IR 05000348/1994007
| ML20029E067 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 04/28/1994 |
| From: | Cantrell F, Morgan M, Ross T, Michael Scott NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20029E062 | List: |
| References | |
| 50-348-94-07, 50-348-94-7, 50-364-94-07, 50-364-94-7, NUDOCS 9405160200 | |
| Download: ML20029E067 (22) | |
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gg* "'%r,g UNITED STATES
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NUCLEAR REGULATORY COMMISSION y
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101 MARIETTA ST8EET, N.W., SUITE 2900
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o ATLANTA, GEORGIA 30323-0199
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p Report Nos.:
50-348/94-07 and 50-364/94-07 Licensee:
Southern Nuclear Operating Company, Inc.
P.O. Box 1295 Birmingham, AL 35201-1295 Docket Nos.:
50-348 and 50-364 License Nos.:
NPF-2 and NPF-8 Facility Name:
Farley Nuclear Plant, Units 1 and 2 Inspection Conducted: March 14 - April 10, 1994
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$r V/2S/94 Inspectors:
_T. M. Ross, Seador Resident Inspector Date Signed
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M. J. Morgan, Resident Inspector
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'IlMl1t M. A. Sco t,Jesident Inspector
' Date Signed V28/91 Approved by:
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Ddte Signed M oy V. CantreT, Chief Reactof Projects Section IB Division of Reactor Projects SUMMARY Scope:
This routine resident inspection was conducted onsite in the principal areas.
e of plant operations, maintenance, surveillance, safety system verification, review of nonroutine events, and follow-up of previous inspection findings; and to a lesser degree, health physics, physical security, fire protection,
. engineering attributes, and technical support. Deep backshift inspections were conducted on March 16, 24, and 25, and April 6, 7, and 9, 1994.
Results, as summarized by SALP functional area:
0_perations In general, operations personnel performed very well_ in controlling plant outage and refueling activities in accordance with applicable plant procedures-and in compliance with Technical Specifications.
Licensed operators consistently demonstrated a high degree of knowledga and attentiveness.
However, an unresolved item was identified for poor communications. during the establishment and/or restoration of system conditions which resulted in several incidents (see paragraph 3.b.4).
These incidents were not considered typical of Shift Supervisor performance observed in the past. One violation 9405160200 940502 DR ADOCK 05000340 PDR
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was identified regarding inadequate administrative control of scaffolding in the vicinity of safety-related equipment (see paragraph 3.a.4).
Control of personnel overtime was evident and did not appear to be excessive.
Maintenance and Surveillance Maintenance personnel conducted assigned activities in accordance with applicable procedures. Mechanics and technicians demonstrated familiarity with administrative and radiological controls, and good craft skills. The failure of a power operated relief valve to operate identified inadequate torquing of fasteners during previous maintenance on the valve.
Plant management responsible for maintenance exhibited excellent initiative and problem solving capability in addressing a number of potentially serious equipment issues.
Surveillance testing by responsible personnel were consistently conducted in a methodical step-by-step manner in accordance with applicable test procedures.
Responsible test personnel were very knowledgeable of the details of assigned surveillance activities. One violation was identified regarding noncompliance with Technical Specifications during the performance of quarterly surveillance testing of nuclear instrumentation system power range channels (see paragraph 5.1).
Engineering and Technical Support Engineering and technical support from various organizations (i.e.,
Operations, Technical, Plant Modifications, Maintenance, Planning and Systems Performance) was consistently of a high caliber.
The overall planning, installation, testing, and problem resolution of plant modifications and other outage-related work were well supported.
No violations or deviations were
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identified in this area.
Plant Support Health physics (HP) personnel provided good support of outage-related activities.
The HP steam generator trailer was an excellent initiative that resulted in significant dose reductions.
Some minor lapses in foreign material control were observed (see paragraph 3.b.2).
Security personnel were consistently alert and appeared to be implementing the plant's security plan
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appropriately. The large influx of contractor personnel into the protected area was well controlled. Compensatory fire protection measures during the outage were effective, however, some minor problems regarding fire watches
were noted (see paragraph 7.a).
No violations or deviations were identified.
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I REPORT DETAILS 1.
Persons Contacted Licensee Employees
- W. Bayne, Safety Audit and Engineering Review Site Supervisor
- C, Buck, Technical Manager S. Casey, Systems Performance Supervisor
- R. Coleman, PMD Manager P. Crone, Instrumentation and Controls Superintendent L. Enfinger, Administrative Manager H. Garland, Mechanical Maintenance Superintendent
- R. Hill, General Manager - Farley Nuclear Plant J. Kale, Chemistry / Environment Superintendent
- J. McGowan, Safety Audit and Engineering Review Manager M. Mitchell, Health Physics Superintendent l
- C. Nesbitt, Operations Manager l
J. Odom, Superintendent Unit Operations J. Osterholtz, Assistant General Manager - Plant Support J. Powell, Superintendent Unit Operations
- L. Stinson, Assistant General Manager - Plant Operations
- J. Thomas, Maintenance Manager
- B. Yance, Systems Performance Manager
- L. Williams, Training Manager NRC Personnel
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S. Koenick, Intern l
M. Morgan, Resident inspector l
- T. Ross, Senior Resident Inspector
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- M. Scott, Resident Inspector
- Attended the exit interview Other licensee employees contacted included, health physics, operators, technical staff, security, maintenance, I&C and office personnel.
Acronyms used throughout this report are listed in the last paragraph.
