ML20245F423
| ML20245F423 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 06/15/1989 |
| From: | Aiello R, Patterson C, Rogge J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20245F400 | List: |
| References | |
| 50-424-89-14, 50-425-89-15, NUDOCS 8906280135 | |
| Download: ML20245F423 (37) | |
See also: IR 05000424/1989014
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d 1 UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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o REGION 11
"g 101 MARIETTA ST., N.W. .
.'..,, ATLANTA, GEORGIA 30323
Report Nos.:- . 50-424/89-14 and 50-425/89-15
Licensee:. Georgia Power Company
P.O. Box 1295
Birmingham, AL 35201
Docket Nos.: 50-424 and 50-425 License Nos.: NPF-68 and NPF-81
Facility Name: Vogtle 1 and 2
Inspection Conducted: March 18 - May 5, 1989
Inspectors: ,
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J. Edogge Senior Resident Inspector Date Signed
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R. K Aiello, Resident Inspector
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Date Signed ;
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C.-AT Patterson, Project Ehgineer (April 3-6) Date Signed
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J. ,ErMenning, Hatch Sent'or Resident (April 1-2)
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Date Signed
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R.-t'. Prevatte, Summer Senior Resident (April 1-2)
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Date Signed
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PT C At5pkins, Summer Resident <1 April 1-2) Date Signed
Accompanied By: Rick Mc hort r (March 27-30) i
Approved By: [ <<LIuAn-
M. V,'/Sinkule, Secti6n Chief
['"'/6T
Date Signed
Divistion of Reactor Projects
SUMMARY )
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Scope:
This routine inspection entailed resident inspection in the following areas: a
plant operations, radiological controls / chemistry, maintenance, surveillance,
security, startup testing (Unit 2), engineering technical support, and quality
programs and ada.L istrative controls affecting quality.
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Results:
In. the areas inspected, fourteen violations were identified. Of these, one
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violation was cited, and thirteen violations were non-cited pursuant to the
discretionary provisions of the NRC Enforcement Policy. The cited violation
was identified in the area of operations, and it involved six examples of
failure to establish or implement procedures. One of the six examples
pertained to Unit 1 only (paragraph 5.f), three pertained to Unit 2 only
(paragraphs 4.b(3)(q), 4.b(3)(r), and 4.b(3)(s)), and two pertain to both
units (paragraphs 2.b(1) and 3). Of the thirteen non-cited violations, five
pertained to Unit 1: one in the area of radiological controls / chemistry
(paragraph 4.b(2)(d)), two in the area of surveillance (paragraphs 4.b(2)(a)
and 4.b(2)(b)), and two in the area of emergency technical support (para-
graphs 4.b(2)(c) and 4.b(3)(h)). The remaining eight non-cited violations
pertainer! to Unit 2. Three were identified in the area of plant operations
(paragraphs 4.b(2)(f), 4.b(3)(m), and 4.b(3)(p)), three were identified in the
area of radiological controls / chemistry (paragraphs 4.b(2)(h), 4.b(2)(i), and
4.b(2)(j)), one was identified in the area of maintenance (paragraph 4.b(2)
(k)), and one was identified in the area of engineering technical support
(paragraph 4.b(2)(e)).
Two inspector followup items were also identified involving the adjustment of
the P-9 setpoint when steam dumps are removed from service (paragraph 3) and
the resolution of restoring the safety system monitor panel to a condition to
correctly indicate the operability status (paragraph 5.d).
Two strengths and one weakness was noted within the report. The areas of
maintenance and startup testing (Unit 2) were noted as strengths with the area
of operations noted as a weakness.
- Maintenance (paragraph 2.b(7)) was considered a strength primarily due to
the planning and execution of the work schedule. Short system outages on
Unit 1 and short plant outages on Unit 2 were effectively conducted. Most
noteworthy was the elimination of a 10-day scheduled outage during the
Unit 2 test program due to this proficiency.
- Startup Testing on Unit 2 (paragraph 3) was a second strength even though
one procedure error resulted in a preventable transient. The transient
was preventable because the identical error was identified during Unit I
test program. More significant was the proficient and efficient conduct
of the Remote Shutdown Test and the Loss of Offsite Power Test.
- Operations evidenced weakness in the area of procedure establishment and
implementation of the basic operating procedure 12004-C " Power Operation."
Examples included in the cited violation are failure to open bypass ,
isolation valves (paragraph 3), to secure from long-cycle cleanup !
(paragraphs 3 and 4.b(3)(s)), and to perform the transfer from auxiliary
Other operations errors were noted in
to
themain
LERsfeedwater
(paragraphs(parag(raph
4.b 2) and 3).
4.b(3)). This concern has been verbally
expressed to licenset management.
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. REPORT DETAILS
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1. . Persons Contacted
Licensee Employees
- G. Bockhold, Jr., General Manager Nuclear Plant .
- A. L. Mosbaugh, Plant Support Manager
- R. M. Odom, Nuclear Safety & Compliance Manager / Plant Engineering
Supervisor
- J. E. Swartzwel_ der, Manager Operations
W. F. Kitchens, Assistant General Manager Plant Operations
R. L. Legrand, Manager Chemistry and Health Physics
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- H. M. Handfinger, Manager Maintenance
G. A. McCarley,-ISEG Supervisor
- G. R. Frederick, SAER Supervisor
W. E. Mundy, Quality Assurance Audit Supervisor
C. L._Coursey, Maintenance Superintendent
Other licensee employees . contacted - included craftsmen, technicians,-
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supervision, engineers, operations, maintenance, chemistry,- quality
control inspectors, and office personnel.
- Attended Exit Interview
An alphabetical list of acronyms and initialisms used throughout this -
report are listed in the last paragraph.
2 .- Operational' Safety Verification - (71707)(93702)(71715)
Unit 1 operated this inspection period _ in _ Power Operations (Mode 1) at
100% reactor power.
Unit 2 began this inspection period in Mode ^ 4 (Hot shutdown). On-
March 18, 1989, Unit 2 entered into Mode 3 (Hot Standby). Later that'same
day.(night shift), Unit 2 experienced an inadvertent SI due to personnel
error followed by an. NUE declaration. On March 19, following the SI,
Unit 2 experienced a CVI due to 2RE-2565 radiation monitor. Additionally,
on March 19, (night shift), Unit 2 expei enced a FWI due to P-14 on SG #4
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caused by personnel error. On March 28, , nit 2 entered Mode 2 (Startup),.
went critical, and commenced low power r'aysics testing. On April 5, MFP
"A" tripped .resulting in a MDAFW pump actuation. On April 7, Unit 2
entered Mode 1. Later that same day, a FWI occurred as a result of a Hi Hi
SG 1evel. The unit later entered Mode 2. The unit reentered Mode 1 on
= April 8. On April 9, MFP "A" tripped resulting in a MDAFW pump actuation
and subsequent Mode 2 entry. The unit reentered Mode 1 on April 10. The
-main turbine was tied to the grid on April 11. Later that same day, the
reactor was tripped from the remote shutdown panel and placed in Mode 3 as
part of a required test. While recovering, the unit received an AFW
actuation during transfer of controls from remote shutdown panel with both
MFW pumps tripped. On April 12, the unit entered Mode 2 and went critical
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with subsequent entry into Mode 1. On April 14, the unit conducted a LOSP
test with subsequent entry.into Mode 3. Following the LOSP test, the. unit
went into a three day maintenance outage. On April 15, the unit entered
Mode 2, went critical, and entered Mode 1. On April 16, the unit was tied
to the grid. On April 18, the main turbine was removed from the grid and-
tripped to conduct secondary system repairs. On April 19, the main
turbine was returned to service and tied to the grid. On April 22, a unit
turbine trip occurred due to a loss of stator cooling. This was followed
by a FWI on SG #3 Hi Hi level and subsequent AFW start. The unit then
entered Mode 2. Later the same day, the unit reentered Mode 1. On
April 23, the main generator was tied to the grid. On April 24 with all
of the 30% plateau testing complete, the unit commenced power ascension to
50% for 50% power plateau testing. On May 2, the unit was increasing
power to 75%. for 75% power plateau testing when a reactor trip occurred
with the plant at 63% from a turbine trip following a test of the
electrical overspeed trip circuit. On May 3, the unit reentered Mode 2
achieved Mode 1, and was operating at 75% at the end of the inspection
period.
a. Control Room Activities
Control Room tours and observations were performed to verify that
facility operations were being safely conducted within regulatory
requirements. These inspections consisted of one or more of the
following attributes as appropriate at the time of the inspection.
- Proper Control Room staffing
- Control Room access and operator behavior
- Adherence to approved procedures for activities in progress
- Adherence to technical specification limiting conditions for
operation
- Observance of instruments and recorder traces of safety-related
and important-to-safety systems for abnormalities
- Review of annunciators alarmed and action in progress to correct
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Control Board walkdowns
- Safety parameter display and the plant safety monitoring system
operability status
- Discussions and interviews with the On-Shift Operations
Supervisor, Shift Supervisor, Reactor Operators, and the Shift
Technical Advisor (when stationed) to determine the plant
status, plans, and to assess operator knowledge
- Review of the operator logs, unit logs and shift turnover sheets
No violations or deviations were identified.
b. Facility Activities
facility tours and observations were performed to assess the
effectiveness of the administrative controls established by direct
observation of plant activities, interviews and discussions with
licensee personnel, independent verification of safety system status
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and LCOs, licensee meetings and facility records. During these
inspections, the following abjectives were achieved:
~(1) Safety System Status (71710) - Confirmation of system
operability was -obtained by verification that flowpath valve
alignment, control and power supply alignments, component j
conditions, and support systems for the accessible portions of
the ESF trains were proper. The inaccessible portions are
confirmed as availability permits. An additional indepth
inspection of the Unit 1 SI system was performed to review the
system lineup- procedure with the plant drawings and as-built
configurations and to compare valve remote and local
indications. Walkdowns were expanded to include hangers and
supports and electrical equipment interiors. The inspector
observed that the lineup was not in accordance with license
requirements in that the SI RCDT pump discharge to RWST
isolation (1-1204-U4-002), SI RWST INL FI-0928A and FI-0928B
isolation valves were found open. DCs were properly issued by
the SS to correct these deficiencies. These valve misalignments
did not render the SI system inoperable. Several valves were
noted to have missing label plates. Rooms A9 and A10 need a
great deal of attention from a Health Physics and cleanliness
point of view.
