for failure to evaluate
the effects of instrument accuracy, water temperature, or pump speed on surveillance
testing of TDAFWP 1/2-1302-P4-001 and MDAFWPs 1/2-1302-P4-002/003 in
accordance with TS Surveillance Requirement (SR) 3.7.5.2.
Description. The inspectors review of design calculation X4C1302S12, Auxiliary
Feedwater Pump Discharge Line Orifice Sizing, Rev. 1 and calculation X4C1302V04,
Auxiliary Feedwater Pumps Technical Specifications Verification, Rev. 5 identified the
following concerns:
- The results of calculation X4C1302S12 were not incorporated into calculation
X4C1302V04. Specifically, the resistance values for orifices currently installed in
the system were not used in the calculation to determine if the system could
satisfy the intended safety function by providing above the total minimum flow of
510 gpm. The inspectors determined that the failure to incorporate the results of
calculation X4C1302S12 into calculation X4C1302V04 resulted in
non-conservative acceptance criteria for the MDAFW pumps.
- The acceptance criterion for the TDAFW pumps was not based on a fluid
specific gravity of 1.0. A lower value was used based on an assumed
condensate storage tank (CST) temperature of 100 degrees F. This is a
non-conservative assumption since the actual CST temperature was lower than
this value on every test conducted over the last two years.
any allowances for instrument uncertainties.
17
The inspectorss review of 20 completed pump surveillances 14810-1, TDAFW Pump
Operability, Response Time and Check Valve IST, Rev. 36.2 and 14810-2, TDAFW
Pump Operability, Response Time and Check Valve IST, Rev. 30.2 from March 2005
through April 2007 identified the following concerns:
temperature of 100 degrees F.
- The test results were not normalized to a required speed of 4230 rpm. The Terry
turbine speed was changed from 4200 to 4230 rpm by a plant change
MDD-91-V1M016, Auxiliary Feedwater Rated Speed Change, April 4, 1991 to
increase the TS margin. The inspectors review identified that five surveillance
tests had recorded the as-found turbine speed to be 4200 rpm. The licensee
generated CR 2007105895 on May 23, 2007 to address this issue.
- The test results were not corrected for the applicable instrument error. The
licensee generated a documentation of engineering judgment (DOEJ) DOEJ-
SJ-C070401401-001, Channel Uncertainty - AFW Pump Suction and Discharge
Pressures - P-5110, P-5107, dated May 7, 2007, which demonstrated that the
combined instrument uncertainty for both suction and discharge gauges is 28.6
psig, when using information from the plant computer. The licensee also
evaluated the effects of the temperature correction and concluded that the
recent TDAFWP test results when corrected for both instrument uncertainty and
temperature met the acceptance criterion. The licensee generated CR
2007105436 and CR 2007105713 to address the inspectors concerns.
The inspectors review of completed pump surveillances 14807-1, Motor Driven Auxiliary
Feedwater Pump and Check Valve In-Service and Response Time Test, Rev. 29.2 and
14807-2, Motor Driven Auxiliary Feedwater Pump and Check Valve In-Service and
Response Time Test, Rev. 24.1 from March 2005 through April 2007 identified that
similar to the TDAFW pumps, the test results for the MDAFW pumps were not corrected
for the instrument error.
Analysis. The inspectors determined that SNCs failure to consider the effects of
instrument accuracy, water temperature, or pump speed during surveillance testing of
TDAFW pumps 1/2-1302-P4-001 and MDAFW pumps 1/2-1302-P4-002/003 to ensure
compliance with TS SR 3.7.5.2 is a performance deficiency. This URI is more than
minor because it is associated with the Mitigating Systems Cornerstone attribute of
Procedure Quality. It impacts the cornerstone objective of ensuring the availability,
reliability, and operability of the MDAFWPs and TDAFWPs to perform their safety
function. The inspectors could not make a determination of impact on operability of the
AFW pumps due to questions concerning the validity of the system curves and pump
curves identified in Section 1R21.2.19.
Enforcement. This URI was entered into the licensees corrective action program as CR
2007105436, CR 2007105713, CR 2007105870, and CR 2007105895 with actions to
evaluate the ability of the AFW pumps to meet the TS SR 3.7.5.2. This issue is
18
identified as URI 05000424,05000425/2007006-04, Ability to Satisfy TS SR 3.7.5.2.
This item is unresolved pending an NRC review of the effects of instrument
uncertainties, water temperature, and pump performance in the evaluation of results of
surveillance testing of TDAFW pumps 1/2-1302-P4-001 and MDAFW pumps
1/2-1302-P4-002/003. (URI 05000424,05000425/2007006-03, Ability To Satisfy TS SR 3.7.5.2)
.2.19 Capability of Auxiliary Feedwater System to Meet Design and Licensing Requirements
a. Inspection Scope
The inspectors reviewed the design basis documentation to identify design requirements
related to flow, developed head, available NPSH, vortex formation, minimum flow and
runout protection for all AFW system operating conditions. The AFW flow assumptions
in the UFSAR accident analysis were also reviewed. Design calculations and periodic
test documentation and results were reviewed to verify that the AFW system design and
licensing performance requirements were satisfied. Calculations were reviewed to verify
the adequacy of available NPSH and measures taken to prevent suction vortexing.