2.
Farley Nuclear Plant (FNF; tatus o 4 Activities a.
The twelfth Unit I refueling outage (VIRF12) began on March 3, 1994, and continued throughout the inspection report period.
b.
Unit 2 has operated continuously since December 29, 1993, at approximately 99.5% power.
This unit operated at slightly less
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than full power due to administrative limits on Tavg (i.e.,
Average Tavg is being maintained at less than 575 degrees Fahrenheit).
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c, Other NRC inspections or meetings at the site 1)
During the week March 14, Region Il conducted an Emergency Preparedness followup inspection that will be documented in NRC inspection report (IR) 50-348, 364/94-05.
2) During the week of March 14, Region II conducted a routine Security inspection that will be documented in IR 50-348, 364/94-06.
3) During the weeks of March 21 and 28, Region II conducted an inspection of Unit 1 steam generator (SG) eddy-current testing and repairs, and other inservice inspection practices, that will be documented in IR 50-348, 364/94-09.
4) The responsible Region Il Section Chief, Floyd Cantrell, was on site March 21 and 22, to review resident inspector activities and meet with senior plant management.
3.
Review of Plant Operations (71707) and Refueling (60710)
a.
Plant Tours Routine plant tours, particularly of the control room and the auxiliary building, were performed to verify that operating license and regulatory requirements were being met.
In general, inspectors looked for the existence of unusual fluid leaks, piping vibrations, pipe hanger / seismic restraint settings, valve and breaker positions, equipment caution / danger tags, material and equipment conditions, overall housekeeping, fire protection features, and instrument calibration dates.
Tours were conducted both on dayshift and backshifts.
1) Walkdowns of Safety-Related Equipment / Areas Limited walkdowns of accessible portions of the following i
safety-related systems and surrounding areas were performed:
m Unit 1 and 2 AFW Pump Rooms a Unit I and 2 4160 VAC Switchgear Rooms a Unit 1 Main Steam Valve Room "B" Bay a Unit 1 and 2 CCW Heat Exchanger and Pump Rooms a Unit 1 and 2 RHR Heat Exchanger and Pump Rooms m Unit 1 Containment Spra~y Pump Rooms
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Unit I and 2 Cable Spreading Rooms s
a Unit 1 and 2. Vital 4KV Switchgear Rooms a Unit 1 and 2 Charging Pump Rooms Breaker / switch / valve line-ups (both locally and in the control room), equipment conditions, and housekeeping were examined.
System lineups were verified to meet operability requirements.
Safety-related equipment material conditions-
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and area housekeeping for Unit 2 looked very good.
Unit 1 material conditions and housekeeping were generally well controlled considering outage activities.
2)
Tagout/ Clearances During the inspection period, the following tagging orders /
clearances were reviewed and verified to be properly implemented:
a 94-1832-1, Main Feedwater (MFW) and Auxiliary FW valves to the B SG u
94-1159-1. 18 Battery isolation for testing a
94-0092-1, 18 Diesel Generator (DG)
3) Walkdown of Unit 1 Main Steam Valve Room - B Bay (71710)
The Main Steam (MS) Valve Room is an interface area between containment and the turbine building.
It contains some of the piping and major components for the safety-related portions of the MS and MFW Systems, and includes the injection portion of the Auxiliary Feedwater (AFW) System.
An inspector conducted a detailed physical examination of the "B" Bay.
In general, the following attributes were checked - mechanical integrity of components; clearances in place for components being worked; components and piping protected or sealed to prevent foreign material entry; major supports and attachments in place; major physical components matched P& ids; and, defects or damage not being
worked were identified by deficiency tags or equivalent.
The applicable P& ids for these systems were D-175007, Rev 19
[AFW); D-175033, Rev 16 [MS]; and D-170117, Rev 17 [MFW).
Aside from the normal activities associated with an outage and some minor housekeeping problems at the floor level, the portions of these systems located in the B Bay of the MS Valve Room were found to be satisfactory condition and consistent with official P& ids.
4) Walkdown of Unit 1 Containment - Scaffolding Deficiencies The inspectors conducted several general area walkdowns of the Unit 1 containment during the course of the twelfth refueling outage (UlRF12).
Every level of containment was inspected for cleanliness, equipment material conditions, and safety system integrity.
The inspectors were especially sensitive to any signs of poor work practices and/or leakage.
Overall interior conditions of the Unit I containment were found to be acceptable.
The inspectors did not identify any circumstances that would call into question
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safety system operability or inadequate work practices, except as noted below.
During a containment tour on March 24, an inspector observed that a number scaffold permits for scaffolds in the vicinity of safety-related equipment were incomplete.
More specifically, six scaffolds were in use near safety-related equipment that had not been approved for use by the Shift Foreman Operating (SF0) as required by Step 7.10 of FNP-0-GMP-60, Revision 10, " General Guidelines And Precautions For Erecting Scaffolding." These scaffolds constituted approximately one third of the major equipment scaffolds examined by the inspector.
In addition to these scaffolds, subsequent tours of the Unit 1 Auxiliary Building and DG Building identified two other scaffolds in use near safety-related equipment with incomplete permits (i.e., not signed by the SFO).
The eight deficient scaffolds were brought to the Shift Supervisor's (SS) attention and promptly corrected.
Similar problems with incomplete scaffold permits were observed during the ninth Unit 2 refueling outage (U2RF9).
These prior' problems were identified as non-cited violation (NCV) 93-28-02 in NRC inspection report 50-364/93-28.