The licensee's program for maintaining control room drawings was
reviewed. On April 28 and May 4,1989, the unit control rooms
and TSC drawings were inspected. This inspection included a
detailed walkdown of the SI system (discussed above) and a
review Of the following drawings to determine legibility,
current revision verification and verification that procedure
valve lineups were appropriate:
1X4DB119 Rev 20 1X4DB130 Rev 22 1X4DB129 Rev 23
1X4DB133-1 Rev 23 1X4DB136 Rev 22 1X4DB161-1 Rev 22
1X4DB170-1 Rev 23 1X4DB120 Rev 14 1X4DB138-2 Rev 15
1X4DB136 Rev 22 1X4DB121 Rev 24 1X4DB131 Rev 19
1X4DB139 Rev 18 IX4DB138-1 Rev 16 1X4DB122 Rev 26
1X4DB132 Rev 14 1X4DB133-2 Rev 26 IX4DB135-1 Rev 21
1X4DB137 Rev 15 IX4DB161-2 Rev 22 IX4DB161-3 Rev 20
1X4DB170-2 Rev 22 1X4DB116-2 Rev 15 1X4DB117 Rev 18
IX4DB118 Rev 20 CX40B173-557 Rev 1 CX4DB173-558 Rev 1
CX4DB173-553 Rev 1 i
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The inspector determined that the procedures for controlling the
distribution of drawings were satisfactory. The drawings
adequately represent the plar,t's current configuration. Three
drawings IX4DB133-1 Rev 23, 1X4DB122 Rev 24, and IX4DB122
Rev. 26, (NSCW, SI, and RHR respectively) are too congested and
therefore, difficult to read. It was also determined that most
of the safety-related drawing ABNs were not legible. Three in
particular which are examples of the worst case are 1X4DB161
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t Rev. 22,1X40B121 Rev. 24, and 1X4DB122 Rev. 26 (AFW, SI, and
RHR respectively). Administrative procedure 00101-C, " Drawing
Control," Step 3.4.4, requires that drawing legibility be
ensured prior to distribution and engineering procedure 50009-C,
"As-Built Notices," Step 4.6.3, requires ABNs to be legible and
-reproducible. This constitutes a violation of administrative
procedure 00101-C and engineering procedure 50009-C.
This violation is one example of violation 50-424/89-1a-01 and
50-425/89-15-01, " Failure To Implement Procedures 00101-C and
50009-C Resulting In TS 6.7.1.a Violation."
(2) Plant Housekeeping Conditions - Storage of material and
components and cleanliness conditions of various areas
throughout the facility were observed to determine whether
safety and/or fire hazards existed.
(3) Fire Protection - Fire protection activities, staffing, and
equipment were observed to verify that fire brigade staffing wa.c
appropriate and that fire alarms, extinguishing equipment,
actuating controls, fire t ihting equipment, emergency
equipment, and fire barriers were operable.
(4) Radiation Frotection - Radiation protection activities,
staffing, and equipment were observed to verify proper program
implementation. The inspection included review of the plant
program effectiveness. Radiation work permits and personnel
compliance were reviewed during the daily plant tours.
Radiation Control Areas were observed to verify proper
identification and implementation.
(5) Security - Security controls were observed to verify that
security barriers were intact, guard forces were on duty, and
access to the Protected Area was controlled in accordance with
the facility security plan. Personnel were observed to verify
proper display of badges and that personnel requiring escort
were properly escorted. Personnel within Vital Areas were
observed to ensure proper authorization for the area. Equipment
operability or proper compensatory activities were verified on a
periodic basis.
(6) Surveillance (61726)(61700) - Surveillance tes#s were observed
to verify that approved procedures were being ; sed, qualified
personnel were conducting the tests, tests were adequate to
verify equipment operability, calibrated equipment was utilized,
and TS requirements were followed. The inspectors observed
portions of the following surveillance and reviewed completed
data against acceptance criteria:
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Surveillance No. Title
14000-1 Rev. 17 Operations Shift And Daily Surveillance.
Logs
14000-2 Rev. 2 Operations Shift And Daily Surveillance
Logs
14220-1 Rev. 3 Main Turbine Valves Weekly Stroke Test
14228-: Rev. 1 Operations Monthly Surveillance Logs
14230-1 Rev. 4 Weekly Train A & B Verification Offsite
To Onsite Class 1E A.C. Distribution
System Circuit Breaker Alignments While
In Modes 1-4
14235-2 Rev. 1 Onsite Power Distribution Operability
Verification
14450-2 Rev. 1 RCS Pressure Isolation Valve Leakage
Test
14495-1 Rev. 3 TDAFW System Flow Path Verification
14551-2 Rev. 1 CCW Flow Path Verification
14808-2 Rev. 2 CCP And Check Valve Inservice Test
14825-2 Rev. 1 RCS Quarterly Inservice Valve Test
14905-1 Rev. 21 RCS Leakage Calculation
Surveillance procedure 14825-2 was conducted during the night
shift on March 22, 1989. The resident inspector conducted a
review of the data on the following morning. It was noted that
data sheet 1 (test section 5.3.1) requiring independent
verification was not documented for PORV block valves 2-HV-8000A
and B. The inspector promptly brought this to the attention of
the Operations Superintendent, OSOS, and unit SS. The SS took
the necessary corrective action to complete these steps of the
procedure on the following shift. It is apparent that an
inadequate operator and supervisory review was conducted on the
previous shift.
(7) Maintenance Activities (62703) - The inspector observed ;
maintenance activities to verify that correct equipment i
clearances were in effect; work requests and fire prevention I
work permits, as required, were issued and being followed;
quality control personnel were available for inspection
activities as required; retesting and return of systems to
service was prompt and correct; and TS requirements were being
followed. Maintenance Work Order backlog was reviewed.
Maintenance was observed and work packages were reviewed for the
following maintenance activities: ,
MWO No. Work Description
18901524 Replace NSCW Torque Switch Limiter Plate Due
To Valve 1HV-1668A Not Stroking Properly
28902508 Stroke Steam Dump Valves
28902598 Main Feed Isolation Valve Repair
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28902715 investigate / Rework / Replace Cards As Required
To Restore MFP Slave Relay K-620 To Proper
Operation
28903135 Reset Power Range Detector Current Per Start
Up Test Procedure 2-6SE-01 & 03
During this inspection, the inspectors noted that maintenance
planning and execution was effectively conducted during short
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system outages (Unit 1) and plant outages (Unit 2). Most
noteworthy was the elimination of a 10-day scheduled outage
during the Unit 2 test program due to this proficiency.
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One example of one violation was identified (paragraph 2.b(1)).
3. Startup Test Program Implementation / Verification -
Unit 2
(72302)(724008)(71715)
The inspector reviewed the present implementation of the Startup Test
Program. Inspected Test Program attributes including review of
administrative requirements, document control, documentation of major test
events and deviations to procedures, operating practices, instrumentation
calibrations, and correction of problems revealed by testing.
Periodic facility tours were made to observe Startup Test activities in
progress. The inspector verified that procedural prerequisites and
initial conditions were mat. Verification was performed by the
inspector's review of records (valve lineup sheets, test equipment
calibration status, system status checklists, or appropriate sign-offs
listed in procedure were maintained current) or by direct observation
(monitoring instrumentation indications, valve positions, equipment
position switches, or personnel actions). Discussions were held with
responsible personnel, as they were available, to determine their
knowledge of the Startup Test Program. Schedules for Startup Test Program
completion and progress reports were routinely monitored. Specific
inspections conducted are listed below:
Initial Criticality and Low Power Test Sequence
The initial criticality and low power test sequence directing the test
activities as contained in procedure 2-600-04 was reviewed during testing.
The following specific tests were partially witnessed:
(a) Step 6.2, Initial Criticality per Procedure 2-600-02
(b) Step 6.3, Determination of Low Power Physics Testing Power Range
(c) Step 6.4. Boron Endpoint, Isothermal Temperature Coefficient ;
Measurement i
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(d) Step 6.4.11, Flux Map 2-6SE-02
(e) Step 6.11, Control Bank A Worth
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Power Ascension Test Sequence (72509)(72582)(72583) l
The power ascension test secuence directing the test activities as ,
contained in procedure 2-600-13 was reviewed during testing. The ;
following specific tests were partially witnessed.
(a) Step 6.1.1, Adjustment of Nuclear Instruments to 50% Trip Level
(b) Step 6.1.7, Main Feedpump Operation per 12004-C
(c) Step 6.1.8, Perform 12004-C
(d) Step 6.1.10, 2-6AB-01, Dynamic Auto Steam Dump Control
(e) Step 6.1.11, 2-6AE-01, Automatic Steam Generator Level Control
Position Indication Test
(f) Step 6.1.20, 2-600-08, Remote Shutdown Test
(g) Step 6.1.23, 2-600-09, Loss of Offsite Power Test
(h) Step 6.4.5, 2-6SE-02, Flux Map At 30" Power
(i) Step 6.4.7, 2-6SE-03, Operational Alignment Of The Nuclear
Instruments
(j) Step 6.5.3.1, 2-6SC-02 Load Swing Test
(k) Step 6.10.2, 2-6AE-01, Automatic Steam Generator Level Control
(1) 2-600-06, MFW Dynamic Response Test
On April 2, 1989, during performance of Step 6.1.8 which directed
operation of the plant to proceed per procedure 12004-C, the inspector
observed the unit perform the transfer from auxiliary feedwater to main
feedwater for the #3 Steam Generator. Procedure 12004-C, Step 4.1.4,
specifies that the transfer is to be completed as follows:
4.1.4 TRANSFER Auxiliary Feedwater to Main Feedwater one Steam
Generator at a time by performing the following:
a. STABILIZE the SG NR level between 45% and 55%,
b. Slowly CLOSE the Auxiliary Feedwater Supply Valve and
OPEN the BFRV while maintaining SG level in program
band,
c. When the Auxiliary Feedwater Supply Valve is fully
closed, Stabilize SG level and then PLACE the BFRV in
automatic,
d. Repeat valve transfer for remaining Steam Generators.
Prior to the start of the transfer, the inspector noted that the
Balance-of-Plant Operator discussed the transfer with the operator
controlling Steam Generator level. The operators decided that the best
way to make the transfer was for the B0P operator to close the Auxiliary
Feedwater Supply Valve and the other operator would " punch" the BFRV into
automatic. The operators then commenced the transfer without discussion
with the Shift Supervisor. The B0P operator did however involve the shift
supervisor in the transfer by directing him to display narrow range and
wide range computer trends of #3 Steam Generator on the ERF computer.
Upon closing the Auxiliary Feedwater Supply Valve, the SG Water level
initially lowered. The second operator placed the BFRV into Automatic as
previously planned. The BFRV automatic control began to slowly open in
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order to restore steam generator levels. The response time necessary for
. the controller and~ valve were slow. and resulted in eventual' overshoot of.
