IR 05000445/1992047
| ML20126J332 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 12/24/1992 |
| From: | Yandell L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20126J302 | List: |
| References | |
| 50-445-92-47, 50-446-92-47, NUDOCS 9301060120 | |
| Download: ML20126J332 (25) | |
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APPENDIX _B
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U.S. NUCLEAR REGULATORY COMMISSION'
REGION IV
-Inspection Report: 50-445/92-47 ,
50-446/92-47 Operating License: HPF-87 Construction Permit: CPPR-127 Expiration Date: August 1, 1995 Licensee: TV Electric Skyway Tower 400 North Olive Street L. B, 81 Dallas, Texas 75201 Facility Name: Comanche Peak Steam Electric Station, Units 1 and 2 Inspection At: Glen Rose, Texas inspection Conducted: October 11 through November 25, 1992-Inspectors: W. B. Jones, Senior Resident Inspector '
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G. E. Werner, Resident inspector R. E. Lantz, license Examiner P. A. Goldberg, Reactor Engineer Reviewed by: _
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L. A. Yandell, Chlef, Project Sectio l D/te 2.4 42
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Division of Reactor Projects Inspection Summary Inspection Conducted October 11 through November 25. 1992 (Report 50-445/92-47)
Areas Inspected: Routine, unannoun' red. inspection of onsite followup of plant events, refueling activities, operational safety verification, maintenance and surveillance observations, engineered safety features walkdown, a temporary instruction (2515/113, Reliable Decay Heat Removal), and previously identified item *
Results:
- The operators response to the reduction in feedwater flow, with the reactor at full power, was excellent (paragraph 2.1). The subsequent startup and shutdown'for the refueling outage were well controlled (paragraphs 4.1 and 4.2).
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. Communications between the reactor operators was generally goo i
'However, an instance was noted where the operator assuming licensed duties was not fully cognizant of plant-status. It was also noted that managements' expectations were not met for self-verification while placing an instrument air system purge into service (paragraph 4.3).
. The operations shift supervisor responded very well to the security compromise and appropriately classified the event. The licensee effectively communicated between the different organizations to establish the conditions which were to be met prior to exiting from the NOUE (paragraph 2.2).
- Several configuration control issues were identified which impacted the design modification process and work control implementation (paragraph 2.3).
- Two radiological uptake events and several personnel contaminations-resulted from contract personnel not being fully cognizant of radiation work requirements and not utilizing good work practices. Additional poor work practices were observed for permanent plant personne Remedial radiation worker training, and enhanced supervisory and radiation protection technician oversight, was required to ensure adequate contractor radiological work practices were utilized (paragraph 2,4).
- The AL. ARA program was well implemented for refueling outage work activities. One instance of improper control of radioactive materials was note General area housekeeping within the radiologically controlled area'was good (paragraphs 4.4 and 4.5).
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. Heavy loads were not adequately controlled in-the area of the reactor vessel head. A violation was identified for the > inadequate riggin activities (paragraph 3.1).
- Excellent communications were noted between the reactor operators and contract personnel during the reactor ~ core offload (paragraph.3.2).
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- Appropriate measures-were established to identify th'e-fuel assemblies with failed rodlets and to assure they were not returned to the; core for Cycle.3. operation (paragraph 3.3).
l . Maintenance ac_tivities were-generally performed in accordance with L
procedure requirements. A violation was identified for inadequate:
i control of tools used on stainless steel safety-related components l (paragraph 5.1). .
- Surveillance activities were performed in accordance with the procedure requirements (paragraph 6).
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- The design modification process was effectively utilized and should provide enhancement of the radiation monitoring-system (paragraph 7),
- The licensee implemented appropriate-measures to assure reliable decay heat removal (paragraph 8.1).
Summary of Inspection findings:
- Violation 445/9247-01 was opened (paragraph 3.1).
- Violation 445/9247-02 was opened (paragraph 5.1).