The licensee's failure to effect adequate longterm corrective actions to resolve continuing difficulties with implementing FNP-0-GMP-60 is considered a violation (VIO), and identified as VIO 50-348/94-07-01, " Unapproved Scaffolds Near Safety-
Related Equipment."
b.
Routine Plant Operations Review The inspection staff periodically reviewed shift logs and plant operating records including instrument traces, chemistry reports, auxiliary logs, operating / standing o.rders, night order entries, and equipment tagout records.
Inspectors routinely monitored operator alertness / demeanor, control room staffing and access, shift turnovers, and operator performance during routine operations.
Random off-hours inspections were conducted to ensure that operations and security performance remained at acceptable j
levels.
Control room annunciator status and alarms were reviewed.
1) Technical Specifications Compliance FNP compliance with selected Technical Specifications (TS)
Limiting Condition of Operation (LCO) were verified throughout the inspection report period by the inspection staff. During this period the inspectors did not identify any circumstances constituting a possible noncompliance of Technical Specifications (TS), with the exception of two violations and an unresolved item described elsewhere.
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2) Unit 1 Core Refueling Activities The inspectors observed refueling activities for selected fuel assemblies during both unloading and reloading of the Unit I reactor core.
Unloading of the Unit I reactor core was completed on March 18.
Reloading the core with cycle 13 fuel was completed on April 4.
All refueling activities witnessed by the inspectors were conducted in strict compliance with the governing procedure FP-ALA-R12,
" Refueling Procedure J.M. Farley, Unit No.1 Nuclear Plant Cycle XII - XIII." The inspectors monitored refueling activities from the control room, spent fuel pool (SFP), and reactor cavity. At FNP fuel handling manipulations are performed by a contractor under the direct oversight of licensee personnel.
During both refueling evolutions, the inspectors confirmed that procedural prerequisites and instructions were signed off appropriately. The inspectors also independently verified a selected number of prerequisites, with particular attention paid to those involving compliance with TS Section 3/4.9, " Refueling Operations."
Following completion of core
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reload, the inspectors independently compared the final core map specified in FP-ALA-R12 against the licensee's final core reload video tape. No discrepancies were identified in the final Unit I reactor core configuration.
In general., refueling personnel performed extremely well.
Refueling operations were conducted in a controlled and methodical manner without incident. The only adverse finding identified by the inspectors involved implementation of the foreign material exclusion (FME) program in the SFP
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area.
Instances of inadequate accountability of personnel and material into and out of the FME controlled area immediately around the pool were brought to the attention of
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plant management. These were promptly addressed.
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3) Unit 1 Fuel Sipping Incident On March 18, during fuel sipping operations in the Unit 1 SFP, a SFP bridge crane operator inadvertently tilted fuel assembly 2830 a few degrees from vertical while attempting to remove-it from. sipping can #2.
This fuel assembly had.
not cleared the sipping can, before the operator tried to'
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move laterally. The bridge operator recognized almost
.H immediately that fuel assembly 2B30 was not clear of the can and promptly-secured bridge travel. After raising the I
assembly a few more inches, he was then able to place it in the designated SFP storage location.
Subsequent
~ investigation by the licensee determined that the hoist motor limit switches were not set properly to accommodate i
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the additional height of the sipping cans which project about six inches above the SFP racks.
The licensee's initial corrective actions to establish a fuel movement spotter in the SFP area during sipping operations, readjust the SFP hoist limit settings, and counsel responsible personnel were considered adequate. The resident staff will followup on the licensee's ongoing investigation to ascertain how the limit switches were improperly set initially. This will be identified as inspector followup item (IFI) 50-348/94-07-02, "SFP Hoist Limit Switch Misadjustment."
4)
Establishing System Conditions - Poor Communications Between March 31st and April 2nd, the licensee's operations staff for Unit I was responsible for three separate incidents involving inadequate communications. All of these incidents occurred as a consequence of not establishing proper system / component conditions prior to testing or returning to service. A brief summary of these incidents is as follows:
a.
On March 31, Operations prematurely reenergized the 208 VAC section of Motor Control Center (MCC) 1G, before the completion of post-modification testing (PMT), inorder to return the "B" Train Control Room Ventilation System (CRVS) to service.
About six hours later, operators realized that several fans powered by MCC 1G were rotating backwards. A subsequent investigation determined that during prior electrical modifications. of MCC.lG, _ certain power leads had been inadvertently reversed.
b.
On April 2, Operations failed to properly align the 1A Containment Spray (CS) system during preparations to perform FNP-1-STP 16.10, " Containment Spray System
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Check Valves and Pump Flow Test - A Train." The subsequent start of the 1A CS pump per STP-16.10, resulted in an uncontrolled discharge of Refueling Water Storage Tank.(RWST) water into containment and the 1A CS pump room. Three individuals were sprayed with water and received minor contamination on their shoes and clothes.
A licensee review of active tagging orders disclosed that the isolation valves for two test connections on the discharge side of the 1A CS pump had been left oper,.
c.
On April 2, Operations attempted to perform surveillance testing on the containment sump to Residual Heat Removal- (RHR) system interlock (i.e.,
FNP-1-STP-11.13) without understanding the status of
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other solid state protection system (SSPS) tests.
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Instrumentation and Control (I&C) technicians had l
previously installed temporary test equipment in the SSPS for time response testing.