SG 1evel to approximately 64%. The ERF computer displays were valuable in
monitoring the. inventory of water in the steam generator during- the
transient'. A'second effect was observed in the #1 SG involving lowering
. level. The. #1 SG had been transferred to its BFRV on the prior shift.
The B0P operator directed a plant equipment operator to. fail the feedpump
miniflow valve open. The inspector questioned the Operations Manager on
why the procedure had not been followed for the . transfer and why the
miniflow-valve had to be failed open. The inspector also noted that the
prior Step 4.1.3g had been signed off complete when in fact, only the #1'
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and'#3 BFIVs were open. Procedure 12004-C. Step 4.1.3.g, states: _
4.1.3.g OPEN the Bypass Feed Isolation Valve an'd VERIFY the
Feedwater Isolation Valve is closed for each SG.
The Operation Manager counseled the operators on not going in automatic
control too soon. The .failing of the miniflow valve was~ explained as a-
necessary evolution in that the flow from one feedpump feeding two steam
generators is at the point when the miniflow valve closes (500 gpm) which
affects 'the output pressure of the feedpump and hence flow to the steam
generators. By failing the miniflow valve open the feedpump performs in a
smoother manner. Later, the inspector learned that had all four BFIVs
been open, that normal leakage through.the BFRVs would account for.about
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500 gpm and the miniflow valve would have closed prior to or during the
swapover on the first steam generator.
The fact- that procedure 12004-C was not followed in Steps 4.1.3.g and
4.1.4 constitutes a violation of TS 6.1.7 requirements and is one example
of violation 50-424/89-14-01 and 50-425/89-15-01, " Failure to Implement
Procedure 12004-C, Steps 4.1.3.g and 4.1.4, For Performing Transfer From
Auxiliary Feedwater To Main Feedwater."
The inspector observed the subsequent transfer to the #4 Steam Generator
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computer. When the Shift Supervisor called up the display, he obtained
the #1 SG trend instead of the #4 SG. The transfer had already commenced'
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and'was essentially complete by the time the proper display was achieved.
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The following of procedure 12004-C, Step 4.1.4, by the operators resulted
in a smooth transfer. '
On April 3, 1989, during the performance of 2-6AB-01, " Dynamic Auto Steam
Dump Control," the plant experienced a SG level transient when a test
signal specified by procedure was incorrect. Procedure 2-6AB-01,
Step 6.3.3, directed that a test signal be- inserted equivalent to the
signal generated at a T-ref of 553 F by using Attachment 10.5.
Attachment 10.5 called for connection of a Ronan calibrator Model X85
(2.3 volt signal) (pins 26- and 27+). The reversal in polarity resulted
in the steam dumps being commanded to full open when the controller was
placed in the T-avg control mode per Procedure Step 6.3.5. At the time of
the transient, six of the twelve dumps were isolated. The resultant swell l
in SG 1evels resulted in a feedwater isolation. Further details of the
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event ; isi contained in LER _50-425/89-15. This same error ~ occurred on .
R Unit i during- the startup program; however, an LER did.not result. Unit 2
procedure development did not incorporate the Unit' 1 procedure change.
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Failure to establish an appropriate procedure is an example _of a violation:
of 10 CFR Part_50,' Appendix B, Criterion-V, and of TS 6.7.1'.a.
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50-425/89-15-01, " Failure To -Establish An Adequate Procedure For Thel
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Testing Of Steam Dumps." (Refer to the discussion'on LER 50-425/89-14'in.
paragraph 4.b(3)(r) for additional information.-)
The inspector questioned why the identical error on Unit I did not result.
l in a more severe transient. While no specific answers are known,.
speculation was made regarding the number of steam dumps that are-
inservice. On Unit 2 six of the twelve were inservice, .and the test
procedure called for verification that three valves be unisolated and
ready for testing (PV-507A, B, and C). If Unit 1 bac' only three
unisolated dumps, then the transient would not have resulted in es severe
l a level-swell. A review by the inspector on procedure 12004-C noted that
no guidance or control regarding steam dumps existed. Inspection of Unit
I revealed that one steam dump was not inservice.
The above events regarding establishment and adherence to procedures was
discussed :with the General Manager on April 6,1989. The._ inspector
addressed observations regarding:
- failures to follow procedure 12004-C,
- failure of the Shift Supervisor to closely control the operator
actions,
- failure to have appropriate procedures in place for control of
steam dumps and feedwater pump miniflow valves,
- excessive eating of food in the' control room, and
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telephone distractions to the operators.
In response to the above, the General Manager took action to address these
concerns by having by operations manager review and discuss these events 1
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with operators and supervisors.
On April 7, 1989, a feedwater isolation occurred which illustrated another .
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failure of the operators to implement procedure 12004-C. On April 6, with
the unit in Mode 3 on long-cycle cleanup, the shift f.upervisor directed
that in order to support another surveillance that long .;ycle cleanup be
secured from the control room. Following the surveillance, the cleanup j
was not restored. The following shift decided to replace the existing
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copy of 12004-C due to the number of items which had been signed off and
however no longer represented the plant configuration. Since the action i
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to secure long-cycle cleanup had been accomplished in the control room,
the shift supervisor assumed that all of Step 4.1.3 directing the stopping
of feedwater recirculation in long-cycle cleanup were not applicable.
This error resulted in the failure of the plant to close six manual
isolation valves and produced a situation wherein all four steam
generators were cross connected. On April 7, with reactor power at
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approximately 7%, operators noticed that the #1 SG BFRV was at 60% demand,
two steam generators gave the indication that only one- BFRV was
maintaining level, the operators notified I&C to investigate the
' indication problem. Since SGs 1 and 4 are on the same side of
containment, the physical piping layout results in these two SGs_ being
related. While resolving the problem, the operators decided to stroke the
- 2 MFIV as part of a maintenance functional test. As soon as the MFIV was
opened, flow from the other SGs was diverted to the #2 SG until a
feedwater isolation occurred due to Hi Hi #2 SG water level. The root
cause is related to the first shift supervisor failing to implement
procedure 12004-C, Step 4.1.3, in securing from long-cycle recirculation.
This item is an additional example of violation 50-424/89-14-01 ' and
50-425/89-15-01, " Failure To Implement Procedure 12004-C To' Secure From
Long-Cycle Recirculation." (Refer to the discussion of LER 50-425/89-15
in paragraph 4.b(3)(s) for additional information.) l
The proper control of the steam dumps was addressed by the inspector as a
concern in that the basis for the P-9 reactor protection interlock assumes
that all dumps are available with normal pressurizer pressure control.
TS 2.2.1, Table 2.2-1, item 18.3, specifies a trip setpoint of ,,50% where
the reactor trip on turbine trip can be blocked. The inspector asked for
a review by the licensee to determine if the actual setpoint should be
adjusted downward when ' dumps were not available. Followup of this item
will be tracked as IFI 50-424/89-14-02 and 50-425/89-15-02, " Review
Licensee Evaluation Regarding Adjustment Of The F-9 Setpoint When Steam
Dumps Are Removed From Service."
The above sections represent a weakness in the area of operations to
implement and adhere to the basic " Power Operation" procedure 12004-C. It
becomes apparent when combined with other operations procedure /imple-
mentation as documented in LERs 50-424/89-07, 50-425/89-02,.50-425/89-03,
50-425/89-04, 50-425/89-06, 50-425/89-08, 50-425/89-11, and 50-425/89-16
(see paragraph 4) that additional management attention and oversight are
needed. Response by licensee management has been noted; however,
effectiveness of this effort will require more time to evaluate.
The startup test program has been relatively successful with only one
noted failure discussed above regarding the steam dump testing. More ;
noteworthy was the proficient and efficient conduct of the Remote Shutdown l
Test and the Loss of Offsite Power Test. Key in the successful
accomplishment was the decision by management to perform the test only
during the day shif t at specific times. This decision affected the
appropriate personnel the ability to be well rested and prepared for the
tes ting.
Three examples of one violation and one inspection followup item were
identified.
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4.'-Review'ofLicensee' Reports (90712)(90713)(92700)
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a. In-Office Review of Periodic and Special Reports
This . inspection consisted of reviewing the below listed ' reports to_-
determine whether the.. information reported by the licensee was
technically adequate and consistent with the inspector. knowledge of-
n the material contained within the report. Selected material-within-
,
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the report was questioned randomly to verify accuracy and to provide.
a reasonable assurance that other NRC personnel have an appropriate
document: for their activities.
Monthly Operating- Report:- The inspector reviewed the Unit 1 and 2'
monthly operating reports dated March 15, 1989. This review included
the data revision for an earlier Unit I report. The inspector had no
comments.
No violations or deviations were ideritified.
b. Licensee Event Reports and Deficiency Cards
Licensee' Event Reports and Deficiency Cards were reviewed for
potential generic impact, to detect trends, and to determine whether
corrective actions appeared appropriate. Events which were reported
pursuant to 10 CFR 50.72,-were reviewed as they occurred to determine
if the technical specifications and other regulatory requirements
were satisfied. In-office review of LERs may result in further.
followup to verify ' that the stated corrective actions have been -
completed or to identify violations in addition to those described in
the-LER. Each LER is reviewed for enforcement action-in accordance
with 10 CFR Part 2, Appendix C, and if the violation is not being
cited the' criteria specified in Section V.G of the Enforcement
Pclicy was satisfied. Review of DCs was performed.to maintain a
realtime status of deficiencies,. determine . regulatory compliance,
follow the licensee corrective actions, and assist as a basis for
closure of the.LER when reviewed. Due to the numerous DCs processed
only those DCs which resuh in enforcement action or further
inspector followup with the licensee at the end of the inspection are
listed below. The LERs and DCs denoted with an asterisk indicates
that reactive inspection occurred at the time of the event prior to
receipt of the written report.
(1) Deficiency Card Review
(a) DC 1-89-831, " Inadvertent Addition Of Radioactive Gas To
Decay Tank Number 10."
On April 18, 1989, the licensee discovered that radioactive
gas was apparently added to waste gas decay tank number 10
without the lab being notified for determining the quantity
of gas contained in the tank. This deficiency will be
followed up on when submitted as an LER.
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(b) *DC12-89-985, " Unit 2 Turbine Trin Following Standby Stator
Cooling-Pump Trip."