- Unresolved item 445/9247-03 was opened (paragraph 2.3).
e Unresolved item 445/9247-04 was opened (paragraph 2.4).
- Inspection Followup Item 445/9019-05 was closed (paragraph 9.1).
- Unresolved Item 445/9220-07 was closed (paragraph 9.2).
- Unresolved Item 445/9240-01 was closed (paragraph 9.3).
- Unresolved Item 445/9231-02 was closed-(paragraph 9.4).
Inspection Summary (Report 50-446/92-47)
Inspection Conducted October 11 through November 25. 1992 (Report 50-446/92-47)
Areas Inspected: A previous unresolved item and inspection followup item were evaluated and closed.
Results: Not applicable.
Summary of Inspection Findin_qs:
- Inspection Followup Itern 446/9019-05 was closed (paragraph 9.1).
- Unresolved item 446/9220-07 was closed (paragraph 9.2).
Attachments:
e Attachment 1 - Persons Contacted and Exit Meeting
-l-4-DETAILS 1 PLANT STATUS (71707)
At the beginning of this inspection period, the plant was at 100 percent power. On October 12, 1992, a manual reactor trip was initiated because of the Steam Gent.rator 4 feedwater regulating valve going closed, it was found that the spring within the associated pressure regulator had failed, causing the flow control valve to close. On October 14 a plant startup was commenced and the unit returned to full power.
On October 22 the licensee began reducing reactor power in preparation for the second refueling outage. The main generator output breakers were opened on October 23. The reactor was placed in a subcritical condition by inserting the control rod control banks. Later that morning a control rod urgent failure alarm was received and four Control Bank B Group 1 control rods dropped. The reactor operator initiated a manual reactor trip and verified all the shutdown and control rod banks were inserted. The licensee is currently investigating the cause of the control rod urgent failure alarm.
The unit entered Mode 5 on October 26 and Mode 6 was entered on November 3.
The core was completely off-loaded November 9 and . Mode 6 exited. The core remained off-loaded through the remainder of this inspection period.
2 ONSITE FOLL0ilVP OF EVENTS (93702)
2.1 Reactor Trip On October 12, 1992, an annunciator was received which identified a Steam Generator 4 feedwater/ steam flow mismatch. The operator took immediate action to recover feedwater flow by placing the feedwater flow controller in manual and input a 100 percent demand signal to the respective feedwater flow control valve. The bypass valve around the feedwater flow control valve was.also opened to provide additional feedwater flow. The operators were not able to provide sufficient flow to the steam generator and the unit was manually tripped prior to the steam generator level reaching the LO-LO reactor trip setpoint. The auxiliary feedwater system started and the unit was stabilized in Mode The inspectors were promptly notified of the manual reactor. trip and the expected safety features activations. The required 10 CFR Part 50.72(b)(2)(ii) notification was made to the NRC Operations. Center within the 4-hour period following the event. Later that day, the-inspectors met with the licensee to ascertain the cause for the feedwater regulating valve f ailing closed and independently reviewed the plant response. The inspectors verified that the plant response was consistent with the accident analysis in Final Safety Analysis Report 15.2.7, " Loss of Normal Feedwater." The reactor operators' activities were appropriate and appreciably mitigated the transien m _ .m _ _ . . _ _ _ _ _ . . _ _ ~ . - ._ _ . _ _ _ _ _._ _-_ _ _ . . . _ . _ _
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The inspectors met with maintenance personnel to discuss the feedwater
. regulating-valve failure. The licensee identified that a: spring in the-associated feedwater regulating valve had broken causing a loss-of-air -
pressure to the-feedwater regulating valve. During the assessment of the failure, it was determined that an incorrect pressure regulator had been installed for the feedwater regulating valve. A pressure regulator with a range of 0-60 psig-was installed where a 0-125 psigLregulator was require The pressure regulator was replaced with the 0-125 psig-pressure regulato The licensee submitted Licensee Event Report (LER) 92-22, " Manual Reactor Trip Due to feedwater flow Control Valve f ailure," on November 11. This report identified that the root' cause for the event was a lack of self-verification-in 1990 when the pressure regulator was replaced. An error in the master parts list was found which identified the 0-60 psig pressure regulator as acceptable. The inspectors will review the licensee's corrective action as identified -in the LER and Operations Notification and Evaluation (ONE)
form 92-972 during the LER followu .2 Declaration of Notification of Unusual Event (NOVE)
On-October 21 the inspectors were notified that a security officer had been physically assaulted and that a plant security compromise existed. Based on this information, the operations shift supervisor declared a NOUE at 4:51 =
