IR 05000458/1992026
| ML20127D426 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 09/04/1992 |
| From: | Harrell P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20127D404 | List: |
| References | |
| 50-458-92-26, NUDOCS 9209150041 | |
| Download: ML20127D426 (14) | |
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APPENDIX B U.S. NUCLEAR REGULA10RY COMMISSION
REGION IV
NRC Inspection Report:
50-458/92-26 Operating License: NPF-47 Docket:
50-458 Licensee: Gulf States Utilities P.O. Box 220 St. Francisville, Louisiana 70775 Facility Name:
River Bend Station Inspection At:
St. Francisville, Louisiana Inspection Conducted: July 5 through August 15, 1992 Inspectors:
E. J. Ford, Senior Resident Inspector D. P. Loveless, Resident inspector
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Approved:
. i ef, roject Section C Date P. H.
r inspection Summary leapection, Conducted Julv 5 through August 15. 1992 (Re_ port 50-458/92-26_1 Areas Inspected: Routine, unannounced inspection of licens?e event report followup, onsite-followup of events, operational safety verification, maintenance and surveillance observations, and engineered safety features-walkdown.
Resuly:
The radiological protection department made appropriate and conservative o
actions when transporting a potentially_ contaminated individual offsite (paragraph 4.a).
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_ One unresolved item was identified to evaluate the adequacy of o
procedural controls that led to the damage of the Division I diesel-
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generator valve.. train.. In addition, the adequacy of initial corrective action prior to running the engine will be evaluated (paragraph 4.b).
The licensee perfor.ned a valid core verification ana repaired several o
minor damaged core components (paragraph 4.c).
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-2-The licensee's actions were conservative in minimizing shutdown risks o
during the current outage by ensuring decay heat removal was available and licensed operator oversight of safety-related work (paragraph 5.a).
Plant-wide housekeeping was found to be poor during a plant tour o
conducted July 20, 1992 (paragraph 5.b).
A noncited violation was identified for an inappropriate radiological o
A posting (paragraph 5.c).
One violation was identified for failure to appropriately control the
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o replacement of the soft seat rings in the safety / relief valve air accumulator check valves (paragraph 6).
The licensee sat;sfactorily ensured the adequacy of decay heat removal o
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upon initiation of core reload by temporary in-vessel temperature instrumentation, operator monitoring of temperature sensors, and cooling system alignment (paragraph 7.a),
One violation was identified for the failure to test the safety relief o
valve air accumulator check valves in accordance with Paragraph IWV-3522(a) of the ASME Code, as required by Technical Specification 4.0.5-(paragraph 7.b).
A portion of the reactor water cleanup system was reviewed and it was in o
proper alignment for the plant conditions and in good material condition (paragraph 8).
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-3-DETALLS 1.
Persons Contacted R. E. Barnes, Supervisor, Codes and Standards W. J. Beck, Supervisor, Balance of Plant Design J. E. Booker, ihnager, Nuclear Industry Programs E. M. Cargill, Director, Radiological Programs J. W. Cook, Technical _ Assistant T. C. Crouse, Manager, Administration W. L. Curran, Cajun Site Representative J. C. Deddens, Senior Vice President S. V. Desai, Principal Engineer
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P. D. Graham, Plant Manager G. R. Kimmell, Director, Quality Assurance
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D. N. Lorfing, Supervisor. Nuclear Licensing 1. M. Malik, Supervisor, Quality Operations W. H. Odell, Manager, Oversight J. P. Schippert, Assistant Plant Manager - Operations Radwaste and
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Chemistry J. E. Spivey, Engineer, Senior Quality Assurance Engineer
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R. J. Vachon, Senior Compliance Analyst C. W. Walker.. Supervisor, Quality. Control
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C. W. Walling, Supervisor, Process Systems The above listed personnel attended the exit meeting conducted on August 21, 1992.
In addition to the above personnel, the inspectors contacted other personnel during this inspection period.
2.
Plant-Status At the beginning of this inspection period, the reactor was defueled and the plant was in Day 115 of a 156-day outage that commenced on March 12, 1992.
On July 22, the first fuel bundle was returned to the reactor vessel and the plant reentered Mode 5.