Maintenance, in-service testing (IST), periodic surveillance testing, corrective actions,
and design change history were reviewed to assess the system for potential component
degradation and subsequent impacts on design margins or performance.
b. Findings
Introduction. The inspectors identified an URI for failure to verify the AFW system
design was bounded by all operating and design limitations.
Description. The inspectors reviewed design calculation X4C1302S12, Auxiliary
Feedwater Pump Discharge Line Orifice Sizing, Rev. 1 and identified that this
calculation established the limiting AFW system flow rates and system resistance values
after the system limiting flow orifices were modified during plant pre-operational testing.
This calculation developed a hydraulic model for the current configuration of the AFW
system. The system resistance was calculated by measuring flows and corresponding
dPs. This calculation predicted relatively small margins for the ability to satisfy the
intended safety function by providing above the total minimum flow of 510 gpm to intact
steam generators while also not exceeding the value of 1050 gpm used as a bounding
condition for mass introduction in the containment analysis for a faulted steam
generator. For minimum flow cases, this calculation predicted that the TDAFW pumps
could deliver a flow of 520.5 gpm and the MDAFW pumps could deliver a flow of 547
gpm. For faulted steam generator cases this calculation predicted a maximum flow of
1038.5 gpm. This review identified the following concerns:
- The accuracy of the AFW system resistance values obtained from plant testing
did not evaluate the effect of accuracy of the instrumentation used on the
calculation results. Information provided in the calculation states: In view of the
scattered test data, it is obvious that the inaccuracy and uncertainty were
involved in pressure / flow measurements. To minimize the effect, the
19
representative orifice dP (i.e., the average dP) is used for new orifice sizing. In
addition to the data averaging, one set of data points was completely discarded
because it was not consistent with the other data without any analysis of the
reasons for the differences. Considering the fact that a typical instrument error
associated with just the flow measurement office is 2 percent and the very small
system margins, it is not clear that this analysis can demonstrate the systems
ability to satisfy the intended safety function by providing above the total
minimum flow of 510 gpm to intact steam generators while not exceeding the
value of 1050 gpm used in the containment analysis for a faulted steam
generator.
- The inspectors reviewed surveillance procedure 28210-C, Main Steam Line
Code Safety Valve Setpoint Verification, Revs. 15, 16, 17, and 18 for the steam
generator safety relief valves (SRVs) 1PSV3001, 1PSV3011, 1PSV3021,
1PSV3031, 2PSV3001, 2PSV3011, 2PSV3021, and 2PSV3031 completed from
March 2003 through March 2005. This review identified that this procedure
established the as-found acceptance criteria for these valves to be a range 1149
to 1209 psig (1185 +3%, -2%) and the as-left 1173 to 1197 psig (1185 +1%,
-1%). This allowable band for SRV settings was not applied to the analysis to
determine the worst case acceptable high and low steam generator back
pressure which could further decrease the available margin for the minimum flow
cases for the higher SRV setpoints, while decreasing the available margin for the
faulted steam generator case for the lower SRV setpoints.
- The inspectors reviewed modification MDD-91-V1M016, Auxiliary Feedwater
Rated Speed Change, dated April 4, 1991, and LDCR 2003034, Revise FSAR to
Show Values for AFW System Performance Based on Calculation X4C1302S12,
dated November 26, 2003. The review identified that in 1991, the TDAFWP
speed setting was changed from 4200 to 4230 rpm by modification
MDD-91-V1M016, which did not evaluate the effect of the speed increase on the
AFW systems margin for the faulted steam generator containment analysis. In
2003, the UFSAR was changed to reflect the results of calculation X4C1302S12
by LDCR 2003034, which did not evaluate the effects of the cumulative margin
reduction changes, lack of the instrument uncertainty considerations by the
calculations, effects of the allowable speed variations, and potential negative
effects of the allowable SRV set point acceptance criteria.
- The inspectors reviewed minor design change MDC-00-V1M036, 1A AFW Pump
Rotating Element Replacement and Thrust Bearing Change, dated September
11, 2000, and identified that when the 1A MDAFW pump rotating element was
replaced, the pump characteristics were changed. The effect of the pumps
higher TDH and flows on the AFW system margin for the faulted steam
generator ccontainment analysis was not evaluated. CR 2007105979 was
generated to address this concern.
Analysis. The inspectors determined that SNCs failure to consider the effects of
instrument accuracy, speed change, and allowable SRV set point tolerances on the
20
AFW system capability to perform the intended safety function by providing a total
minimum flow of 510 gpm to intact steam generators while not exceeding the value of
1050 gpm used in the containment analysis for a faulted steam generator was a
performance deficiency. This URI is more than minor because it is associated with the
Mitigating Systems Cornerstone attribute of Design Control. It impacts the cornerstone
objective of ensuring the availability, reliability, and operability of the AFW system to
perform its safety function in that the licensee lacked information to analyze the
capability of the AFW system to ensure the system could satisfy the intended safety
function during all conditions. The inspectors could not make a determination of impact
on operability of the AFW system due to questions concerning the validity of the system
curves and pump curves.