When the SSPS mode selector switch was placed in "0PERATE" per STP-11.13, several CS relays in the SSPS were actuated due to electrical short circuiting via I&C test equipment.
However, because the majority of the CS system was already tagged out, only the CS Additive Tank Outlet Valve stroked open.
The resident inspector staff is continuing to review the licensee's investigation process and corrective actions concerning these three incidents.
This is considered an unresolved issue (URI), and identified as 50-348/94-07-03,
" Improper Release of Equipment."
5)
Licensee Use Of Overtime An inspector examined the licensee's use and control of overtime during UlRF12, for both units. At FNP, plant management routinely requests its staff to work regular 12-hour shifts during refueling outages.
The inspector reviewed overtime records and discussed control of overtime with site senior management.
The results of this inspection are as follows:
The Systems Performance (SP) group recorded the a
highest average number of overtime hours (i.e., 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of overtime per person per week).
Health Physics and Maintenance groups recorded the m
second and third highest average numbers of overtime hours (i.e., 20 and 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> of overtime per person per week, respectively).
m The maximum number of hours any individual can work during a week is administratively limited to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
The inspector verified this limit has not been exceeded.
m Plant management has not given routine approval for exceeding the overtime limit guidelines specified in the TS; during the last two outages, approval has been given on only four separate occasions.
m The overtime trend has been decreasing for all plant work groups over past two years.
In general, operations personnel performed very well in controlling plant outage and refueling activities in accordance with applicable plant procedures and in compliance with Technical Specifications.
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Licensed operators consistently demonstrated a high degree of knowledge and attentiveness.
However, poor communications during the establishment and/or restoration of system conditions resulted in several incidents.
These incidents were not considered typical of Shift Supervisor performance observed in the past. One violation was identified regarding inadequate administrative control of scaffolding in the vicinity of safety-related equipment.
4.
Maintenance Observation (62703)
The inspectors observed / reviewed portions of various predictive, preventative and corrective _ maintenance activities to determine conformance with facility procedures, work requests and NRC regulatory requirements. Work requests and instructions were also evaluated to-determine the status of outstanding jobs and to ensure that proper-priority was assigned to safety-related equipment.
a.
WA-W00407305, W00407298, and W00407306; Remove /Re-installation of HFA Relays for Calibration from the BlG Load Sequencer The above work authorizations provided instructions for the removal, calibration, and re-installation of HFA relays 68G1, 27XG, and 68G2, respectively.
The inspector observed the removal of these relays by electrical maintenance (EM) personnel.
Relay leads were verified and marked to facilitate proper re-installation. The electricians paid special attention to the location of leads from relay 68G2 to a diode mounted on the relay box to ensure there would be no confusion during its replacement.
Documentation and control of work was good, b.
WO-360015; 1A Component Cooling Water (CCW) Pump Room Cooler Replacement Contract construction personnel, using work order instructions and design change information, replaced the 1A CCW Pump Room cooler.
An inspector observed the installation of a new, increased capacity cooler unit to replace the old one as part of PCN 890-1-6986. Workmanship of the newly installed unit looked. good.
Appropriate care was taken during rigging operations of this 4,000'
pound cooler unit, such that no damage occurred to equipment in the immediate area.
Supports and other structures removed for this work were properly re-installed.
c.
4160 VAC Breaker Prop Latch, Springs - Corrective Actions The FNP EM organization has completed its evaluation of past performance problems involving prop latch springs in Allis-Siemens 4160 VAC breakers _(see IR 94-04).
The inspectors have reviewed and discussed the results of this evaluation with the licensee, and considered their conclusions and corrective actions acceptable. A detailed description of the prop latch. spring issue and proposed resolution is documented in a licensee memorandum
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ENG-94-0487, "REA 94-0487 Siemens MA350C Breaker Prop Latch Spring Evaluation," dated March 8,1994.
In summary, the licensee's corrective action program involves additional inspections, and repairs as necessary, of 4160 VAC Allis-Siemens breakers, to ensure the following critical characteristics:
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Proper hinge pin alignment and retaining screws are secured.
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Only spring clips with the 45 degree bend are used to attach prop latch springs.
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Proper orientation of prop latch spring attachment.
The EM organization conferred with the breaker vendor regarding its corrective actions.
They also plan to revise applicable procedures to ensure adequate instructions exist for future use.
The licensee. plans to inspect, and repair all Unit 1 breakers during the present UlRF12 outage.
In addition, the licensee intends to address the breakers on Unit 2.
These breakers were inspected for hinge pin integrity during the recent Unit 2 refueling outage (U2RF9); however, the critical characteristics identified above were not fully understood.
Consequently, these breakers will be reinspected as they become available during routine PMs and associated component surveillance outages over the next several months.
The residents considers this proposed action plan reasonable.
d.
WO-500557; 2C DG Fuel Injector and Mechanical Component Leak Check An inspector observed performance of this leak inspection by mechanical maintenance (MM) personnel during start of the 2C DG for monthly surveillance testing (see paragraph 5.c.)
No leaks were identified, e.
Unit 1 Reactor Trip Bypass Breaker A - Maintenance and Testing Reactor trip and bypass circuit breakers are inspected, cleaned, and tested every refueling outage. An inspector observed the performance of selected portions of these activities for the "A" Bypass Breaker.
These activities were accomplished by skilled electricians in accordance with the instructions of Electrical Maintenance Procedure (EMP) 1402.1.
Visual inspections and performance tests of the shunt trip actuator (STA) and undervoltage trip actuator (UVTA) were completed with satisfactory
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results.