On July-28, the last fuel bundle was placed in the reactor and the reactor was completely refueled.
At the end of this inspection period, the reactor was in Mode 4 and the plant was in Day 157 of the current refueling outage.
3.
Followup of an Licensee Event Report (92700)
(Closed)
Licensee Event Report 92-005:
Reactor Scram Caused by a t.
Generator Trip Due to High Winds Causing Transformer Damage h
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This licensee event report described an event, on March 5, 1992, in which the unit scrammed from 100 percent power because of turbine control valve fast closure. This fast closure and attendant turbine trip resulted from a generator trip when high winds blew sheet metal siding loose from the turbine building and it shorted the main transformer.
The damage to the turbine building resulted in potentially contaminated fiberglass insulation and other debris being scattered onto the buildings and grounds within and beyond the protected area.
This event was previously reviewed in detail and documented in NRC Inspection Report 50-458/92-08 prior to the issuance of the licensee event report; therefore, this licensee event report is closed.
4.
Onsite Followup of Events (93702)
a.
Transportatjon of. potentially Contaminated Individual
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On July 7,1992, the licensee declared a Notificat, ion of Unusual Event when a potentially contaminated individual was tra-sported to West Feliciana Hospital.
The individual, a contract radiation
>rotection technician, was working in the drywell behind the
)iological shield wall when he fell through an opening in the grating and sustained back injuries.
The responding emergency medical technician placed the individual on a solid-back board before radiation protection could conduct a complete radiation survey.
An additional radiation protection technician was dispatched with the individual. The individual was surveyed at the hospital and determined to be free from contamination.
This was evidence of appropriate, conservative actions by the radiological protection department.
b.
Damage to a Diesel Valve Train Following Maintenance
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On July 8, 1992, while licensee personnel were adjusting the valve settings on the Division I emergency diesel generator, the engine failed to turn using the barring device. A valve-train inspection was conducted by the licensee and it was noted that the Cylinder 5 rear intake valve adjusting screw wa1 set too far into the rocker arm assembly. The adjusting screw was in contact with the valve and was loaded to a point that it could not be loosened without turning the engine in the reverse direction first.
Prior to this-event, the licensee had performed a series of complex inspections of the engine under Surveillance Test Procedure STP-309-7614 " Diesel Generator Inspection - Divisic9 I
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and Division II."
The review of this inspection is documented.in
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-5-NRC Inspection Report 50-458/92-24. Apparently, the adjusting screw was inappropriately positioned during its reinstallation following the inspections.
The licensee stated that backing down the adjusting screw prior to the reassembly of the valve rocker arm was a routine good practice and should have been within the skill of the craft. The inspector questioned the adequacy of the reinstallation procedure Refueling Outage Procedure Rf0-424 " Inspect Turbo Charger Bracket Bolting and Gaskets and Retorque Fuel Injector Retaining Bolts," utilized in conjunction with Maintenance Work Orders R152114, R152124, and R152113. This procedure will be reviewed for adequacy by Regional inspectors during a scheduled review of the Division 1 diesel generator inspection data.
This item is considered unresolved pending further NRC review and resolution of the appropriatennss
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of the procedural controls (458/9226-01).
The licensee disassembled Cylinder 5 for inspection and determined that.the valve had contacted the piston head and overstressed the valve train. During the inspection, numerous parts were found to be damaged and were replaced.
The engine had.been barred through the area of binding twice before the conditions were found. As a result, the licensee's
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engineering organization, supported by the vendor, performed calculations to analyze.the stresses in the engine..The analysis
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showed that the weakest points of the valve train were the-push rod, subcover casting, and the valve.
The inspection results corroborated this analysis.
Further analysis showed that it was impossible for the engine barring device to generate forces approaching the magnitude of the combustion forces on the power train.
The barring device was determined to apply a maximum of 7,000 psi to the rim of the-flywheel. The licensee concluded that the' power train components were never in any danger from the binding-in the valve train while barring the engine.
Following the replacement of the damaged parts, several test runs of the diesel were performed. On July 25, while: performing the retightening of the cylinder head air start valves required by_
Paragraph 1.6.12 of Procedure RF0-424, licensee personnel found that the intake push rods on Cylinders 1 and 7 appeared to be slightly bent.