On April.22, 1989', a-turbine trip. occurred.as'a result of a
loss of stator; cooling during a routine swapping of' stator
cooling . pumps. When~ the standby. pump.was started, both -
pumps tripped, causing. the turbine. to . trip. While
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attempting to stabilize- the plant, 'a feedwater isolation
Loccurred due to Hi-Hi SG 1evel on SG:#3, leading to an AFW -
actuation. when the running MFP tripped. The reactor was
stabilized at 2% with the SG being fed from AFW. This
deficiency will be followed up.when submitted as an LER.
(c) DC 2-89-1027' " Reactor Trip From 60% Power On A Turbine
Trip."
On May 2',.1989, the unit received a reactor vip from 60%
power on a turbine trip.: AFW' actuated on Lo.Lo SG 1evel
following the trip. All systems functioned -as required.
~The turbine trip' occurred while Engineering and a GE. Vendor-
representative were investigating a test malfunction alarm
which'was received during the weekly turbine trip device
operability test. The' cause of the turbine trip is still
under' investigation. This deficiency will be followed up
when submitted as an LER.
(2) .The following LERs were reviewed and. are ready for closure
pending verification that the licensee's stated corrective
actions have been completed.
(a) 50-424/89-06, Rev. O, " Inadequate Functional Test Leads To
Improper Termination Of Limiting Condition For Operation."
On January 30, 1989, the Gaseous Waste Processing System's
Outlet Analyzer,1 ARC-1119, failed to pass the surveillance
requirements of Technical Specification 4.3.3.10. The TS-
1- required grab samples to be taken and analyzed at least
once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. A micro fuel' cell in the analyzer was
replaced and tested on February 7,1989. On February 23,
1989, a review of the work order discovered that the
equipment was placed in service, even though a complete
surveillance test of the analyzer had not been performed to
verify that the surveillance requirements were met. The
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surveillance test was then performed satisfactorily. This
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event was caused by personnel error. Procedural
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inadequacies contributed to this event. The appropriate
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procedure was revised. The appropriate personnel have been
counsels Proper checks now exist to ensure all required
testing is performed prior to exiting a LCO. This item
represents a violation of NRC requirements which meets the
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criteria for non-citation. . In order to track this item,
the following licensee-identified item is established.
NCV 60-424/89-14-03, " Failure To Perform Required Testing 3
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Per Surveillance Requirements Results In TS 4.3.3.10
, Violations - LER 50-424/89-06."
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(b) 50-424/89-07, Rev. O, " Failure To Take Required
Temperatures Results In Inadequately Performed
Surveillance."
On February 16, 1989, while performing Procedure 14001-1,
" Shift Area Temperature Log," the plant operator noted that
there was no entry for Fuel Handling Building Room B008 for
the two previous shifts. The Shift Supervisor was notified
of the missed readings, which are required per Technical
Specification 3.7.10. The current temperature was taken
for Room B008 (76 F), and as it was well within the normal
maximum technical specification limit (104 F), no
compensatory action was required. The cause of this event
was personnel error. Two plant operators failed to take
the required reading and their respective shift supervisors
failed to note the missing temperatures when the data
sheets were reviewed. Corrective actions included
counseling of the operators and shift supervisors on the
importance of ensuring that all required technical
specification surveillance temperatures are obtained and
data sheets thorcaghly reviewed. This item represents a
violation of NRC requirements which meets the criteria for
non-citation. In order to track this item, the following
licensee-identified item is established.
NCV 50-424/89-14-04, " Failure To Take Required Temperatures
Results In Inadequately Performed Surveillance Resulting In
A TS Violation - LER 50-424/89-07."
(c) 50-424/89-08, Rev. O, " Inadequate Review Of Drawing Change
Results In Use Of Improper Breakers."
On February 23, 1989, it was discovered that 125V DC
breakers for motor-operated valves in the Turbine Driven
Auxiliary Feedwater pump system were not the proper size.
The breakers, as installed and as shown on design
drawings, were 15 amp thermal magnetic but should have been
sized as 30 amp themol magnetic per the design criteria.
Therefore, the plant has operated in a condition prohibited
by Technical Specifications. Technical Specifica-
tion 3.7.1.2 requires at least three independent steam
generator auxiliary feedwater pumps and flowpaths to be
operable. The undersized breakers were discovered as a
result of an investigation of the same problem in Unit 2.
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LCO L1-89-121 was entered. The breakers' were replaced..
successfully tested, and the LC0 was exited. The cause of-
this event was due to inadequate review by the responsible-
c engineer whenia drawing change notice corrected the M0V
horsepower rating form 0.66 hp' to 1.0 hp.- Corrective
actions . included a- review of all 125V -DC MOV- breaker.
protection. This review indicated-this incident to be an.:
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isolated . case. This' item represents a violation 'off NRC
requirements which meets the criteria for non-citation. In
order to track this item, the following licensee-identified
item is established.
NCV 50-424/89-14-05 . " Failure .To Coriduct An. Ade' quate -
Engineering Review Of The AFW Electrical System Which Led
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To AFW Inoperability Resulting In a TS 3.7.1.2. Violation -
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LER 50-424/89-08."
(d) . 50-424/89-10, Rev. 0, . " Valved Out Radiation Monitor Leads
< To_Unmonitored Liquid Waste Release."
On March 14,1989, a plant operator was preparing to
perform a liquid waste release per. procedure 13216-1,
" Liquid -Waste Release." The operator verified - that
radiation monitor 1-RE-0018 was registering normal'
background levels and that isolation release . valve
1-RE-0018 would close on a high radiation signal. ' The
release began and the operator checked the signal from
1-RE-0018 and found it was not registering above background
levels. A brief search found that the inlet valve to
1-RE-0018 was closed. This valve, 1-1901-X4-144, was -
opened; 1-RE-0018 registered the proper activity level; and
the liquid waste release continued. The release was
completed and the closure of the inlet valve resulted in
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liquid waste being released unmonitored which is a
condition prohibited by Technical Specification 3.3.3.9.
The operator omitted the performance of a pre-release line
flush which would have ensured that the inlet valve was
opened. Corrective actions included counseling the
operator and changing procedure 13216-1 to require
independent verification of the inlet valve being open.
This item represents a violation of NRC requirements which
meets the criteria for non-citation. In order to track
this item, the following licensee-identified item is
established.
NCV 50-424/89-14-06, " Failure To Follow Procedures While
Conducting A Liquid Waste Release Resulting In A TS 3.3.3.9
Violation - LER 50-424/89-10."
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l(e) 50-425/89-05,.Rev. 0, " Inadequate Review Of A Modification
' Results In A Technical Specification Violation."'
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0n. March 17, 1989, while1 investigating a proble'm'with the-
AutomaticL Surveillance Technical system, field' voltage
measurements - were taken 'that . revealed an ' electrical short
on valve 2HV-19051, the Reactor Coolant Pump #1 thermal-
. barrier isolation valve. The valve was required.to bei
operable upon entry into Mode 4, which ~ had occurred on
March.4. A Surveillance had been performed on February.4,
1989, to prove operability of 2HV-19051; however, a change
to the ASTEC system wiring. on February 10 resulted in valve
2HV-19051 being inoperable. The cause of this event was;
the issuance of an incorrect. As-Built Notice. Corrective
actions included counseling the appropriate engineering
. personnel-involved, training for all engineering personnel'
recently transferred from the Unit' 2 test organization on
use of the ABN, and issuing a second ABN to restore-the.
system to its . original configuration. This item represents-
a violation of NRC requirements which meets the criteria.
for non-citation. In order to track this item, the
following licensee-identified-item'is established.
NCV 50-425/89-15-04. " Failure To Meet A Mode Change
,
Prerequisite Resulting In A TS 3.7.12 Violation Requiring
Valve 2HV-19051 To Be Operable Prior To Entering Mode 4 -
LER 50-425/89-05."
(f) *50-425/89-06, Rev. O, " Operation Of Incorrect Handswitch
Results In Safety Injection."
On March 18, 1989, while warming main steam' lines as part
of procedure 12002-2, " Unit Heatup To Normal Operating.
Temperature And Pressure," automatic Engineered Safety
Features actuation. A- step of the procedure . called for
handswitches HS 40047/48 to be operated to reset the main
steam isolation signal. However, handswitches HS 40068/69
were operated. These switches reset the low steamline
pressure safety injection and steamline isolation logic,
removing the blocking signal. Since the main steam line
pressure was below the safety injection setpoint pressure,
the SI occurred. Appropriate ECCS pumps and valves
actucted resulting in approximately 2900 gallons being
injected into the Reactor Coolant System. The SI was
manually reset and injection into the RCS was terminated.
.The cause of this event was personnel error. The operator
failed to ensure that the proper switch was being operated.
Corrective actions will include counseling the operator on
the importance of verifying that the proper device is being
operated, changing the color of SI handswitches, adding
cautions to the handswitches, and incorporating details of
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this event into . training. This item was ? formally ,
" ' discussed following the' Enforcement-Conference on March 22,.
K .1989. ,This itcm represents a violation'of NRC requirements
which meets the criteria- for' non-citation. In-order to'
~ track this item, the following licensee-identified item is
. established,
u- .NCV 50-425/89-15-05, " Failure- To Follow Procedures -
~Resulting In Inadvertent SI Actuation - LER 50 425/89-06."
,
(g) 50-425/89-07, 'Rev. 0, " Lockup Of: A Computer. Communications-
Device Results In Containment Ventilation Isolation."
On March 19,1989, 'while restoring the Plant Effluent.
Radiation Monitoring ' System to service ' the plant
experienced 'an automatic Engineered Safety Features
actuation which resulted in a ' Containment Ventilation
Isolation. Appropriate valves and dampers actuated to.
isolate containment ventilation. _ Control room operators
verified that no abnormal radiological' conditions existed
using'2RE-0002/0003. The monitor that actuated the CVI,
2RE-2565, was placed in bypass. The CVI was. reset and
equipment that actuated was returned to normal _ operating
position. Due to an earlier SI,. power was-lost to most of
the. PERMS system. On ' restoration of power, the compute _r
parameter files are initialized with a -9.99E-20 value.
The_ computer replaces this value with pa* 1 meters received
from each monitor. Due to a communication failure of a
multiplexer, communication with the monitors was lost and
no value was received for 2RE-2565. When the mutiplexer
was reset the computer detected the original power failure
, for 2RE-2565. On a power failure, the computer gives'the
monitor the current" parameter on file and assigned the
monitor -9.99E-20 value. This resulted in 'a high alarm,
causing the CVI actuation. Corrective action is a
procedure revision to require 2RE-2565 to be placed in
bypass when the computer is initialized to receive
parameters.