Enforcement. This URI was entered into the licensees corrective action program as CR
2007105979 with actions to evaluate the ability of the AFW systems ability to satisfy the
intended safety function. This issue is identified as URI 05000424, 05000425/
2007006-05, Capability of Auxiliary Feedwater System to Meet Design and Licensing
Requirements. This item is unresolved pending an NRC review of the effects of
instrument accuracy, speed change, and potential negative effects of the allowable SRV
set point acceptance criteria on the AFW systems ability to satisfy the intended safety
function. (URI 05000424,05000425/2007006-04, Capability Of Auxiliary Feedwater
System To Meet Design And Licensing Requirements)
.3 Review of Low Margin Operator Actions
a. Inspection Scope
The inspectors performed a margin assessment and detailed review of five risk
significant and time critical operator actions. Where possible, margins were determined
by the review of the assumed design basis and UFSAR response times and
performance times documented by job performance measures (JPMs). For the selected
components and operator actions, the inspectors performed an assessment of the
Emergency Operating Instructions (EOIs), Abnormal Operating Instructions (AOIs),
Alarm Response Instructions (ARIs), and other operations procedures to determine the
adequacy of the procedures and availability of equipment required to complete the
actions. Operator actions were observed on the plant simulator and during plant
walkdowns.
The following operator actions were observed on the licensees operator training
simulator:
- Recovery of offsite power, actions to recover offsite power utilizing the EDG,
RAT, and SAT. (EOP 19100-C, ECA-0.0 Loss of All AC Power, SOP 13145-
1, Diesel Generators, and SOP 13427A-1, 4160V AC Bus 1AA02 1E
Electrical Distribution System)
C, Loss of Residual Heat Removal.)
21
C, Reactor Trip or Safety Injection, and EOP 19011-C, ES-1.1 SI
Termination)
Additionally, the inspectors walked down, table-topped and reviewed the following
operational scenarios:
- Operation of Plant Wilson to provide power to safety buses. (SOP 13418-C,
Standby Auxiliary Transformer, and SOP 13419-C, Diesel Generator
Extended Allowable Outage Time (AOT))
- Local operation of the TDAFWP without AC or DC power. (SOP 13610-1,
Auxiliary Feedwater System.)
b. Findings
No findings of significance were identified.
.4 Review of Industry Operating Experience
a. Inspection Scope
The inspectors reviewed selected operating experience issues that had occurred at
domestic and foreign nuclear facilities for applicability at the Vogtle Nuclear Plant.
The inspectors performed an independent applicability review, issues that were
identified as applicable to the Vogtle Nuclear Power Plant were selected for a detailed
review. The issues that received a detailed review by the inspectors included:
- NRC Information Notice (IN), IN 86-14, Overspeed Trips of HPCI, RCIC and AFW
Turbines
- IN 95-03, Generic Letter 98-02 Wolf Creek Loss of RCS Inventory while Shutdown
- IN 2005-21,Plant Trip and Loss of Preferred AC Power from Inadequate Switchyard
Maintenance
- IN 2006-18, Significant Loss of Safety Related Electrical Power at Forsmark, Unit 1,
in Sweden
- IN 90-25, Loss of Vital AC Power with subsequent Reactor Coolant System Heat-up
- IN 2007-014, Loss of Offsite Power and Dual-Unit trip at Catawba Nuclear
Generating Station
22
b. Findings
No findings of significance were identified.
.5 Review of Permanent Plant Modifications
a. Inspection Scope
The inspectors reviewed two modifications related to the selected risk significant
components in detail to verify that the design bases, licensing bases, and performance
capability of the components have not been degraded through modifications. The
adequacy of design and post modification testing of these modifications was reviewed
by performing activities identified in IP 71111.17, Permanent Plant Modifications,
Section 02.02.a. Additionally, the inspectors reviewed the modifications in accordance
IP 71111.02, Evaluations of Changes, Tests, or Experiments, to verify the licensee
had appropriately evaluated them for 10 CFR 50.59 applicability. The following
modifications were reviewed:
- DCP 2019003401, Replace TDAFWP Turbine Speed Control Panel and associated
components
- 92-VAN0203/C929020301, Replace 1E Transformers 1AB04X, 1AB05Xm 1AB15X
b. Findings
No findings of significance were identified.
4. OTHER ACTIVITIES
4AO6 Meetings, Including Exit
Exit Meeting Summary
On May 25, 2007, the inspectors presented the inspection results to Mr. T. Tynan, Site
Vice President, and other members of the licensee staff. The inspectors returned all
proprietary information examined to the licensee. No proprietary information is
documented in the report.
On June 4, 2007, a telephone exit was conducted to disposition two items related to
the AFW system as URIs (Section 1R21.2.18 and 1R21.2.19) to Mr. J. Williams and
other members of the licensee staff.