The inspector questioned the-adequacy of the initial corrective actions taken.between July 8 and 23.
Other cylinders were not, inspected for generic adjustment problems, as found in Cylinde? 5.
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Had the-licensee performed these inspections, the other bent push rods would have been identified prior to the diesel generator
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runs, which had the potential for further damaging the engine.
Although the bent rods were found during a scheduled inspection, it is unclear that the bent push rods would have been readily identified by all inspecting personnal.
The appropriateness of the licensee's initial corrective actions will be reviewed by Regional inspectors during a scheduled review of the Division I diesel generator inspection data.
This review will include the evaluation of whether the licensee met regulatory requirements of its corrective action program and the requirements
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of Technical Specification 4.8.1.1.2.f.l.
This item is considered a second example of the unresolved item (459/9226-01).
The licensee determined that the valves were seating properly
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based on diagnostic measurements of cold compression and peak firing pressures taken during engine runs on July 23 and 25.
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licensee determined that there was no reason to remove any cylinder head unless damage was seen on.the top end inspections.
The licensee concluded that the cause was the same as that which caused the damage to Cylinder 5;'however, the valves did not contact the pistons.
Following a complete inspection, the licensee replaced three additional intake pushrods on Cylinders 1, 7, and 8.
Additionally, the licensee inspected all pushrods in the Division 11 emergency diesel generator.
No discrepancies were identified.
The licensee performed the additional required inspections and runs of the Division I emergency diesel generatur and declared it operable.
c.
Damaged Core Comoonents During Refueling On April 11, 1992, during the offload of the core, the refueling operators had difficulties grappling Fuel Bundle LYV299. After several attempts, the bundle grappled light illuminated and the bundle was transported to the inclined fuel transfer system. Once seated in the cart, the operators noted that the bail handle appeared to be bent, 'The bundle was transported to the fuel preparation machine, where it was further inspected.
The-licensee pointed out that the damage could have occurred at an earlier time.
The exact cause and date of the damage is unknown.
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Following inspection and discussions with the fuel supplier, the bail-handle was replaced on July 10. The_ licensee stated that although River Bend Station has never had a bent bail handle, they-are not uncommon in the industry,
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On July 28, during the core vctification process, the licensee determined that two bundles had bent channel fasteners.
The licensee concluded that these had been bent as the bundle was inserted into the core and that the festener had hung up on the adjacent bundle.
Both bundles were transported to the lower pool and repaired.
The licensee also identified 14 fuel bundles in the peripheral locations that were not fully seated in their fuel support pieces.
The licensee reseated tho bundles utilizing the refueling machine.
The licensee then verified the seating of every peripheral fuel bundle by viewing the nozzle area with a video camera.
Conclusions The radiological protection dooartment made appropriate and conservative
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actions when transporting a potentially contaminated individual offsite.
One unresolved item was identified to evaluate the adequacy uf procedural controls that led to the damage of the Division I diesel generator valve train.
In addit %, tne adequacy of initial corrective
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action prior to running the er.gine w:11 be evaluated.
The licensee performed a valid core verifica' tion and repaired several minor damaged core components.
5.
09erational Safety Verification (71703 a.
Control Room Tours During the first part of this inspection period, the reactor was defueled and Technical Specifications do not address requirements for control room manning. However, the inspector noted that a larger than normal number of licensed reactor operators were continuously manning the control room and providing ~ oversight of the safety-related work in the plant.
Following the refueling, a reactor operator was continuously at the controls, verifying -
appropriate safety system status and decay heat removal.
On July 29, 1992, the inspector verified that appropriate alternate shutdown cooling was in effect with Residual Heat
' Removal System B in the-upper fuel pool cooling assist mode.
Altnough not required by Technical Specifications, the inspector verified that the high pressure core spray and Residual Heat
. Removal System C were available te inject water into the reactor, if necessary.
The licensee frequently maintained a larger number of systems available than the minimum required by Technical Specifications to minimize-shutdown risks.
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Both of the preceding items are considered to be positive actions
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-to ensure decay heat removal and the oversight of safety-related work.