(h) 50-425/89-09, Rev. O, " Procedure Misinterpretation Leads
To Late Surveillance Testing." ,
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On March 20, 1989, a diesel fuel oil shipment arrived l
onsite for offloading into the Diesel Fuel Oil Storage
tanks. A technician obtained and analyzed a sample. The
technician and his foreman interpreted a note in the
analyses scheduling procedure to mean that the
neutralization number and mercaptan were not required to be j
performed. In fact, only the mercaptan was exempt from the !
analysis and neutralization number was required to be
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performed. After the analysis found the other fuel ,
properties to be satisfactory, the shipment was unloaded l
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into the DF0S tanks. Meanwhile, a second diesel fuel oil
shipment arrived onsite, a sample was'obtained and analyzed
as before and unloading into the DF05. tanks began. A-
laboratory supervisor reviewed the data sheets and question
the omission of the neutralization number from the data
sheets. After the requirement was clarified, - the
technician obtained the original samples from each shipment
and determined that the neutralization number of each was
within technical specification requirements. The cause of
this event was the misleading nature of the procedure note.
The procedure note was rewritten and clarified. This item
represents a violation of NRC requirements which meets the
criteria for non-citation. In order to track this item,
the following licensee-identified item is established.
NCV 50-425/89-15-06, " Failure To Establish An Adequate
Sampling Procedure For Diesel Fuel Oil Per is 6.7.1.a - LER
50-425/89-09."
(i) 50-425/89-10, Rev. O, " Radioactive Discharge Without Permit
Leads To Technical Specification Violation."
Technical Specification 3/4.11.1 requires that releases of
radioactive materials to unrestricted areas be sampled and
analyzed for appropriate alpha, beta, and gamma emitters.
On March 8, 1989, the contents of the Unit 2 Turbine
building drain tank, 2-2412-T4-002, were sampled for gamma
emitters to determine if a release permit was required. On
March 9, a plant operator released the tank contents to the
Unit 2 Waste Water Retention Basin without a permit. On
March 14, during a review of releases, it was found that no
permit had been issued for the March 9, release. The
permit ensures that required samples have been taken,
analyzed and are within allowable limits for releases.
Procedure 13211-2, " Turbine Building Drain System,"
required that sample analysis be used to determine how
drain tank contents are to be processed but did not specify
that a release permit may be required. The cause of this
event was that the operator did not obtain a radioactive
release permit prior to releasing. Procedure 13211-2 has
been revised to provi-de specific instructions that a
radioactive release permit may be required for releasing ;
the contents of a turbine building drain tank. Also, at ;
shift briefings, operators were reminda.d that waste permits
are required prior to release of radioactively contaminated
tank contents. This item represents a violation of NRC
requirements which meets the criteria for non-citation. I r.
order to track this item, the following licensee-identified
item is established.
NCV 50-425/89-15-07, " Failure To Obtain A Radioactive
Release Permit Prior To Releasing Radioactive Materials To
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Unrestricted Areas Resulting In A TS 3/4.11.1 Violation -
LER 50-425/89-10."
(j) 50-425/89-12, Rev. O "Oper' ting Incorrect Switch Results
In Inoperable Monitor Requi >.ng Entry Into TS 3.0.3."
On March 30, 1989, while performing maintenance on
2RE-2562A, an Instrument and Controls Technical
inadvertently placed' 2RE-2562A and 2RE-2562C in purge
instead of activating the paper drive on 2RE-2562A. This
caused 2RE-2562C to be -inoperable. Later the same day, a
chemistry foreman discovered 2RE-2562C to be inoperable and
notified the control room. .An entry into TS 3.0.3 was made
due to an existing limiting condition for operation.for the
Reactor Coolant System Leakage Detection System and
2RE-2562C being inoperable. With 2RE-2562C inoperable the-
LCO for Technical Specification 3.4.6.1 could not be met.
2RE-2562C was restored to service and TS 3.0.3 exited. The'
cause of this event was personnel error. The I&C
technician failed to pay attention to detail when
activating plant equipment. The purge switch was activated
instead of the paper drive. Corrective actions included
counseling the individual and issuing a memo to all I&C
personnel concerning attention to detail when performing .
maintenance / trouble shooting on plant equipment. This item
represents a violation of NRC requirements which meets the
criteria for non-citation. In order to track this item,
the following licensee-identified item is established.
NCV 50-425/89-15-08, " Failure To Follow Procedures While
Performing Maintenance On 2RE-2562A Resulting In The Plant
Operating In A Condition Prohibited By TS Thus Requiring
Entry Into TS 3.0.3 - LER 50-425/89-12."
(k) 50-425/89-13, Rev. O, " Flood Barrier Removal Leads To
Auxiliary Feedwater Inoperability."
Technical Specification 3.7.1.2 requires that three
independent steam generator AFW pumps and associated flow
paths be operable in Modes 1, 2, and 3. On March 30, 1989,
plant personnel were conducting a routine walkdown. They ,
found a flood protection barrier removed from the wall l
between che AFW discharge piping room (room 105) and the
Turbine Driven AFW pump room (room 106). The barrier was
replaced and the TS action statement was exited. The cause
of this event is an apparent personnel error by removing
the barrier without the proper review and approval. Work
had been performed on a check valve in room 105. When a
functional test was performed on March 23, the existence of
a flood barrier and precautions to be observed were not
addressed by those requesting the test or by those
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implementing'the. work order..-A sign will be installed near
the flood ~ barrier and information. will be added to the-
equipment file advising of the: flood barrier's existence.
This item represents a violation of NRC requirements which
meets' the criteria'.for non-citation. . In order to' track
this- item, the following licensee-identified- item . is-
established.
. .NCV 50-425/89-15-09, " Failure To Maintain The Auxiliary
Feedwater System Operable Resulting. In A Condition
Prohibited By TS 3.7.1.2. - LER 50-425/89-13."
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(1) *50-425/89-16, Rev. O, " Unplanned Auxiliary Feedwater
Actuation On Recovery From Remote Shutdown Test."
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On April 11, 1989, while recovering from a Remote Shutdown
Test, -an automatic Engineered Safety Features actuation
(auto start signal to motor. driven Auxiliary Feedwater
pumps) occurred. During the Remote Shutdown test, both-
Main Feedwater Pumps were manually tripped and AFW was in-
service. With both .MFPs tripped an AFW actuation signal
was generated; however, .while control- was at the Remote
Shutdown Panel, the signal is interrupted. . - When control
was returned to the control room, the signal was
reinstated. As the AFW pumps were already in operation,
the AFW actuation signal caused the discharge. valves of the
L Train. A to stroke full open. Control room operators
immediately throttled AFW flow to-_ prevent overfilling of
the steam generators. MFP "A" was reset to allow return of
the remaining trains to the control room. All AFW systems
were' restored to readiness. The cause of this event was a
situation that was not anticipated by the procedure.
Procedure 18038-2, " Operation From Remote Shutdown Panels,"
will be revised to caution. operators of a possible
actuation of transfer of control to the control room.
(3) .The following LERs were reviewed and closed.
(a) 50-424/87-81, Rev. 0, " Excessive Valve Weight Could Have
Prevented Fulfillment Of Safety System Function."
On May 5,1987, two valves supplied by Anchor Jarling Valve
on the sludge mixing recirculation line of the Refueling
Water Storage Tank were found to weigh significantly more
than shown on the A/DV drawings. The initial analysis from
an employee of Bechtel Power Corporation indicated that the
valves weighed in excess of the seismic design capacity of
their associated pipe supports and that if a line failure
had occurred in the non-safety related portion of the
sludge mixing line during a seismic event, the valves could
have been closed and allowed the RWST water volume to be
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available for plant shutdown. On March 6, -1989, the
Project Field Engineering-0ffice advised plant personnel
that there was an error in the application af potential
failure point and that the potential failure point was
actually between the valves and the RWST. Thus, if a
seismic event caused a line failure to occur, the broken
line could have potentially drained the RWST to a level
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below minimum requirements for plant shutdown. The cause
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of this condition was determined to be the failure of A/DV
to advise Bechtel of a change in valve weights from those
originally shown on the valve drawings and an error by a
Bechtel Power employee in the initial review of this
condition. Corrective actions included adding an
additional pipe support and reviewing other safety related
valves for weight discrepancies. The-inspector has no
further questions.
(b)*50-424/88-16, Rev. O, " Water Leakage Into Control
Room / Potential Exists For A Safety System Failure."
On June 3, 1988, smoke from an electric duct heater
actuated smoke detection alarms. Although sprini cr heads
did not actuate, water from the preaction valve leakoff
lines ran into the upper cable spreading room and seeped
into the control room from the ceiling. Water entered some
process panels and led to spurious equipment actuations in
the Reactor Coolant System which were promptly addressed
and corrected by control room personnel. On June 5, 1988,
it was concluded that a condition existed which alone could
have prevented the fulfillment of the safety function of a
systera needed to mitigate the consequences of an accident.
The cause of this event is an inadequate design of the
control room ceiling penetrations which are supposed to be
watertight. Corrective actions were verified complete.
This item resulted in a NRC violation 50-424/88-24-01.
(c) 50-424/88-19, Rev. O, " Inadequate Installation Leads To
Containment Ventilation Isolations."
On June 10, 1988, a CVI occurred due to an apparent power
supply failure in radiation monitor 1RE-2565C. The
appropriate dampers and valves actuated as designed.
Control room personnel verified that no abnormal condition
existed. 1RE-2565C was bypassed and the CVI signal was
reset. Later, the same day, another CVI occurred, when
plant personnel removed 1RE-2565C from bypass in order to
reenter monitor setpoints. Again the proper dampers and
valves actuated and control room personnel verified that no
abnormal radiation condition existed. 1RE-2565C was again
placed in bypass and the CVI signal was reset. An
investigation demonstrated that the cause of the CVI was an
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inadequate installation . which left ' a flow' transmitter
. shield wire _ exposed that electrically grounded, simulating.
a loss of. power. Corrective action included _ insulating the
shield wire ~and new default values were installed.
1 (d) 50-424/88-20..Rev' 1, " Inadequate Breaker Leads.:To
Condition Prohibited By Technical Specification."
On June 29, 1988, it was determined that. ten containment
penetrations may not have adequate redundant overload
protection, as. required by Regulatory Guide 1.63. The
redundant protection was not provided because-in each of
the~ ten penetration circuits one of the two breakers use_dL
was magnetic-only, which did not provide adequate overload
. protection for the penetration. The other. breaker pro ~ided-
v
was a ' thermal-magnetic and provided adequate ' overload
protection for the penetration. Since the magnetic-only
breakers did not provide the redundant overload protection,
the requirements of . Technical Specification 3.8.4.1 for
operability was not satisfied. When it was determined that.
redundant overload protection may not have been adequate
over the entire range, the identified containment
penetrations were declared inoperable and the requirements
of Technical Specification 3.8.4.1 were satisfied while the
breakers were being replaced. Prior' to. the operation of
Vogtle Unit 1, a construction test was performed for each
breaker to verify its tripping function. All tests were
performed satisfactorily and the breakers declared
operable. The inspector has reviewed documentation which
indicated that the- corrective action was complete. The
magnetic-only breakers were replaced with thermal-magnetic
breakers.