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
M. Agcaoili, Engineer II, Engineering Support
R. Burns, Senior Engineer, Engineering Support
M. Byrd, Senior Engineer, SNC Corporate Vogtle Support
G. Coady, Engineer I, Engineering Support
J. Fridrichsen, Mechanical/Civil Supervisor, SNC Corporate Vogtle Support
W. Gover, Senior Engineer, Engineering Support
G. Gunn, Nuclear Operator Plant Instructor, Training & Emergency Preparedness
D. Hines, Senior Engineer, SNC Corporate Vogtle Support
T. Honeycutt, Senior Engineer, SNC Corporate Licensing
S. Kerstiens, Senior Engineer, Engineering Support
D. Midlik, Senior Engineer, SNC Corporate Licensing
R. Moye, Project Planning Supervisor, SNC Corporate Vogtle Support
R. Reddy, Senior Engineer, SNC Corporate Vogtle Support
M. Sharma, Nuclear Specialist I, Performance Analysis
K. Stokes, Senior Engineer, Engineering Support
S. Swanson, Engineering Support Manager, Engineering Support
H. Thompson, Assistant Team Leader, Assistant Plant Manager Support
S. Waldrup, Operations Support Superintendent, Operations
A. Wesley, Engineering Supervisor, Engineering Support
J. Williams, Site Support Manager, Assistant Plant Manager Support
NRC
G. McCoy, Senior Resident Inspector, Vogtle Nuclear Plant
M. Cain, RII, Engineering Branch 1, Chief (Acting)
ITEMS OPENED, CLOSED, AND DISCUSSED
Open/Closed
05000424,05000425/2007006-01 NCV Safety-Related 480 VAC Molded Case Circuit
Breakers On MCC 1ABF, 1BBF, 2ABF, and
2BBF Not Tested. (Section 1R21.2.9)
05000424,05000425/2007006-02 NCV Maximum Differential Pressure for Containment
Emergency Sump Isolation Valves Not
Calculated. (Section 1R21.2.17)
05000424,05000425/2007006-03 URI Ability to Satisfy TS SR 3.7.5.2. (Section
1R21.2.18)
Attachment
2
05000424,05000425/2007006-04 URI Capability of Auxiliary Feedwater System to Meet
Design and Licensing Requirements. (Section
1R21.2.19)
Attachment
3
DOCUMENTS REVIEWED
Calculations
1XCF02, 1E Battery Sizing, Rev. 16
MX3CA27, Degraded Grid/Undervoltage Relay Setting for Unit 1 and Unit 2, Rev. A01
X3CA03, Reserve Auxiliary Transformer Sizing, Rev. 5
X3CA18, Vogtle Unit 1 Load Study, Rev. 7
X3CA26, Unit 1 and Unit 2 Relaying, Rev. 6
X3CF07, DC Breaker Sizing, Rev. 14
X3CH01, 120 Vac Vital Bus Loading and Cable Sizing, Rev. 6
X3CK03-C, Control Cable Sizing Details for Cutler-Hammer Freedom Series Starters, Rev. 2
X3CT08,Fire Event Safe Shutdown Circuit Analysis, Rev. 15.
X4C1000U01, Differential Pressure Calculations, Rev. 15
X4C1202S16, NSCW Flows & Pressures Post-Accident Conditions, Rev. 1
X4C1202V11, NSCW Sys-P/T & Flows During Normal Operation, Rev. 5
X4C1302S12, Auxiliary Feedwater Pump Discharge Line Orifice Sizing, Rev. 1
X4C1302S08, FSAR Auxiliary Feedwater System, Rev. 5
X4C1302T01, Condensate Storage Tanks, Rev. A
X4C1302V04, Auxiliary Feedwater Pumps Technical Specifications Verification, Rev. 5
X4C1302V06, Condensate Storage Tank Verification, Rev. 1
X4C2107V01, Diesel Generator Building - Normal and Emergency Operation, Rev. 3
X4C2107V03, Diesel Generator Building Minimum Ventilation Requirements, Rev.1
X4C2403V02, Verification of Diesel Fuel Oil Transfer, Rev.0
X5CP1668, NSCW Cooling Tower Spray Header Shutoff and Bypass Valve Controls Train A
and Train B, Rev. 6
Operating Procedures
13011-1, Residual Heat Removal System, Rev. 65
13145-1, Diesel Generators, Rev. 65
13415-1, Reserve Auxiliary Transformers, Rev. 12.4
13419-C, Diesel Generator Extended AOT, Rev. 6.2
13427A-1, 4150V AC Bus 1AA02 1E Electrical Distribution System, Rev. 3.1
13431-1, 120V AC 1E Vital Instrument Distribution System, Rev. 8
13481-C, Standby Auxiliary Transformer, Rev. 15
13610-1, Auxiliary Feedwater System, Rev. 41
18019-C, Loss of Residual Heat Removal, Rev. 25
18034-1, Loss of Class 1E 125V DC Power, Rev. 8
19000-C, E-0 Reactor Trip or Safety Injection, Rev. 32
19011-C, ES-1.1 SI Termination, Rev. 26.1
19100-C, ECA-0.0 Loss of All AC Power, Rev. 31
ABW 2901, Plant Wilson Blackstart Procedure, Rev. 16