The inspectors met with senior licensee management personnel in the technical
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support center and observed the shift supervisor's actions in the control room, Based on the information provided, the inspectors found that the NOVE declaration was appropriate. The licensee continued.to assess the security compromise. The NOVE was exited at 8i47 a.m. after certain predetermined conditions were me The licensee submitted LER 92-23, " Physical Assault of Security-Officer Results in Security Compromise," on November 20. A special-inspection was in progress during this inspection period to assess the licensee's actions. The
- results of this inspection will be documentedLin NRC Inspection-
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Report 50-445/92-50; 50-446/92-50, 2.3 Configuration Control The licensee has ' identified several. configuration control ' issues since the beginning of the second refueling outage. The principle areas identified were-discrepancies- between design and as-built plant: conditions,- plant status including temporary modifications,-temporary jumpers not installed in accordance with design, and plant:evolutlons resulting in unanticipated
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alarms. The-licensee conducted a daily review of each ONE form, which included discussing. any configuration control issues. An-overall assessment was scheduled each week- to determine if rn: apparent configuration. control
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problem- trend existed. . The-inspectors noted that this review was not always conducted even though configuration control issues had emerged since- the .last meeting,
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- The inspectors reviewed several of the configuration control issues and categorized the incidents into two general areas. -These areas were electrical design modification and work contro The inspectors'found that the licensee had taken prompt action to evaluate-the *
electrical design modification configuration control issues. This included assessing the potential impact these issues may have on the unit and whether the design modification implementation process and postmodification testing was adequate to identify similar issues. The licenseo noted that the design modification packages which contained errors had been performed by contractors. The contractors were notified of the configuration control problems. The licensee is requiring that each contractor respond to these concerns. In addition, the licensee will be sampling as-built electrical-cabinets to assess the accuracy of the electrical-drawings. The inspectors will continue to review the licensee's program for reviewing and accepting contractor initiated design document The inspectors reviewed the configuration control ONE forms which identified concerns with the work control process. Three events were selected-to
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determine the adequacy of the licensee's corrective actions. The three events reviewed were: (1) a loss of component cooling water flow to _both control room heating, ventilation, and air conditioning units rendering them inoperable (FX 92-1223); (2) a clearance issued to electrical maintenance for transformer work with the-transformer still' energized (FX 92-1214); and-(3) ,
the partial draining of. the nonsafety chilled water system surge tank to the component cooling water system drain tank (FX 92-1287).
The inspectors noted that the licensee did not appear to adequately assess the events for the corrective actions which should have been promptly implemented.
An_ example of this concern involved the control room heating, ventilation, und
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air conditioning system event, with the reactor' vessel defueled, where a Units 1 and 2 interface concern appeared to have-not been addressed. A second-example invelved the energized transformer and the need to assure that-temporary modificationsLbe properly reviewed for' impact from the outage: system configurations. The last event appeared to;have. occurred,-in part, because; the order in which theLclearance was lifted may not have been appropriat This concern was presented to the licensee's senior management at the end of; the inspection period. These examples are considered an unresolved item
- pending further review in conjunction with other-related issues that will be-evaluated during the subsequent inspection period (445/9247-03). The
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inspectors will continue to evaluate the ' licensee's immediate and long-term :
corrective actions taken for thel identified configuration control problems'.
2.4- Radiological Worker Practices On November 16 and'17:two radiological contaminat' ion-events occurred which-resultedJin uptakes to contract employees. -The licensee initiated an.