1
(e) *50-424/88-22, Rev.1, " Failed Potential Transformer i.eads
To Turbine / Reactor Trip."
On July 14, 1988, a generator / turbine / reactor trip occurred
as a result of an overexcitation condition on the generator
.
field. Control rods inserted. The Main Feedwater system
l
isolated and the Auxiliary Feedwater system actuated.
Control room operators responded . properly to assist in
plant stabilization. An investigation revealed that the
failure of a potential transformer caused the primary fuse
to blow. The resultant transient caused the GENERREX
voltage regulator to malfunction, increasing generator
voltage to the Volts / Hertz relay setpoint, which
subsequently initiated a generator / turbine / reactor trip.
Corrective action includes replacing all primary PT fuses,
PT 2A, and the malfunctioning circuit boards in the
GENERREX system. The GENERREX system's operational history
has been evaluated and additional adjustments are not
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considered necessary at. this time. Engineering review of
design enhancements to .the present GENERREX system will i
continue to be performed as part of the Trip Reduction i
Program. The failed PT was analyzed and a winding failure
was identified. Improved test methods to detect this-type
of PT failure were evaluated. However, a more appropriate
'
test method has not been identified. This LER was closed
in report 50-424/88-37.
(f) 50-424/88-23, Rev. O, " Inadequate Design Leads To Condition
Prohibited By Technical Specification."
On July 29, 1988, LER 50-424/88-20 was issued, identifying-
that several electrical penetrations may not have been
provided with adequate redundant overload protection. As a
result of the interpretation for deportability of that
event, two previously identified deficiencies have been
re-evaluated for deportability. As a result of the
re-evaluation, an event that was discovered on August 14,
1987, was determined to be reportable on July 28, 1988.
The other event was discovered on July 7,1987, and
determined to be reportable on August 11, 1988. It was
determined that for each event, redundant overload
protection may not have been adequate for the entire range
of protection . as required by Regulatory Guide 1.63.
Technical Specification 3.8.4.1 required that electrical
penetration overload protection may not have been provided
for several penetrations, Unit 1 may have been operating in
a condition prohibited by TS until the event was
discovered. For each event the limiting condition for
operation action statement for TS 3.8.4.1 was implemented
on the event discovery dates of July 7,1987, and
August 14, 1987. The event on August 14, 1987, involved
electrical penetrations No.12 and No. 69, concerning the
- 12 and #14 size conductors. The other event on July 7,
1987 involved penetration No. 03, 14, 34, 41, 60, and 61,
concerning #10 size conductors. The inadequate overload
protection was discovered during a broadness review for
Unit 2 by the designer, Bechtel Power Corporation. The
inspector verified the work complete by reviewing the
closed MW0s.
(g) 50-424/88-26, Rev. O, "Use Of Improper Tools Leads To
Containment Ventilation Isolation."
On September 7, 1988, an electrician was in the process of
installing shorting bars into fuse holders following the
completion of an electrical switch replacement. The y
electrician unintentionally created a short between two 120 ?
volts AC circuits. Various alarms and indicators actuated,
including those for a CVI. The appropriate CVI valves and
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dam;:ers ' actuated. Control room personnel verified that no
abnormal radiation condition existed by observing redundant ~
-monitors. The control room personnel and the electrician
immediately confirmed that the electrical short had
'
initiated the "JT . The cause of this event was the use of
an improper toct -y the electrician. Fuse pullers provided
to the electrician would not fit between the inserted
shorting bars, so he used needle-nose pliers to perform the
insertions. These pliers made the electrical short by
simultaneously contacting two shorting bars following one-
shorting bar's insertion. Appropriate personnel were
advised to avoid the use of needle-nose pliers or makeshift
tools for installation of fuses or shorting bars. The
proper size fuse-pullers were made available.
(h) 50-424/88-30, Rev. 0, " Surveillance Missed Due To
Inoperable Rod Position Deviation Monitor."
On October 27, 1988, while preparing a licensing document
change, it was discovered that a plant computer design-
feature _ for monitoring deviations between Digital Rod
Position Indication System and Domand Position Indication-
System had not been implemented within the plant computer
software as intended. The absence of this feature means
the Rod Position Deviation Monitor is operable for this
function and that surveillance 4.1.3.2 has not been met,
when required, since issuance of the Unit I license. The
surveillance required operability determination of the
digital rod position indicators. For this determination,
the DPIS must be verified to be with + or - 12 steps of the
DRPIS every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, except when the RPDM is inoperable,
then the requirement is at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. As the
plant staff were unaware of the software omission, they did
not take the required action to manually make the
comparisons every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> as required. The cause of this
event was the omission of appropriate rod supervision.
programs in the original vendor supplied computer software
specifications. Corrective actions include increased
frequency of the surveillance and an evaluatica to
determine if either changes to W computer software are
feasible or changes to licensing documents are required.
The inspector reviewed documentation which indicated that
the corrective action was complete. This item represents a
violation of NRC requirements which meets the criteria for
non-citation. In order to track this item, the following
licensee-identified item is established.
NCV 50-424/89-14-07, " Failure To Conduct Surveillance
Resulting In A Violation Of TS 4.1.3.2 - LER 50-424/88-30."
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(i)-*50-424/88-41, Rev. O and 1, " Containment Purge Supply
Isolation Valve Inoperable Due To Failure To Fully Close."
On December 13, 1988, while. performing 'a revised Type C
Local Leak Rate Test- for. surveillance of the containment
purge supply isolation valves in Penetration 83, it was.
discovered that the 24-inch containment purge supply
isolation valve 1-HV-2626A was not fully seated.
'
This
condition is prohibited by Technical Specification 3.6.1.7
which requires that this valve be closed and sealed closed. ;
LC0 1-88-922 was entered for 1-HV-2626A failing the leak
'
. ate test. This event occurred because the valve did not
fully close, even though the limit switch indicated that
the valve was closed. Corrective actions included issuing
LC0 1-88-922,1mmediate manual seating of the valve and
successfully repeating the LLRT, and . establishing
conservative administrative controls to ensure that each
24-inch purge isolation valve, if cycled, will be either
manually seated or have an LLRT performed, as appropriate.
Procedures 13125-1, Rev. 8, and 13125-2, Rev. 2, were
verified by the inspector to have been revised.
(j) *50-424/89-05, Rev. O, " Trip Of Main Feed Pump On High !
Vibration Resulting In Manual Reactor Shutdown."
l
On February 10, 1989, Control Room operators received Main
Feedwster Pump Turbine "A" high vibration alarms. A check
of the vibration monitor system showed a vibration of only
1.2 mils. (The vibration system alarms at 3 mils and trips ,
at 5 mils). Shortly thereafter, MFP " A ". tripped. t
Steam /feedwater flow mismatch alarms were received on all >
four steam generators. Turbine load was manually reduced l
to approximately 700 MWe and control rods placed in Auto to ;
'
follow load. Steam dump valve controllers were manually
operated to attempt to match steam / feed flow. SG #4
reached 20% level and the Shift Supervisor directed the !
reactor to be manually tripped. Feedwater isolation and 1
start of Auxiliary Feedwater pumps occurred as expected.
However, the Turbine Driven AFW pump tripped on overspeed ;
after startit:g. The cause of the MFP high vibration trip l
was not positively identified. The cause of the TDAFW pump
overspeed trip, although not positively identified, may l
have been caused by particulate contamination of the lube :
oil, which serves as the control system hydraulic fluid. !
Corrective actions included temporarily installing l
vibration instrumentation to collect MFP vibration data. !
Additional surveillance were also performed on the TDAFW
pump to ensure operability.
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(k) 50-424/89-09, Rev. 0, " False Radiation Monitor Signal
Caused Containment Ventilation Isolation And TS 3.0.3
Entry."
On March 13, 1989, radiation monitor 1RE'-0003 spiked high
causing a Containment Ventilation Isolation. Appropriate
valves and dampers actuated from the CVI signal to isolate
containment ventilation. LC0 1-89-155 was entered for
1RE-0003. Radiation monitor 1RE-0002 was out of service
for a surveillarice and 1RE-2565 was not operable because of
reliability concerns. Technical Specification 3.3.2,
Table 3.3-2, requires a minimum of two of the three channel
be operable, but there is a provision for operation with
only one channel in operation. An entry was made into TS 3.0.3 since all three channels were inoperable. Control
room operators verified that no abnormal radiological
-
conditions existed using IRE-0002, which was functional but
not operable. Later that same day, 1RE-0002 was declared
operable, the high alarm on.1RE-0003 was cleared, the
monitor placed in. bypass, and the CVI signal was reset.
The cause of this event was the failure of the detector
tube. The tube was replaced; however, the replacement tube
did not function properly and required replacement due to
degradation of the voltage plateau. The replacement tube
was monitored and the monitor was declared operable.
(1) 50-425/89-01, Rev. O, " Spurious. Signal Resulting From
Circuit Board Causes Control Room Isolation."
On February 14, 1989, a Control Room Isolation occurred due
to a spurious signal from radiation monitor channel
2RE-12116. Prior to this actuation, the Safety Parameters
Display Console had received intermittent trouble light
indications from the channel. Control room operators
verified no high radiation condition existed. The
monitor's output was blocked, a LC0 was entered, the CRI
signal was reset, and normal ventilation was established.
Radiation monitor channel 2RE-12116 was returned to service
and the LCO exited on February 18. The event was caused by
a random failure detected on the Central Processing Unit
board in the Digital Processing Module. This random event
caused the internal timer to lock up and initiate a system
reset signal. During a system reset, the monitor's fail
safe function initiates a high alarm signal which caused
the CRI actuation. Corrective actions included initiation
of a LC0 for the monitor, replacement of the defective
circuit board, observation of the monitor for proper
operation and return of the monitor to service.
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L (m)'*50-425/89-02 Rev.0- " Opening Discharge Valves Causes Plant
Operation'0utside Of Technical Specifications."