13415-1, Reserve Auxiliary Transformers, Rev. 12.4
13418-C, Standby Auxiliary Transformer, Rev. 15
17033-2, Annunciator Response Procedures for ALB33 on EAB Panel, Rev. 17
17036-1, Annunciator Response Procedures for ALB36 on EAB Panel, Rev. 17
Attachment
4
NMP-AD-006, Frequently Performed Test and Evolutions, Rev.2
00656-C, Vehicle Control, Rev. 6
Operations Training Related Documents
RQ-JP-18034-001, Locally Remove Diesel Generator From Service, Rev. 9
RQ-JP-18032-001, Respond to Loss of 120V AC Instrument Power 1AY1A (RO Actions),
Rev. 16
RQ-JP-18032-002, Respond to Loss of 120V AC Instrument Power (BOP Actions), Rev. 15
RQ-JP-13427-001, Return ESF Bus From Diesel Generator to Normal Supply, Rev. 17
RQ-JP-13418-002, Energize the Standby Auxiliary Transformer (SAT), Rev. 6
RQ-JP-13418-003, Aligning the SAT to Class 1E 4160VAC Bus, Rev. 9
RQ-JP-18034-002, Turbine Driven AFW Pump Local Manual Control Without DC Power, Rev. 1
RQ-JP-13610-001, Reset of the TDAFW Pump Trip and Throttle Valve, Rev. 6
RQ-JP-13610-002, Locally Operate TDAFW Pump Using HV-5106, Rev. 4
RQ-JP-13610-003, Locally Operate TDAFW Pump Using HV-5106 With Inability to
Maintain Speed Control, Rev. 4
V-LO-TX-16001, Primary Systems, Rev. 3.0
V-RQ-SE-07106, Inadvertent Safety Injection, Rev. 0
Test Procedures
14805-2, Residual Heat Removal and Check Valve IST and Response Time Tests, Rev. 31
14807-1, Motor Driven Auxiliary Feedwater Pump and Check Valve In-Service and Response
Time Test, Rev. 29.2
14807-2, Motor Driven Auxiliary Feedwater Pump and Check Valve In-Service and Response
Time Test, Rev. 24.1
14810-1, TDAFW Pump Operability, Response Time and Check Valve IST, Rev. 36.2
14810-2, TDAFW Pump Operability, Response Time and Check Valve IST, Rev. 30.2
22505-C, Switchgear Panel Voltmeter Calibration, 09/30/06, 09/29/06
23202-C, G.E. IAC Overcurrent Relay Calibration, Rev. 19, for DG1A, performed 1/22/07
23202-C, G.E. IAC Overcurrent Relay Calibration, Rev. 18, for 1NXRA, performed 3/26/05
23222-C, General Electric Type IAV53K Over and Under Voltage Relay Calibration, Rev. 14
03/1605
23226-C, G.E. Type ICW51 Power Relay Calibration, Rev. 6, for DG1A, performed 1/21/07
23230-C, G.E. Model INC77B Negative-Phase Sequence Time Overcurrent Relay calibration,
Rev. 7.1, for DG1A, performed 1/24/07
23232-C, G.E. Type IJCV51A and IJCV51B TOC with Voltage Restraint Calibration, Rev. 5,
performed 1/21/07
23244-C, G.E. Type IJD Percentage Differential Relay Calibration, Rev. 8, for DG1A, performed
1/24/07
23250-C, Westinghouse Type HU & HU-4 Differential Relay Calibration, Rev. 14, for 1NXRA,
performed 3/25/05
23278-C, Westinghouse Type KF Underfrequency Relay Calibration, Rev. 8 for DG1A,
completed 1/24/07
24613-1, Safety Features Sequencer Train A Channel Operational Test and Channel
Calibration, Rev. 27.3, performed 10/13/06
Attachment
5
25506-C, Motor Control Center Preventive Maintenance, Rev. 16, performed 10/13/00
25506-C, Motor Control Center Preventive Maintenance, Rev. 28, performed 31/5/05
27710-C, 125 VDC Circuit Breaker Inspection and Testing, Rev. 38, for 1CD101, performed
3/18/05
27710-C, 125 VDC Circuit Breaker Inspection and Testing, Rev. 39, for 1CD111, performed
4/8/05
27731-C, 480 Volt Switchgear Cubicle/Transformer Maintenance, Rev. 26, 03/16/05
27828-C, SCI Non-1E 7.5KVA Inverter Calibration, 10/04/06, 10/01/06
28210-C, Main Steamline Code Safety Valve Setpoint Verification, Rev. 15, 16, 17, and 18
28705-C, 4.16KV/13.8KV Circuit Breaker Inspection and Testing, Rev. 23, for 1AA0221,
performed 3/23/05
28705-C, 4.16KV/13.8KV Circuit Breaker Inspection and Testing, Rev. 22.1, for Incoming
1AA02 Feeder Breaker, performed 1/24/05
28707-C 480V Air Circuit Breaker Maintenance and 60 Month Check, Rev. 27, performed
3/17/05
28816-C, Class 1E Battery Modified Performance Test, performed 9/27/06
54727-1, Device 451NRA1, RAT 1NXRA Neutral Backup Ground Relay Functional Test, Rev 2,
performed 3/22/05