Bypass breaker interlock tests, and power supply and control circuit insulation resistance tests were also completed'
satisfactorily.
Although, the "A" Bypass breaker UVTA met its acceptance criteria the inspector identified a concern that current and past test results indicated a negative trend.
In fact, an extrapolation of
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l the results seemed to suggest that the UVTAs for certain reactor trip and bypass breakers could be below the acceptable voltage drop-out limit before the next outage.
In response to the inspector's concern, the UVTAs for all four Unit 1 breakers were retested by plant electricians using more sophisticated equipment.
Subsequent test results convinced FNP management to replace all UVTAs prior to restart of Unit 1 as a preemptive measure. The inspectors will also review the performance history of Unit 2 UVTAs.
f.
MWR-278141; Unit 1 Printed Circuit Card Fuse Inspections In response to a vendor infogram, the licensee has been checking the ratings of fuses installed on 7300 series printed circuit cards in the Unit I and Unit 2 protection cabinets.
An inspector observed I&C technicians remove, examine, and reinstall several of these circuit cards from Cabinet No. 2 (Protection II) of Unit 1.
The licensee had originally planned to examine just 64 circuit cards for Unit 1.
However, due to the number of failures, the sample size was expanded to the entire population of over one-thousand cards.
[ Note, a failure was any over-sized fuse].
During the Unit 1 examinations all improperly rated fuses (both under-sized and over-sized) were replaced.
g.
Power-0perated Relief Valve (PORV) Failure and Repair On March 7th, during the shutdown of Unit 1, operator attempts to open pressurizer power-operated relief valve (PORV) PCV-445A, failed while trying to establish a vent path for the reactor coolant system (RCS).
The other pressurizer PORV, PCV-4448, performed properly.
Subsequent investigation of the problem revealed tho? the air dome fasteners on the valve actuator had not been torqued to the correct value.
In October 1992, when these fasteners were last torqued, the responsible I&C technician selected the wrong (i.e.,
lower] torque value for the specific fasteners being us,ed. As a result, air to the PCV-445A actuator leaked from the dome [where the diaphragm cover meets the diaphragm at the dome bolt circle]
preventing the valve from operating.
Although PCV-445A satisfactorily passed its last stroke test on March 11, 1993, the licensee now believes that relaxation of the fasteners from the lower torque value became sufficient to cause excessive air leakage.
The condition and records of all other similar air-operated valves (A0V) in Unit I were checked by the licensee.
Only one other actuator was found to have incorrectly torqued dome fasteners.
This actuator was on PCV-444C, a pressurizer spray valve, which had already been declared inoperable on March 6 (shortly after Unit I was shutdown) because it would not stroke.
PCV-444C has the same type and size of actuator as the PORV PCV-445A,.but had
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failed due to clogging of the air actuation line filter - not diaphragm leakage. The other spray valve operated satisfactorily.
As longterm corrective action, the torquing requirements contained in the applicable FNP I&C preventive maintenance procedure for PCV-445A type of actuators was clarified to only list the torque value for their specific size of dome fasteners.
The licensee-also reviewed maintenance records for the Unit 2 PORVs, and other similar A0Vs, and determined that all Unit 2 actuators were torqued correctly.
These corrective actions were considered reasonable and timely.
No violations or deviations were identified in this area.
Maintenance personnel conducted assigned activities in accordance with applicable
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procedures. Mechanics and technicians demonstrated familiarity with administrative and radiological controls, and good craft skills.
Pl ant management responsible for maintenance exhibited excellent initiative and problem solving capability in addressing a number of potentially serious equipment issues. The inadequately torqued fasteners for a PORV actuator is an example of management resolving an equipment issue.
5.
Surveillance Observation (61726)
Inspectors witnessed surveillance test activities performed on safety-related systems and components, in order to verify that such activities were performed in accordance with facility procedures, TS and NRC regulatory requirements.
Portions of the following surveillance tests were observed:
a.
FNP-1-STP-905.1, Auxiliary Building Battery Load Profile Test, 18 Battery An inspector observed the performance of STP-905.1 under work authorization WOO 411670 with clearance 94-1159-1 in effect.
Responsible electricians were familiar with the details of the procedure and followed them accordingly. Test data was verified to meet STP acceptance criteria.
This battery test demonstrated that the battery had a large amount of margin.
Test equipment used was within calibration due dates.
Although, the IB battery had been replaced during the last Unit 1 outage, the inspector saw some brown corrosion buildup on several of the cell posts. After STP 905.1 was completed, electricians were to conduct routine preventative maintenance (PM) on the 1B battery.
Among other things, this PM activity breaks down battery components to clean battery leads, cell posts, and connectors.
Subsequent observations by the inspector verified that the corrosion was removed.
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b.
FNF-0-STP-80.11, Diesel Generator 1-2A 1200 KW Load Rejection Test An inspector observed the satisfactory performance of STP-80.ll by reactor operators in the control room. The 1-2A DG adequately picked up the rejected electrical loads when offsite power to the IF 4160V bus was intentionally interrupted (i.e., startup transformer output breaker manually opened].
The maximum DG voltage rise achieved was well below the acceptance criteria limit and generator output frequency drift was minimal.
During pretest checks, the licensed operator-performing the test identified an error in the test procedure.
The wrong output breaker for the startup transformer was specified. The Shift Supervisor promptly acknowledged the problem and had a procedure change initiated prior to proceeding with the test.
c.