Technical Specification .Se'ction. 3.4.1.4.2 _ states,"
... Reactor Makeup Water Storage Tank discharge ' valves
- (1208-U4-175, 1208-U4-176, 1208-U4-177, and 1208-U4-183)
.shall be closed and secured in position (in) Mode 5 with
reactor coolant loops not filled.". On February 19, 1989,
the unit made its initial entry into Mode 5. - valves
2-1208-U4-175 and 2-1208-U4-177 were opened. After shift
change, new shift personnel realized that the; reactor
o . coolant system loops were not filled and that the two open-
discharge valves were required to be closed. -A LCO was
initiated, the valves were closed and locked, and the LC0
was terminated. Plant personne1' believed that filling .the -
RCS above the loops to L the reactor vessel flange leve11
constituted a " loops filled" condition, after which opening
the discharge' valves would have been; permissible. With the
-discharge. valves open, an inadvertent dilution event of the-
RCS could have been initiated. A TS interpretation of what.
constitutes " loops filled" has been.added to the Operations
Required Reading Boolg The personnel involved were
counseled regarding the' importance of complying with TS.
Inspector. followup determined that prior to the Mode 5
entry, the.SS had been asked to open these same valves to
allow chemistry to add primary chemicals. At that time,
the SS was : aware that TS 3.9.1 required the valves to be
maintained shut in Mode 6 and thought that the change to
Mode 5 would allow the evolution. TS 3.4.1.4.2 however,
also controls these valves when the RCS loops are not
filled. Operations procedure 12006-C established positive
control of these valves by tagging them closed. These
valves are untagged by operations procedure -13000-2 upon
filling and completing air sweeping of the RCS. The.
removal of the RMWST valves to the CVCS was a discussion
item at the shift turnover, however, neither SS recognized
the consequences. Later in the shift, the deficiency was
identified and corrected. This item was formally discussed
' following the enforcement conference on March 22, 1989.
This item represents a violation of NRC requirements which
meets the criteria for non-citation. In order to track
this item, the following licensee-identified item is
established. !
NCV 50-425/89-15-10 " Failure To Maintain RMWST, Discharge
Valves Shut Closed And Secured In Position While In Mode 5
Resulting In TS 3.4.1.4.2 Violation - LER 50-425/89-02."
(n) 50-425/89-03, Rev. O, "Depressurizing RHR System Leads To
Technical Specification 3.0.3 Entry."
On March 9,1989, with the unit having just entered mode 3
for the initial heatup, preparations were being made to
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perform the Pressure.1 solation Valve Leakage Test. In
order to ensure proper pressure across the valves :to be
-
tested, the Shift Supervisor decided, without an approvedi
.
procedure, to depressurize the Residual Heat Removal
system, using the RHR test return valves. The SS directed.
a momentary opening of these valves. This resulted in the
return .line valves being left open for approximately 14
hours, reducing the flow capacity of both RHR trains, and
leading to operation under Technical Specification 3.0.3
provisions. This event was caused by -(1) operations -
personnel attempting an evolution without approved
procedural guidance, (2) lack of closed loop communication.
and (3) inadequate system status sensitivhy by the
operations' shift team. Corrective actions include s(1)
counseling the Shift Supervisor and briefing of each
operating crew by th Plant General- manager on the
importance of conducting plant evolutions with approved
procedures, (2) changing the appropriate procedure, (3)
stressing precise control room . communications,. (4)
stressing sensitivity to system status in shift briefings
'and requalification training, and (5) improving the locked
valve program.. . This item was. cited as a NRC violation. in
report 50-425/89-12. ~ Remaining corrective actions will.be
verified in closecut of the violation. -
(o) *50-425/89-04, Rev. O, " Reactor Coolant System Leakage
During Check Valve Testing."
On March 9,1989, with Unit 2 in Mode 3, plant operations
personnel performed a pressure isolation valve leakage
test. The Primary Coolant Loop #3 Cold Leg Check Valve
(2-1204-U6-085) exhibited excessive leakage. A
Notification of Unusual Event was declared, because.the
Reactor Coolant System leakage - exceeded the technical
specification limit of 5 gpm specified in Section
3.4.6.2.f. On March 10, 1989, the plant entered Mode 5 and
the NUE was terminated. The event was caused by excessive
wear on internal check valve components. Wear was found
near the' pivot pin which allowed the disc to drop down and
not seat properly. The valve consists of a disc with two
arms which insert into a lock block. The pivot pin goes
into the lock block. The disc arms are notched out for
alignment with the pivot pin. Wear was found on both
notches in the arms which allowed the disc to drop.
' Corrective action included replacement of the internal
components in this valve and the three identical check
valves in the other three loops.
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(p)*50-425/89-08, Rev. O, " Improper Control Of Steam Generator
Water Level Leads To Feedwater Isolation."
On March 19, 1989, unit 2 heatup was in progress. The unit
Balance-of-Plant operator was manually controlling the
steam generators water levels when a technician requested
I his assistance in performing a surveillance test. The B0P
operator left the front panel to go to a back panel area.
When he returned several minutes later, he found that an
automatic feedwater isolation had occurred because SG #4
had exceeded the 78% (narrow range) high-high water level
setpoint. The operator stopped the feed to SG #4, returned
the flow to normal, and long cycle recirculation was
L re-established. The B0P operator intended to leave the
front panel for only a few moments and did not request
relief. This is the direct cause of this event.
Contributing to this event was the Shift Supervisor's
omission in assigning a dedicated Steam Generator Water
Level Controller which is the plant policy when manual SG
feeding it. in progress. The BOP operator was counseled
regarding the importance of maintaining a continuous watch
on operations in progress or else requesting relief if
needed. The SS was advised of the necessity to comply with
plant practice to have a dedicated SGWLC when manual SG
feeding is in progress. This item was formally discussed
following the enforcement conference on March 22, 1989.
This item represents a violation of NRC requirements which
meets the criteria for non-citation. In order to track
this item, the following licensee-identified item is
established'.
NCV 50-425/89-15-11 " Failure To Exercise The Duties And
Responsibilities Of The R0 And SS As Delineated In
Operations Procedure 10000-C - LER 50-425/89-08."
(q)*50-425/89-11, Rev. O, " Valve Closure Leads To
Non-Compliance With Technical Specifications."
Technical Specification 3/4.5.2 requires that the Safety
Injection Pump Cold Leg Injection valve 2-HV-8835 be open
while in Modes 1, 2, and 3. On March 19, 1989, the shift
operating crew closed the Safety Injection pump cold leg
injection valve to the Reactor Coolant System cold legs
(2HV-8835) while performing the system operating procedure
to fill SI accumulators at low RCS pressure in Mode 3.
Closure of this valve prevents both SI pumps from being
capable of providing automatic injection to the RCS cold
legs upon receipt of a SI actuation signal. On March 26,
while considering LER 2-89-003 (both trains of Residual
Heat Removal rendered inoperable due to common valve
manipulations) and similar situations for other
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safety-related. systems, a shift supervisor realized that
the system operating procedure for filling SI accumulators
at low RCS pressure requires closure of 2HV-8835 while in
,
Mode 3. Upon discovering this, a review of the Unit 1 and
Unit 2 accumulator fills was initiated. Nine separate
instances were identified for Unit I when 1-HV-8835 was
closed while in Mode 3, in addition to the single
occurrence on Unit 2, specified previously. The cause of
these events is inadequate procedures which did not prevent
closure 2HV-8835 during Mode 3 or require accumulator fill
prior to Mode 3 entry. The procedures are being changed to
correct these inadequacies. Future followup on this LER
corrective actions will be in closecut of the violation.
This event is one example of violation 50-424/89-14-01 and
50-425/89-15-01, " Failure To Establish An Appropriate
Procedure To Maintain SI Operable While Filling
!
(r) ~ *50-425/89-14, Rev. 0, "Feedwater Isolation Results From
Error In Startup Test Procedure."
On April 3, Unit 2 startup testing was in progress. A test
signal was incorrectly inputted into the steam dump control
circuit causing the steam dumps to fully open instead of
opening 10% to 15% as expected. This led to a steam
generator water level swell and a feedwater isolation due
to SG #4 reaching the high-high level. Main feedwater
isolation occurred as designed, and the safety grade
isolation valves closed, but main feed pump "A" did not
trip. As a result, the Auxiliary Feedwater system did not
automatically start, although it was already being used to
supply SG water. Manual control was taken of the Steam
Generator Feed and unit parameters were stabilized. The
test procedure, which called for an incorrect test signal,
was corrected and the remaining startup tests are being
reviewed to ensure that proper connections are specified.
Sliding links associated with MFP "A" circuits were found
open and are believed to be an oversight from the Unit 2
construction phase. Similar slidici links were inspected
to ensure closure.
This item is part of one ex%+ k of violation
50-424/89-14-01 and 50-425/89-15-01 discussed in paragraph
3.
(s) *50-425/89-15 Rev. O, " Faulty Circuit Cards Results In ESF
Actuations."
On April 5,1989, a spurious trip of Main Feedwater Pump
"A" generated a Feedwater Isolation signal and automatic
actuation of the Auxiliary Feedwater System. On April 7, a
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FWI and AFW actuation occurred when a steam generator
reached its high-high level setpoint during a test of a
Main Feedwater Isolation Valve. On April 9, a second
spurious trip of MFP "A" generated a FWI and subsequent AFW
actuation. The cause of the April 5 and April 9 events was
faulty circuit boards in the Solid State Protection System 1
'
logic circuits. The April 7 event, although not directly I
caused by a faulty circuit card, was a consequence of the
valve lineup used to functionally test repairs made
following the April 5 event. The lineup of long-cycle
recirculation was not properly restored prior to resumption
of startup testing. Corrective actions include replacing
the faulty- circuit boards and counseling . plant operators
regarding proper shift turnover of unusual plant
configurations and the need for procedural compliance.-
This event is part of one example of violation
50-424/89-14-01 and 50-425/89-15-01 discussed in paragraph
3.
One example of a cited violation and thirteen non-cited violations
were identified.
5. Actions on Previous Inspection Findings - (92701)(92702)
a. (Closed) Violation 50-424/87-30-03, " Failure To Properly Close
Valve."
The inspector reviewed the licensee respor e dated July 13, 1987.
Valve No. 1-1208-U4-348 has had the lock removed to preclude future
errors in positioning from the renote operator.
b. (Closed) Violation 50-424/88-05-02, " Lack Of Material Control."
The inspector reviewed the licensee response dated March 10, 1988.
The inspector noted that procedures exist to control the purchase and
receipt of weld rod.
c. (Closed) Violation 50-424/88-24-01, " Failure To Adequately Design
And Install Water Tight Penetration Seals And Perform An Analysis i
Which Evaluates Their Failure."