55026-C, Calculation of Breaker Trip Settings and Trip Test Data for 125V DC Switchgear
Breakers, Rev. 8, for breakers 1CD101 and 1CD111, performed 11/13/88 and 3/5/99
ABW 2902, Plant Wilson 18 Month Blackstart Test Procedure, Rev. 7, performed 4/15/06
ABW 2902, Plant Wilson 18 Month Blackstart Test Procedure, Rev. 7
Design Changes/Modifications
92-VAN0203, Replace 1E Transformers 1AB04X, 1AB05Xm 1AB15X
C929020301, Replace 1E Transformers 1AB04X, 1AB05Xm 1AB15X, 07/28/93
DCP 2019003401, Replacement of TDAFW Pump Turbine Speed Control Panel and
Associated Components, Version 1.0
MDD-91-V1M016, Auxiliary Feedwater Rated Speed Change, April 4, 1991
MDC-00-V1M036, 1A AFW Pump Rotating Element Replacement and Thrust Bearing Change,
September 11, 2000
Design Basis Documents
DC-1202, Nuclear Services Cooling Water System, Rev. 12
DC-1202-A, Nuclear Safety Cooling Tower, Rev. 11
DC-1217, Auxiliary Component Cooling Water System, Rev. 3
DC-1302, Auxiliary Feedwater Water System, Rev. 14
DC-1566, Diesel Generator Building HVAC System, Rev. 4
DC-18000E, General Design Criteria, Rev. 15
DC-1801 Offsite Power System, Rev. 7
DC-1804, 4160 VAC System, Rev. 10
DC-1805, 480 VAC System, Rev. 12
DC-1806, Class 1E dc System, Rev. 11
DC-1807, 120 VAC Power System, Rev. 9
DC-1809, Cable System, Rev. 19
Attachment
6
DC-1810, Raceway System, Rev. 14
DC-1821, Standby Power System, Rev. 12
DC-2402, Emergency Diesel Generator Systems, Rev. 9
DC-2403, Emergency Diesel Generator, Rev. 9
Condition Reports (CRs)
2005100563, 1ABF Material condition is degrading
2005101944, Control Room received 1ND3A 125VDC Ground Annunciator
2005102348, Discharge Test of battery 1Nd1B
2005102356, Failure of the K1 relay to reset, maintains diesel generator field shorted
2006110873, RAT 1B (1NXRB) Gassing trend
2006111881, Battery Charger 1AD1CA ac input breaker tripped
2006112084, Battery Charger 1AD1CA ac input breaker tripped
Work Orders
10202438, PM REG XFMR 1BBB40RX
10203021, Inspect, Clean, and Measure Grounding Resistors (RAT), performed 9/29/05
10203022, Inspect, Clean, and Measure Grounding Resistors (RAT), performed 10/4/03
10203252, Implement 480V MCC breaker cubicle 1ABF05 changes, performed 2/4/03
10203253, 480 V MCC breaker cubicle 1ABF14 phase B failed trip test, performed 1/31/03
10203410, PM REG XFMR 1ABB40RX
1030093101, Functional Test of Transformer 1AB05X
1053523801, PM Essent AC Inverter 1ND3I2
1053523701, PM Essent AC Inverter 1ND3I3
106045201, PM Sequencer Board Train A
Drawings
1X3D-AA-A01A, Main One Line Unit 1, Rev. 25
1X3D-AA-B02A, One Line Diagram Relays & Meters, RAT, Rev. 7
1X3D-AA-C01A, One Line Diagram 4160V/13.8KV Switchgear ANA04/ANAA, Rev. 0
1X3D-AA-C01A, One Line Diagram 13800V Switchgear 1NAA, Rev. 20
1X3D-AA-D02A, One Line Diagram 4160V Switchgear 1AA02, Rev. 12
1X3D-AA-D02B, One Line Diagram 4160V Switchgear 1AA02, Rev. 8
1X3D-AA-F16A, One Line Diagram 480V Motor Control Center 1ABb 1-1805-S3-ABB, Rev.18
1X3D-AA-F36A, One Line Diagram 480V Motor Control Center 1ABF, Rev. 15
1X3D-AA-G01A, Main One Line Class 1E 125V DC and 120V Vital AC Systems, Rev. 10
1X3D-AA-H04A, One Line Diagram 125V DC Class 1E Distr. Train C, Rev. 23
1X3D-AA-M03A, UNIT 1 Power Transformer Data &Tap Settings, Rev. 0
1X3D-AA-M04A, UNIT 1 Load Center Transformer Data and Tap Settings, Rev.5
AX3D-AA-M06A, Unit A Standby Auxiliary Transformer Data and Tap Settings, Rev. 1
1X3D-AA-M08A-13, UNIT 1 Relaying Data, Rev. 1
1X3D-AA-M08A-14, UNIT 1 Relaying Data, Rev. 1
1X3D-AA-M08A-15, UNIT 1 & 2 Relaying Data, Rev. 1
1X3D-AA-M08A-16, UNIT 1 & 2 Relaying Data, Rev. 2
Attachment
7
1X3D-AA-M08A-17, UNIT 1 & 2 Relaying Data, Rev. 0
1X3D-AA-M08A-28, UNIT 1 Relaying Data, Rev. 1
1X3D-AA-MO8A-32, UNIT 1Relaying Data, Rev. 1
1X3D-AA-M08A-35, UNIT 1 Relaying Data, Rev. 1
1X3D-AA-M08A-36, UNIT 1 Relaying Data, Rev. 2
1X3D-AA-M08A-37, UNIT 1 Relaying Data, Rev. 1
1X3D-AA-M08A-56, UNIT 1 Relaying Data, Rev. 3
1X3D-AA-K01A, Diesel Generator 1A & 1B Relays & Meters, Rev. 5