The inspector has noted, on occasion, the wearing of soft shoewear
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in the control room by nonoperations personnel.
This is contrary to the safety expectations of plant management.
The inspector discussed this matter with the Assistant Plant Manager -
Operations, who stated that the need to enforce this industrial i
safety requirement would be discussed and reinforced with control room personnel.
-b.
Plant Toun On July 22, 1992, the inspector accompanied the control
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operating foreman on a tour of portions of the fuel building and the primary containment.
Particular attention was paid to ongoing activities on the refueling floor and preparations to resume refueling activities.
The inspector discussed aspects of the ongoing activities with members of-the refueling team on the refueling platform and observed the movements of fuel _into the vessel by the refueling personnel.
The inspector noted that lighting in the containment, the upper fuel pool, and the reactor vessel was adequate, and water clarity was sufficient to ensure good visual contact with the fuel being moved.. It was noted that_the licensee was appropriately
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implementing captivation controls, industrial safety controls for refueling bridge ingress and egress, and appropriate radiological controls.
The inspector noted that housekeeping was acceptable for the refueling areas.
Howaver, housekeeping was found to be poor during a tour conducted July 20. in the turbine building, auxiliary building, and some
- tunnels. Management response was to make housekeeping in these areas timely, aggressive, and thorough; nevertheless, this showed a decline in attention in this area.
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c.
Radiological Controls On August 4, 1992, the posting of the reactor building-equipment hatch, a radiation area, required correction, in that, the entrance to the area was posted as a radiation area, but posting just out_of sight behind the open reactor building equipment hatch indicated that it_was a high-radiation area.
This is contrary to the requirements of Radiological-Protection Procedure RPP-0005, " Posting of
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Radiologically Controlled Areas," which addresses the
posting.of radiological controlled areas.
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The inspector discussed the matter with radiological protection personnel and reviewed the licensee's Condition Report 92-0667 that described the condition.
It was noted that the inconsistent posting was caused by an oversight by the personnel involved in deposting the area. This oversight resulted in the area being erroneously posted, but in a conservai.ive manner.
As a result of
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discussions later that day with the inspector, the Director of Radiological Controls stated that all areas of the plant would be l-verified to have correct posting.
This was completed by the next day with no further discrepancies uncovered by the licensee.
Because the safety significance of the apparent violation was minor and corrective actions were promptly initiated, this meets the criteria specified in Paragraph VII.B.1 of Appendir C to 10 CFR Part 2 for a noncited violation, and Accordingiy, a Notice of Violation will not be issued.
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d.
Inadeauate liftina Eauipment
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During a tour on July _ 29, 1992, the inspector identified that a Harrington brand trolley was in use in the auxiliary building.
This brand of lifting equipment had been placed on hold by the iicensee pending evaluation of its appropriate usage.
This problem was documented in NRC Inspection Report 50-458/92-24.
The licensee stated that this particular trolley was not in the licensee's lifting equipment prognm.
The trolley is scheduled to be removed, and licensee personnel stated that they would perform additional searches for any inappropriate applications of this type of trolley. The licensee is still performing corrective actions for previous findings in this area. The inspector had no
'further questions.
Coi.clusions The licensee's actions wre conservative in minimizing shutdown risks during the current outage by ensuring decay heat removal and licensed operator oversight of safety-related work.
Plant-wide housekeeping was found to be poor during a plant tour conducted on July 20, 1992.
A noncited violation was documented for an inappropriate radiological posting.
6.
Maintenance Observations (62703)
On July 29, 1992, the inspector observed portions of the work being performed under Maintenance Work Orders R 152605, R 152606, R 152607 R 152608, and R 152609.
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valves =in the main steam safety relief valve instrument air system.
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Earlier in the outage the soft seats in these 2-inch, pistc.1-type check valves had been replac,ed based on the emiration of their equipment qualifiertion lifetime.
The insph or verified that the appropriate administrative controls were obtained an6 im place and welding Tctivities were performed by a qualified indi0idual.
The replaceinant seats were obtained on a valid requisition and dere certified as safety related. Weld material procured Nas properly documented and was in accordance with the weld data sheet.