FNP-0-STP-80.17, DG 2C Operability Test While in the DG building, an inspector observed the successful slow start and load test of the 2C DG. The responsible system operator (S0) followed FNP-0-SOP-38.0, Diesel Generators, in an appropriate manner. Mechanical maintenance was on hand for routine leak checks (see paragraph 4.d).
d.
FHP-0-ETP-3637, Reactor Core Television Mapping An inspector conducted an independent review of the final core map video tape made after the Unit I reactor core was fully loaded.
This video tape was cross-checked against the map that was generated of the loaded core.
The generated core map was also compared to the Westinghouse final core load configuration depicted in FP-ALA-R12, Rev 0, TCN 0, on page 270. The video tape was clear, methodical, and fuel assembly identification numbers and geographical -locators were plainly visible. The inspector did not identify any discrepancies, e.
FNP-1-STP-40.7, ECCS Branch Line Flow Verification and Charging Pump Low Discharge Head Flow Test STP-40.7 was observed by an inspector from the Unit I control room where the test was being directed by an off-shift senior reactor operator (SRO) from the Operations technical support. staff.
Flow values were collected at the individual flow orifices at the RCS
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injection paths in containment.
This information was relayed to the operator-at-the-controls in the control room by phone.
Throttled positions were adjusted slightly on four of the injection valves to bring them into specification.
During performance of the test, one flow cell / meter drifted out of i
calibration [ read 0.5 psid high].
This was quickly identified by i
the I&C personnel in containment and relayed to the control room.
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A-new flow cell / meter was obtained and that portion of the test was re-run satisfactorily, f.
FNP-1-STP-16.12, Containment Spray Pumps Automatic Starting Circuitry Test j
The purpose of STP-16.12 is to test the Unit 1 CS pump emergency start circuits from the BlF and B1G sequencers on a phase B
isolation and SI signal simulation.
During this test CS spray pump breakers are placed in the test position to prevent actual start of the 1A and IB CS pumps.
Both CS pump breakers performed satisfactorily (i.e., closed) upon initiation of the simulated signal.
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FNP-1-STP-256.16, BlF Diesel Generator Loading Sequencer Time Response Testing An inspector observed the conduct of STP-256.16 and reviewed the associated test data.
The sequenced Engineered _ Safety System (ESS) actuation response times were all within allowed acceptance criteria.
Plant electricians performed this test in an orderly step-by-step fashion, including restoration of the BIF DG Loading Sequencer cabinet (e.g., removal of jumpers and opening links),
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FNP-1-PMP-1195,
"A" Train Lose of Offsite Power Test With IC Diesel Generator
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An inspector observed licensed operators start and run the IC DG in accordance with PMP-1195.
The purpose of.this test was to-
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verify IC DG performance following UlRF12 modifications.
Both the DG and operations shift personnel performed satisfactorily, i.
FNP-1-STP-40.1,
"A" Train Sequencer Operability and Load Shedding Circuit Test
The purpose of this STP was to demonstrate the operability of the following circuits - Load shed; Safety injection (SI) with offsite power; SI with loss of offsite power (LOSP); and LOSP. An inspector observed the conduct of this test by operations and EM personnel in strict accordance with STP-40.1.
Responsible test personnel and plant equipment performed satisfactorily. All load shed and sequencer circuits met their test acceptance criteria.
j.
FNP-RCP-271, Isotopic Calibration of Unit 1 Monitors An inspector observed the calibration of the high voltage setting of the RE-21 vent stack air particulate monitor in accordance with RCP-271. A health physics technician used three sealed isotopic sources (i.e., Ba-133, Cs-137, and Co-60) at the detector while an I&C technician adjusted detector high voltage at the drawer.
However, repeated attempts by I&C to calibrate RE-21 failed to align the' drawer within accepted tolerances for the respective
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The RE-21 detector was declared dysfunctional and subsequently replaced.
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FNP-1-STP-40.0, Safety Injection with Loss Offsite Power Test
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The purpose of STP-40.0 is to demonstrate the capability of engineered safety systems and components following a manual initiation of LOSP and SI. During this test, DGs. start, loads strip, ESS loads sequence on, emergency core cooling systems (ECCS) inject to the reactor vessel, and containment is isolated (Phase A). An inspector attended-the pretest briefing of STP-40.0, verified selected initial conditions, and observed the successful conduct of this comprehensive, integrated test.
Operations and I&C personnel followed STP-40.0 step-by-step in a controlled and orderly manner. All plant equipment performed as designed, except for two valves (i.e., MOV 3134 and HV 3659)
previously identified as test exceptions; MOV 518/536, "SW to/from the DG building - B Header," which failed to open; and the 1A RHR pump which did.not automatically start and sequence on.
Shortly after being identified during the test, the 1A RHR was started manually and MOV 518/536 was opened. The licensee is currently investigating these failures to determine their cause and effect corrective actions.
The resident inspectors plan to follow these efforts.
1.
Unit I and 2 Nuclear Instrumentation System (NIS) Power Range (PR)
Channel Quarterly Surveillance During power operations, the licensee has been fulfilling the quarterly surveillance requirement of TS Table 4.3-1 for each NIS PR channel by calibrating the entire channel.
[ Note, a complete channel calibration includes all bistables, reactor trip system interlocks, local and remote indicators, and associated trend recorders]. However, TS Table 4.3-1 only requires a quarterly channel calibration of the Power Range Neutron Flux - High Setpoint while in Modes 1 and 2.