1X3D-BA-D02B, E/D 4160V SWGR 1AA02 1NCM BRKR 1NXRA, Rev.12
1X3D-BA-D02D, E/D 4160V 1NCM BRKR 152-1AA0219, from EDG 1A, Rev. 11
1X3D-BA-D02L, E/D 4160V SPLY FOR BRKR TO XFMR 1AB05X, Rev. 4
1X3D-BB-B01P, Elementary Diagram Electrical System, Generator Tripping, Rev. 7
1X3D-BB-B01V, Elementary Diagram Electrical System, Generator Tripping, Rev. 6
1X3D-BB-B01R, Elementary Diagram Electrical System, Generator Tripping, Rev. 2
1X3D-BC-F04A, Elementary Diagram Aux. Feedwater System 1-1302-P4-003-M01, Rev. 11
1X3D-BC-Q04B, Elementary Diagram Main Steam System Safety Actuation Signal, Rev. 11
1X3D-BD-K05U, Elementary Diagram Nuclear Service Water System 1HV-1668A, Rev.14
1X3D-BD-K05W, Elementary Diagram Nuclear Service Water System 1HV-1669A, Rev.11
1X3D-BD-L01E, Elementary Diagram Component Cooling Water System 1-1203-P4-005-M01,
Rev. 10
1X3D-BG-G03A, Elementary Diagram Aux. FDW Pump House HTG & Vent System 1-1593-B7-
001-M01, Rev. 8
1X3D-BG-F01M, Elementary Diagram DSL Gen Building HVAC System 1HV-12050, 12051,
12053, 12054, Rev.4
1X3D-BH-G03C, Elementary Diagram Diesel Engine Control DGA1, Rev. 7
1X3D-BH-G03D, Elementary Diagram Diesel Engine Control DGA1, Rev. 8
1X3E13-00005-2, Diesel Generator SFS Load Sequencer Logic Diagram, Rev. 2
1X4AH04-2-8, Component Cooling Water Surge Tank & Auxiliary Component Cooling Water
Surge Tank, Dated 1/30/80
1X4DB122, Residual Heat Removal System No. 1205, Ver. 49.0
1X4DB133-1, Nuclear Services Cooling Water System, System No. 1202, Ver. 49.0
1X4DB133-2, Nuclear Services Cooling Water System, System No. 1202, Ver. 55.0
1X4DB134, Nuclear Services Cooling Water System, System No. 1202, Ver. 29.0
1X4DB135-1, Nuclear Services Cooling Water System, System No. 1202, Ver. 29.0
1X4D 135-2, Nuclear Services Cooling Water System, System No. 1202, Ver. 34.0
1X4DB136, Component Cooling Water System, System No. 1203, Ver. 32.0
1X4DB137, Component Cooling Water System, System No. 1203, Ver. 18.0
1X4DB138-1, Auxiliary Component Cooling Water System, System No. 1217, Ver. 30.0
1X4DB138-2, Auxiliary Component Cooling Water System, System No. 1217, Ver. 19.0
1X4DB139, Auxiliary Component Cooling Water System, System No. 1217, Ver. 29.0
1X4DB161-1, Auxiliary Feedwater Water System, Condensate Storage & Degasifier System,
System No. 1302, Ver. 44.0
1X4DB161-2, Auxiliary Feedwater Water System, System No. 1302, Ver. 27.0
1X4DB161-3, Auxiliary Feedwater Water System (Aux Feedwater Pump Turbine Driver), System
No. 1302, Ver. 40.0
1X4DB170-1, Diesel Generator System, Train A, System No. 2403, Ver. 43.0
1X4DB217-1, Diesel Generator Building HVAC, System No. 1566, Ver. 16.0
Attachment
8
1X5DN089-1, Control Logic Diagram, Nuclear Services Cooling Tower System, Rev. 7
1X5DN089-2, Control Logic Diagram, Nuclear Services Cooling Tower System, Rev. 4
1X5DN089-3, Control Logic Diagram, Nuclear Services Cooling Tower System, Rev. 10
1X5DT0026, Level Setting Diagram, ACCW Comp. CW Surge Tk., Rev. 4
AX3D-AA-A03A, Vogtle-Wilson Main One Line Diagram, Rev. 4
Miscellaneous Documents
AX3AC02-01668-6, Low Voltage (480V Load Center) Power Circuit Breaker Instruction Manual,
11/08/99
AX3AC03-00920, Instruction Manual & Parts List for 480Vac MCC, Rev. 11
AX3AD01-00025, Instruction Manual, C&D Batteries, Rev. 5
AX3AD01-00084, Instruction Manual, AutoReg Chargers, Rev. 3
AX3AD02-00120, Instruction Manual, Cutler-Hammer Motor Control Centers, Rev. 4
AX3AD03-05020, Instruction Manual, Brown Boveri 125 VDC Switchgear, Rev. 11
AX3AE13-00001, Safety Features Sequencer Functional Requirements, Rev. 4
AX3AQ03A-00027, 10KVA Inverter Instruction and Operating Manual, Rev.1
DCP 02-VAN0017, Inverters BD1I12 and 1AD1I1 Functional Test
DG1A A.C. Synchronous Generator Data, dated 6/1/78
DOEJ-SJ-C070401401-001, Channel Uncertainty - AFW Pump Suction and Discharge
Pressures - P-5110, P-5107
DOEJ-SM-C070401401-001, Maximum Opening Differential Pressure for the Vogtle RHR Sump
Valves (1/2HV8811A and B) at Steady State Voltage of 88.5 percent