The inspector reviewed the licensee response dated September 15, 1988 i
and reviewed completed MW0s 18900130 and 18900180. During this
inspectica period, a similar actuation of the fire suppression system
occurred which challenged the seal configuration. Observation by the
NRC inspector at that time noted that no water penetrated into the
Control Room.
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d. (Closed)'IFI 50-424/88-43-01, " Verify Resolution Of Restoring The j
, SSMP To A Condition To Correctly Indicate The Operability Status." <
.The licensee corrected the condition by implementing a design change
which removed the Boric Acid Pump Motor handswitches as an input to
the SSMP. The inspector urified the change was implemented on
'
Unit 1. Following the verification, the inspector noted that Unit 2
had- not implemented a similar change. The inspector was informed
that design change MDD 89-V2M035 was being developed for Unit 2. The
inspector considered the late implementation of a Unit 2 change to be
a weakness in the area of engineering support in maintaining the
designs both units identia , , as possible. This change involves the
lifting.of two leads in eatn train panel. To track the accomplishment
of Unit 2 change, the foi'owing inspector followup item is
identified.
IFI 50-425/89-15-03, " Verify Resolution Of Restoring The SSMP To A
Condition To Correctly Indicate The Operability Status."
e. (Closed) Violation 50-424/88-56-01, " Failure To Implement Operations
Procedure 14900-1, Containment Exit Inspection Required By TS 6.7.1."
The inspector reviewed the licensee response dated March 7,1989.
Corrective actions have been observed in'pri. tice by the inspector.
Procedure 43006-C was revised to include controls for health physics
responsibilities.
f. (Closed) Unresolved Item 50-424/88-56-02, " Review Licensee
Evaluation Of Compliance To 10 CFR 50.62."
This item concerned the sensitivity of unit personnel to the proper
operation and maintenance of AMSAC equipment. The licensee has
implemented quarterly and refueling surveillance procedure 54804-1,
revised response procedure 54804, and revised response procedure
17005-1. Unit operating procedures 12004-C has been revised to the
correctly indicate the power level where the equipment becomes i
operational. Failure to comply with 10 CFR 50.62 was the result of a
failure to establish adequate procedures. Failure to comply with 10
CFR 50.62 was the result of a failure to establish adequate
procedures.
This item is considered to be one of the examples of violation
50-424/89-14-01 and 50-425/89-15-01, " Failure to establish adequate
procedures to ensure AMSAC was available.
g. (Closed) Violation 50-424/88-61-01, " Failure To Implement Operations
Procedure 10001-C, Required By TS 6.7.la, To Annotate And Verify
Proper Operations Of Control Room Chart Recorders."
In the licensee response dated March 7,1989, to the Notice dated
January 20, 1989, the licensee committed to full compliance on
January 31, 1989, upv 'ssuance of standing order C-89-01. This
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standing order was reviewed by the resident inspector on March 24,
1989, and was found to be satisfactory.
One example of a cited violation and one inspector followup item were
identified.
6. Exit Interviews - (30703)
The inspection scope and findings were summarized on May 5,1989, with
these persons indicated in paragraph 1 above. The inspector described the.
areas inspected and discussed in detail the inspection results. No
dissenting comments were received from the licensee. The licensee did not
identify as proprietary any of the materials provided to or reviewed by
the inspector during this inspection. Region based NRC exit interviews
were attended during the inspection period by a resident inspector. This
inspection closed five violations (paragraph 5), one unresolved item
(paragraph 5), one inspector followup item (paragraph 5), and nineteen
Licensee Event Reports (paragraph 4.b(3)). The items identified during
this inspection were:
Violation 50-424/89-14-01 and 50-425/89-15-01 contains six examples where
procedures were not either established or implemented as follows:
- " Failure To Implement Procedures 00101-C and 50009-C Resulting In TS 6.7.1.a Violation" - paragraph 2.b(1)
- " Failure to Implement Procedure 12004-C Step 4.1.39 and 4.1.4 for
Performing Transfer From Auxiliary Feedwater to Main Feedwater" -
paragraph 3
- " Failure To Establish An Adequate Procedure For The Testing Of Steam
Dumps" - paragraphs 3 and 4.b(3)(r)
- " Failure To Implement Procedure 12004-C To Secure From Long-Cycle
Recirculation" - paragraphs 3 and 4.b(3)(s)
-
" Failure To Establish An Appropriate Procedure To Maintain SI .
Operable While Filling Accumulators" - paragraph 4.b(3)(q) !
- " Failure to establish adequate procedures to ensure AMSAC was
available" - paragraph 5 f
IFI 50-424/89-14-02 and 50-425/89-15-02, " Review Licensee Evaluation
Regarding Adjustment Of The P-9 Setpoint When Steam Dumps Are Removed From
Service" - paragraph 3
IFI 50-425/89-15-03, " Verify Resolution Of Restoring The SSMP To A
Condition To Correctly Indicate The Operability Status" - paragraph 5.d .
I
NCV 50-424/89-14-03, " Failure To Perform Required Testing Per I
Surveillance Requirements Results In TS 4.3.3.10 Violations - LER
50-424/89-06" - paragraph 4.b(2)(a)
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NCV 50-424/89-14-04, " Failure To.Take Required Temperatures Results In
Inadequately Performed . Surveillance Resulting In A TS Violation - LER
50-424/89-07" - paragraph 4.b(2)(b)
NCV 50-424/89-14-05, " Failure To Conduct An Adequate Engineering' Review
Re
Of
TS The AFW
3.7.1.2 Electrical
Violation - LER System Which Led
50-424/89-08" To AFW Inoperability(c)sulting
- paragraph 4.b(2) In a
NCV 50-424/89-14-06 " Failure To Follow Procedures While Conducting A
Liquid Waste Release Resulting In A TS 3.3.3.9 Violation -
LER
50-424/89-10" - paragraph 4.b(2)(d)
NCV 50-424/89-14-07, " Failure To Conduct Surveillance Resulting In A
Violation Of TS 4.1.3.2 - LER 50-424/88-30" - paragraph 4.b(3)(h)
NCV 50-425/89-15-04, " Failure To Meet A Mode Change Prerequisite
Resulting In A TS 3.7.12. Violation Requiring Valve 2HV-19051 To Be
Operable Prior To Entering Mode 4 - LER 50-425/89-05" - paragraph
4.b(2)(e)
NCV 50-425/89-15-05, " Failure To Follow Procedures Resulting In
Inadvertent SI Actuation - LER 50-425/89-06" - paragraph 4.b(2)(f)
NCV 50-425/89-15-06, " Failure To Establish An Adequate Sampling Procedure
For Diesel Fuel Oil Per TS 6.7.1.a - LER 50-425/89-09" - paragraph
4.b(2)(h)
NCV 50-425/89-15-07, " Failure To Obtain A Radioactive Release Permit
Prior To Releasing Radioactive Materials To Unrestricted Areas Resulting
In A TS 3/4.11.1 Violation - LER 50-425/89-?0" - paragraph 4.b(2)(1)
'
NCV 50-425/89-15-08, " Failure To Follow Procedures While Performing
Maintenance On 2RE-2562A Resulting In The Plant Operating In A Condition
Prohibited By TS Thus Requiring Entry Into TS 3.0.3 - LER 50-425/89-12" -
paragraph 4.b(2)(j)
NCV 50-425/89-15-09, " Failure To Maintain The Auxiliary Feedwater System
Operable Resulting In A Condition Prohibited By TS 3.7.1.2. - LER
50-425/89-13" - paragraph 4.b(2)(k)
NCV 50-425/89-15-10 " Failure To Maintain RMWST, Discharge Valves Shut
Closed And Secured In Position While In Mode 5 Resulting In TS 3.4.1.4.2
Violation - LER 50-425/89-02" - paragraph 4.b(3)(m)
NCV 50-425/89-15-11 " Failure To Exercise The Duties And Responsibilities
Of The R0 And SS As Delineated In Operations Procedure 10000-C - LER
50-425/89-08" - paragraph 4.b(3)(p)
The strengths in the areas of maintenance (paragraph 2.b(7)) and startup
testing (paragraph 3) and the weakness in the area of operations
(paragraphs 3, 4.b(2), and 4.b(3)) were also discussed.
I
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7. Acronyms And Initialism l
!
ABN As-Built Notice
A/DV Anchor Darling Valve
AFW Auxiliary Feedwater System
AMSAC ATWAS Mitigating System Actuating Circuitry
ASTEC Automatic Surveillance Technical System
BFIV Bypass Feed Isolation Valve
BFRV Bypass Feed Regulation Valve
B0P Balance-of-Plant
CCP Centrifugal Charging Pump
CCW Component Cooling Water System
CFR Code.of Federal Regulations
CRI Control Room Isolation
CVCS Chemical & Volume Control System
CVI Containment Ventilation Isolation
DC Deficiency Cards
DF0S Diesel Fuel Oil Storage
DPIS Digital Position Indication System
DRPIS Digital Rod Position Indication System
ECCS Emergency Core Cooling System
ERF Emergency Response Facility
ESF Engineered Safety Feature
FI Flow Indicator
FWI Feedwater Isolation
GE General Electric ,
GPM Gallons Per Minute
HS Hand Switch
HV High Voltage
I&C Instrument and Control
IFI Inspector Followup Item
ISEG Independent Safety Engineering Group
LC0 Limiting Condition for Operation
LER Licensee Event Reports
LLRT Local Leak Rate Test
LOSP Loss of Offsite Power
MDAFW Motor Driven Auxiliary Feedwater System Pump
MDD Minor Departure from Design
MFIV Main Feedwater Isolation Valve
MFP Main Feed Pump
M0V Motor Operatcc Valve
MWO Maintenance Work Order
NCV Non-cited Violation
NPF Nuclear Power Facility
, NR Narrow Range
NRC Nuclear Regulatory Commission i
NSCW Nuclear Service Cooling Water i
i NUE Notice of Unusual Event l
0S05 On-Shift Operation Supervisor
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35'
PERMS Plant Effluent Radiation Monitoring System
PORV Power Operated Relief Valve
PT Pressure Transmitter
PV Pressure Valve
.RCDT Reactor Coolant Drain Tank
RHR Residual Heat Removal System
RMWST- Reactor Makeup Water Storage Tank
R0 Reactor Operator
RPDM . Rod Position Deviation Monitor
RWST Reactor Water Storage Tank
SAER Safety Audit and Engineering Review
SGWLC Steam Generator Water Level Control
SI Safety Injection System
SS Shift Supervisor
SSMP Safety System Monitor Panel
TDAFW Turbine Driven Auxiliary Feedwater Pump
TS Technical Specification
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