A complete NIS PR channel calibration is only required by TS on a refueling outage basis.
NIS PR channels at FNP were designed to accommodate TS required surveillance testing at power- (i.e., Neutron Flux - High Setpoint)
without removing the channels from service.-[See UFSAR Sections 7.2.2.2.1.F and.7.2.3.2 for description of at power testing-of NIS PR channels].
But, inorder to accomplish a complete end-to-end channel calibration (excluding detectors) in 'accordance with plant surveillance test procedures (STPs), the associated NIS detector input cables were disconnected. This effectively rendered the NIS PR channel inoperable for the duration of the calibration (usually 10 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />).
Furthermore, while the channel was being calibrated it was not being placed in a tripped condition as required by Action Statement 2.a of TS Table 3.3-1 because tripping the channel was inconsistent with applicable STPs. This-
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practice was previously questioned by an inspector and identified
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as URI 50-364/94-02-01.
In 1983, the question of TS compliance during surveillance testing of the NIS channels was raised at the site. At that time it was concluded by the licensee that removing a channel from service for surveillance testing did not constitute inoperability. This decision appears to have established the foundation for the licensee's current practice.
No evidence was found to indicate that SNC has ever evaluated the safety consequences (i.e.,
10CFR50.59 or otherwise) of performing at-power testing of NIS PR channels using a method which deviates from that described in the UFSAR.
For the past 11 years, the licensee's standard methodology for performing quarterly surveillance tests of Nuclear Instrumentation System (NIS) Power Range (PR) channels has been inconsistent with the test methodology described in the UFSAR and has not complied with applicable TS.
This is considered a violation, and is i
identified as VIO 50-348, 364/93-07-04, "NIS PR Channel Inoperability."
Plant personnel conducted assigned surveillance activities in a methodical step-by-step manner accordance with applicable test procedures.
Control room operators were routinely kept informed of these activities.
Responsible test personnel were very knowledgeable of the details of assigned surveillance activities. One violation was identified regarding noncompliance with Technical Specifications during the performance of quarterly surveillance testing of nuclear instrumentation system power range channels.
6.
Engineering and Technical Support Engineering provided timely support during the implementation of PCN 890-1-6986 discussed in the maintenance section above.
Furthermore, the positive contribution by technical support and engineering staff from i
various organizations' (i.e., Operations, Technical, Plant Modifications,
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Maintenance, Planning and Systems Performance) during UlRF12 was clearly evident.
i Portions of the following PCNs were reviewed and/or observed in the field by the resident staff:
a.
PCN 89-1-5950, Replacement of Station Service Air Compressors Eventually, this PCN will replace all four existing instrument and service air compressors with newer, potentially more reliable units.
During UlRF12, the service water supply which' cools the compressors was enlarged (two inch pipe replaced with three inch),
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certain obsolete components were removed, and other minor preparations were pre-installed. MWR 258455 was the major work
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control document.
An inspector reviewed the changes to valve
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alignment checklists made by the Operations technical support staff.
These checklists were revised to accurately reflect the addition of newly installed valves and deleted references to removed valves.
b, PCN B90-1-7661, RHR Pump Motor Shaft Coupling Installation
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This PCN modified the 1A RHR pump by adding a coupling between the motor and pump to facilitate maintenance of the pump and/or motor.
Wiring to the pump motor, cooling piping to the pump, and some instrumentation were also modified as a part of the overall work.
The main controlling administrative document was Maintenance Support Request 60001.
IR 50-348/94-09 also discusses aspects of this modification.
The inspectors observed portions of in-process work, reviewed PMT data, and monitored pump operation intermittently over several weeks. An inspectors also reviewed an evaluation of the pump performance PMT data by SP engineers. This evaluation determined that the 1A RHR pump met required performance standards. The inspector concluded that SP's evaluation was consistent with actual test results.
The 1A RHR pump has run well in long continuous service since the modification.
c.
PCN S92-1-8234, Replace Portion of Service Water Cross-Tie Piping to AFW [ Service Water to AFW up to Q1N23V014A]
This PCN replaced a section of carbon steel SW piping, that had become partially clogged by sediment, with stainless steel piping that included a clean out flange.
The inspector observed pipe replacement construction activities.
Although, the old pipe was
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partially occluded with harden sediment it still had sufficient open passage to deliver emergency SW to the lA AFW pump. The
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replacement piping was well built and free of visual defects.
While the section of old piping was removed, pipir.g up stream of
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the removed section was hydrlazed with good results. Although the 1A AFW pump was not flow tested with SW (due to chemistry considerations), it did function properly with the newly installed piping.
Engineering and technical support from various organizations (i.e.,
Operations, Technical, Plant Modifications, Maintenance, Planning and Systems Performance) was consistently of a high caliber. The overall-planning, installation, testing, 'and problem resolution of plant modifications and other outage-related work were very effective.
No violations or deviations were identified in this area.
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Plant Support i
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Fire Protection Review (64704)
During the course of their normal tours, the inspectors routinely examined aspects of the Fire Protection Program such as transient fire loads, flammable materials storage, fire brigade readiness, ignition source / fire risk reduction efforts, and fire protection features.
In particular, the ins.oectors verified that appropriate fire watches were in place to compensate for fire protection features removed from service during UlRF12.
In general, the use of fire watches observed by the inspectors conformed with the site Fire Protection Program. However, two minor discrepancies were identified.