General Electric Application Guide, A-C Ground Indication, SA-75, 12/20/1974
Health Report, 125 Volt Direct Current System, 1st QTR 2007
Health Report, 4160 Volt Alternating Current System, 1st QTR 2007
Health Report, 480 Volt Alternating Current System, Reporting Period 4th QTR 2005-4th QTR
2006
Health Report, Emergency Diesel Generator Systems, 1st QTR 2007
Health Report, High Voltage Switchyard, 1st QTR 2007
LDCR 2003034, Revise FSAR to Show Values for AFW System Performance Based on
Calculation X4C1302S12, November 26, 2003
Letter from Southern Company Services to Vogtle Project Nuclear Operations dated
July 6, 2000
Letter from J.G. Aufdenkampe to S.C. Swanson dated May 11, 2007
Operating Experience Program Evaluation for NRC IN 95-03, Loss of RCS Inventory and
Potential Loss of Emergency Mitigation Functions While in a Shutdown Condition
Procedure 50028-C Engineering Maintenance Rule, Performance Monitoring and Evaluation for
Offsite Power and High-Voltage Switchyard, Reporting Period 1/1/2005 through 3/1/2007
RM-0287, Power Circuit Breakers Type AKR, Rev. 1
Summary of VDGP Response to Supplement 2 to NRC Information Notice 86-014, Overspeed
Trips of AFW, HPCI and RCIC Turbines
S&C Data Bulletin 711-90, Circuit -Switchers-Mark V, 6/3/1991
S&C Instruction Sheet 711-600, Circuit-Switchers- Mark V, 6/26/1982
S&C Specification Bulletin 711-31, Circuit-Switchers- Mark V, 2/520/07
Summary of the Plant Wilson Engine Generator Battery Capacity Test, performed 11/29/2000
Voltage Profile, 4160 Volt Bus 1AA02, for the period 5/14/06 through 5/1420/07
Attachment
9
Voltage Profile, 125 Volt Battery 1CD1, for the period 5/14/06 through 5/1420/07
X7GH14-V500, Vogtle Plant Response to IN 07-14, Dual unit Loss of Offsite Power
Corrective Action documents initiated due to CDBI activity:
CR 2007100218, Cable size on the low side of the standby auxiliary transformer (13.8KV/4160V)
CR 2007100247, Corporate Level CR to Address that Following RHR Pump Test the Actual dP
is Greater than the Assumed dP, May 17, 2007
CR 2007100263, Address the apparent absence of an NPSHa calculation for the AFW system
pumps.
CR 2007104981, Evaluation not performed for toolbox chained to the control panel in the
TDAFWP room.
CR 2007104983, Eyewash stations and metal racks not secured in the EDG buildings.
CR 2007105409, Transformer 1AB05X impedance value shown on drawing does not match.
CR 2007105413, NPSH calculation for the AFW pumps is missing
CR 2007105428, IPC point jumps in indicated temperature and remains high.
CR 2007105436, NRC questioned method of satisfying the TDAFW pump surveillance
requirement.
CR 2007105463, Generate work orders to replace the neutral grounding resisitors on UAT and
RAT.
CR 2007105469, Pneumatic tubing on EDG building 1A HVAC recirculation damper is not
properly installed.
CR 2007105563, Rust around operating levers on breakers in panel MCC 1ABF.
CR 2007105624, Reference files are support documentation and cannot be found.
CR 2007105713, Incorporation of Instrument Uncertainty and Density Corrections Into Technical
Specifications Surveillance Acceptance Criteria.
CR 2007105769, Routine maintenance such as continuous current and interrupting rating of 230
KV circuit switchers not performed.
CR 2007105779, Review of the need for routine maintenance of safety related level switches.
CR 2007105833, Review to confirm ACCW surge tank safety related level switched should be
listed as safe shutdown component.
CR 2007105848, Pressure increase after RHR pump testing.
CR 2007105855, Walkdown MCC 1ABF.
CR 2007105859, Walkdown MCC 1ABF.
CR 2007105870, Potential exists for diesel generator frequencies to vary.
CR 2007105895, Initial Speed for TDAFW Pump Recorded at 4200 rpm.
CR 2007105905, TDAFWP has limited flow margin in some analyzed accidents.
CR 2007105979, Replacement of MDAFW Rotating Element Was Not Completely Evaluated
Against the MSLB Design Basis, May 25, 2007
Attachment