The inspector-reviewed the radiation protection support or survey regaest sheec and the associated Radiation Work Permit 92-3004.
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area had recei d a job specific survey and was documented on an addendum sheet. The crew entering the drywell received an area-specific briefing by the-radiation protection technicians.
Additionally, the-inspector noted that the valve internals were removed to the hot machine shop tor repair.
This showed forethought in the areas of ignit.on prevention controls and radiological protection.
The inspector reviewed the work plan for adequacy.
$tep Rl-6 required
the' mechanic to install a new soft seat ring on the disk assembly and lock weld in accordance with the weld data sheet.
The inspector observed the performance of this sto on several of the disk assemblies.
-The inspector noted that this step was nerformed differently by different mechanics.- It tppeared that differing amounts of torque were applied to the soft seat ing.
LThe inspector questioned the mechanics on the requirements for this replacement. The mechanics informed the insper. tor that no specific guidance-was given for installing the new seat. One' mechanic showed the inspector how he performed.the installation to achieve consistency.
However, it appeared from direct observation of the installation that the way mechanics torqued.the retaining nut was not consistent from one
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mechanic to another.
The inspector reviewed the vendor's manual and determined that it did
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l not give any guidance in the area of soft seat ring replacements. The c
- manual appeared to concentrate on valves with hard seats and; therefore,
was. silent on the soft seat replacement. Additionally, Corrective
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Maintenance Procedure CMP-9173, " Check Valve ReworkJ is too generic and l.
did not provide an adequate level of detail for soft seal ring
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Licensee personnel contacted the vendor who indicated that the soft seat material had to protrude.1/32 to 1/16 inch past the surface of the
washer after torquing, and that failure to adequately tighten the seSt l:
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could cause bypass leakage and the retaining washer to impact the
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stellite seat as opposed to the soft seat ring.
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following rework of the valve, which included torquing the soft seat ring as specified by the vendor, all five valves passed their leak rate tests. The failure to appropriately control work of safety-related activities is a violation of Technical Specification 6.8.1 (458/9226-02).
Conclusions
- One violation was identified for failure to appropriately control the replacement of the soft seat rings in the safety relief valve air accumulator check valves.
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Surveillance Observations (6172sl a..
Temperature Verification of Decav Heat Remoy d On July 22, 1992, the inspector observed portions of the-performance of Surveillance Test Procedure STP-204-0700,
" Alternate Decay ileat Removal Verification." A portion of this procedure was being conducted to log reactor pressure vessel temperature data for the succeeding period after demonstrating that the alternate method of decay heat removal by Residual Heat Removal B - Fuel Pool Cooling Assist Mode was satisfactory. The licensee had installed two temporary indicators in the open reactor vessel to monitor reactor coolant temperature during the mode change to Operational Condition 5 as the core reloading commenced.
The inspector noted that the temporary instruments were installed
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using Prompt Modification Request 92-0014', and STP-20',-0700 had-been appropriately changed to accommodate the temperature verification by Change Notice 92-0826. A review of Attachment 4 to STP-204-0700-did not disclose any temperature readings that were unexpected. The inspector observed the placement-of the temperature sensors in the reactor vessel-and the location of the test instruments being utilized to monitor the sensors and had
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discussions with personnel responsible for obtaining the readings.
Additionally, the licensee's methodology for decay heat removal (upon reloading the reactor vessel) had been reviewed by NRC personnel during telephone discussions with members of plant
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management and representatives of their' engineering dupartment.
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lhe _ licensee had.an adequate means of decay heat remqval for core reload and sufficient means of temperature monitoring to ensure that verification of decay heat removal could be satisfactorily achieved.
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Check Valve Ogerabili_t2 Test On July 31, 1992, the inspector observed portions of the performance of STP-202-3603, " ADS /SRV Accumulator Check Valve leak Rato Valve Operability Test."
This procedure was written to verify that the accumulator leak rate is within the requirenents of Technical Specification 4.0.5 and that the air inlet check valves close in accordance with the same specification.
This, in effect, verifies the operability of this portion of the automatic depressurization system.
The inspector verified that the appropriate procedural approvals had been obtained and an adequate clearance was in place.