In one instance, two fire watches in the' Unit 1 and 2.
cable spreading rooms were armed with out-of-date fire extinguishers (i.e., the 30-day inspection period had elapsed a few days earlier).
In another, a fire watch for welding in the.lA RHR pump room, which was posted as a contaminated area, was not dressed out.
Both of these instances were brought to plant management attention and promptly corrected, b.
Physical Protection (81054)
The inspectors verified by observation during routine activities that security program plans were being implemented as evidenced by: proper display of picture badges; tours and stationing of security personnel; searching of packages and personnel at the plant entrance; and vital area portals being locked and alarmed.
Contractor access to the protected area during UlRF12 was controlled through the Secondary Access Point (SAP).
The SAP is usually opened only for outages or when the Primary Access Point.
(PAP) is not available.
Security personnel and equipment at the PAP and SAP performed well in accommodating the large influx of personnel into the protected area.
Licensee activities observed during the inspection period appeared.
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adequate to ensure physical protection of the plant.
Guards were alert and particularly attentive to open doors.
Their posted positions were well manned with frequent relief.
c.
Health Physics in general, health physics technicians demonstrated a continued vigilance for changing radiological conditions.
Their support and
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coverage of outage activities was exemplary.
In particular, the transport'of the contaminated 1B Reactor Coolant Pump stator from containment, and the_lA RHR pump motor from the radiologically controlled Auxiliary Building, to offsite locations were examples of well orchestrated evolutions.
Furthermore, the Health Physics SG Trailer was an excellent initiative by the licensee for providing remote access control and timely monitoring of personnel exposure during SG examinations and repair.
This initiative made
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j a significant contribution to SNC's aggressive dose reduction goals for UlRF12.
Health physics (HP) personnel provided strong support of outage-related l
activities.
The HP steam generator trailer was an excellent initiative that resulted in significant dose reductions.
However, some minor lapses in. foreign material control were observed (see paragraph 3.b.2).
Security personnel were consistently alert and appeared to be implementing the plant's security plan appropriately.
The large influx of contractor personnel into the protected area was well controlled.
Compensatory fire protection measures were effective during. Although, some minor findings regarding fire watches were made (see paragraph 7.a), no violations or deviations were identified.
8.
Followup of Open Inspection Items (92700)
i a.
(Closed) NCV 50-364/93-28-02, Incomplete Scaffold Permits This NCV is considered closed based on the issuance of VIO 94-07-01 (see paragraph 3.a.4).
b.
(Closad) URI 50-364/94-02-01, NIS PR Channel Inoperability This URI is now considered closed based on'the issuance of VIO 94-07-04 (see paragraph 5.1).
9.
Exit Interview Inspection scope / findings were summarized during management interviews throughout the report period and on April 14, with the general manager and selected members of his staff.
Inspection findings were discussed in detail and the licensee acknowledged these findings.
SNC did not identify as proprietary any material reviewed by the inspectors during this inspection.
ITEM NUMBER DESCRIPTION AND REFERENCE 50-348/94-07-01 (VIO)
Unapproved Scaffolds Near Safety-Related Equipment 50-348/94-07-02 (IFI)
SFP Hoist Limit Switch Misadjustment 50-348/91-07-03 (URI)
Improper Release of Equipment 50-348, 364/94-07-04 (VIO)
NIS PR Channel Inoperability 10.
Acronyms and Abbreviations
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Administrative Procedure ASME -
American Society of Mechanical Engineers (construction Code)
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Component Cooling Water
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Cathode Ray Tube DG
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Die.sel Generator DEH
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Digital-Electro-Hydraulic System (main turbine control)
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Division of Reactor Projects Division of Reactor Safety DRS
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DRSS -
Division of Reactor Safeguards and Security ECCS -
Emergency Core Cooling System EHC
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Electro-hydraulic Control System EM
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Electrical Maintenance ESF
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Engineered Safety Features ESS
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Engineered Safety Systems FHP
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Fuel Handling Procedure FNP
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Farley Nuclear Plant FP
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Fire Protection FW
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Feedwater GMP
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General Maintenance Procedure HP
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Health Physics ISI
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In-service Inspection I&C
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Instrumentation and Control KW
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Kilowatt LC0
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Limiting Condition for Operation LER
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Licensee Event Report L/D
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Letdown LOSP -
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Motor Conu ol Center MOV
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Motor-0perated Valve MSIV -
Main Steam Isolation Valve MTC
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Moderator Temperature Coefficient MW
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Megawatt NWR
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Maintenance Work Request NDE
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Non-Destructive Examination NCV
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Non-cited violation NI
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Nuclear Instrument or NIS (system)
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Out Of Service PCN
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Plant Change Notice Preventive Maintenance PM
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Penetration Room Filtration System psig -
pounds per square inch RCS
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Reactor Operator RWT
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Reactor Water Storage Tank SB0
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Station Blackout SFI
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Shift Foreman Inspecting SF0
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Shift Foreman Operating SFP
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Spent Fuel Pool S/G
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Steam Generator Feedwater Pump
SNC Southern Nuclear Operating Company
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Systems Operator S0P
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System Operating Procedure s
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SRO Senior Reactor Operator
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Shift Supervisor STAR -
"Stop", "Think", "Act", " Review" j
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Surveillance Test Procedure
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Service Water System Tavg -
Temperature (average).in the RCS TCN
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Temporary Change Notice UDP
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Unit Operating Procedure UFSAR -
Updated Final Safety Analysis Report URI
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Unresolved Issue VAC
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Volts Alternating Current
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