This system was out of service during the refueling outaae and was not required by Technical Specifications.
The pressure gauge used for leakage verification had been calibrated within the current calibration cycle.
The inspector reviewed the procedure for adequacy.
The procedural steps first required depressurization of the instrument air header supplying the accumulators. A local leak rate test was then performed to verify the leak tightness of the valve.
During operations, these valves routinely open to make up for any acc nulator leakage.
Therefere, an as-found leak rate test would prove that the valve had been exercised, it closed under its own power, and it was leak tight.
Earlier in the outage, the licensee had performed an equipment qualifications preventiva maintenance task on these valves that required the disassembly of the valves.
Therefore, when STP-202-3603 was performed, it did not verify the exercising and closing of the check valve.
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Technical Specification 4.0.5 requires the licensee to implement the inservice testing requirements of plant pumps and valves required by Sectirn XI of the ASME Pressure Vessel Code, except
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where specific relief is granted by the Commission.
Section IWV-3521 requires that check valves be exercised at least once every 3 months.
The licensee applied for relief under Valve Relief Request 9.
Section 3.6.3.2 of the Safety Evaluation Report granted the relief stating that, " full-stroke exercising these valves during refueling outages... should demonstrate proper valve operability and provides a reasonable alternative to the Code frequency requirements."
Section IWV-3522(a) provides the requirements for the exercisilg precedure for normally open check valves.
This paragraph statas that " Valves that are normally open during plant operation and
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-13-whose function is to prevent reversed flow shall be tested in a manner that proves that the disk travels to the seat promptly on cessation or reversal of flow.
Confirmation that the disk is on its seat shall be by visual observation, by an electrical signal initiated by a position indicating device, by obser/ation of appropriate pressure indications in the system, or by other positive means."
The procedure was originally written to allow the check valve to seat itself upon system depressurization and then perform a leak rate check.
Disasseinbling the valve, then leak rate testing, does not prove that the valve opens and that the disk travels to the saat promptly on cessation or reversal of flow.
This is a violation (458/9226-03).
Conclusions The licensee satisfactorily ensured the adequacy of decay heat removal upon initiation of core reload by temporary in-vessel temperature instrumentation, operator monitoring of temperature sensors, and cooling system alignment.
One violation was identified for the failure.to test the safety relief valve air accumulator check valves in accordance with Section IWV-3522(a) of the ASME Code, as required by Technical Specification 4.0.5.
8.
Engineered Safet' 'eatures Walkdown (71710)
On August 7,1992, the inspector performed a detailed walkdown of the reactor water cleanup system ring header and reactor drain lines.
This piping was replaced during the current refueling outage to reduce radiation exposures to plant personnel in future. outages.
The inspector determined that the system was aligned properly for the plant conditions and that the motor-operated valves were in good repair.
Hangers and supports were properly made up and aligned correctly, and pins were in place, as necessary.
Also, hydraulic snubbers were free to move, as necessary.
Valves in the system were noted to be installed correctly and did not exhibit packing leaks, bent stems, missing handwheels, or improper labeling.
Irstrument lines were properly sloped to provide for valid indications.
The inspector verified the system alignment from the control panels.
All valves were in correct alignment for the part conditions and remote position indication verified that power we. wilable to the motor-operated valves.
Flow and temperatui c..dications were as expected, indicating that the system was functioning properl c,;
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Conclusions
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The portion of the reactor water cleanup system that was included in this review was in proper alignment for the plant conditions and in good
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material condition.
9.
Summary of Open [tems
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The following is a synopsis of the status of all npen items generated
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and closed in this report.-
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The following items were opened:
Unresolved Item 458/9226-01 (paragraph 4.b)
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o Violations 458/9226-02 (paragraph 6) and 458/9226-03 (paragraph.
7.b) Check Valves in Accordance with IWV-3522(a).
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Licensee Event Report 92-005 was closed o
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Exit Meeting
An exit meeting was conducted with licensee representatives identified ir-paragraph 1 on August 21, 1992. During this tuterview, the inspectors reviewed the scope and findings of the report.
The licensee
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did not identify as proprietary, any information provided to, or reviewed _by the inspectors.
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