I investigation into the two events. Both uptakes were found to have occurred l- because of poor radiological work practices. _ The first uptake. event occurred while eight maintenance personnel were reinstalling and tensioning the man-way l -
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-cover to Steam Generator 3. The second uptake event occurred while work was being performed' on a chemical and volume control system valve. = Both events occurred in high radiation / contamination area The licensee's-investigation of the first event identified-that the steam generator platform area was extensively decontaminated by radiation protection personnel prior to the work beginning. The radiation work permit for this work activity required full protective clothing, plastics, and bubble hoods for respirator use. These dress requirements-were later altered by the cognizant radiation protection technicians because of communication difficulties between personnel at the platform and control point and because ,
of the extensive decontamination that had been performed. The new dress requirements were protective clothing, paper suits, and face shield Throughout the course of the work activity, continuous radiation protection coverage was provide At the completion of the work activity and after removing all protective clothing, three individuals set off the portal monitors while attempting to-exit the radiologically controlled are Two workers were found to have facial (skin) contamination, one of which resulted in an uptake. - Bothiskin contaminations and the-subsequent uptake were caused by workers reaching under face shields to reposition their safety glasses and making contact with_their skin. The radiation protection technician present during the_ work activity; witnessed this on one occasio The third worker had contaminated the modesty clothing when he tore his paper suit during the work activity. Radiation protection personnel indicated that the contamination entered through the tear in the paper suit and soaked through the protective clothing, contaminating the individual's modesty clothin The second radiological uptake event occurred in the Safeguards Building,'
Room 80. The room was posted as a high-radiation area and. contamination are Three contract workers were performing maintenance activities on a chemical >
and volume control system valve. After the valve was removed from the system, one worker standing approximately 1 foot from the. valve used demineralized -
water. and Scotch-Brite pads to clean the valve internals._ _Two other contract workers also used Scotch-Brite pads on the valve. After completing'the work l
activity, the workers proceeded to- the portal monitors at the radiological-l control point exi The workers were determined to be contaminated.
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Subsequent examination revealed one worker was contaminated on the face -
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around the_ mouth _ area. .A nasal swipe was-taken which verified an uptake. A-whole-body count was-subsequently conducted. After the whole-body count, the
, -individual was interviewed by the. licensee, it was determined-that'the l worker's face had touched the valve while reaching around_ a pipe. to loosen the.
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fl ange- bol t s . The individual also _ stated his face was close to the flange because of poor eyesight. The individua.1 was not wearing corrective lense The inspectors discussed the uptake. events with the radiation protection manage An apparent trend was indicated involving poor radiological
_ practices _ by contract workers. Based on this indication, the license mandated that the radiation protection technicians provide continuous coverage p
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8-of specified contractor work activities in contaminated areas, incidents of poor radiological practices were to be reported to the radiation protection manage Based on these observations, the licensee determined that remedial radiological protection training was neede The following day, the licensee implemented a radiological safety training session for the contract workers' supervisors and held a refresher of
" Radiation Worker Training" for the workers. On November 19 the inspectors attended one of these training sessions. The training stressed the importance of radiation safety and covered topics such as radioactive material, dosimetry and exposure control, contamination and contamination control, and radiation work permit and general access permit training. Following the formal classroom presentation, each worker had to pass a written examination covering the content of the course, in addition, each worker had to demonstrate the proper donning and removal of protective clothing and follow the instructions delineated on a mock radiation work permit to complete the training exc cis All workers adequately demonstrated the proper donning and removal of protective clothing. However, almost all workers, when questioned by the thspectors and licensee training personnel, were not aware of the requirements delineated in the mockup radiation work permit. Some workers were not aware of the definition of a high-radiation area, some were not aware of the number of simulated high-radiation areas they would progress through to complete the exercise, and some were also not aware of appropriate actions needed to be taken if conditions in the work area changed. At the completion of the exercise and af ter removal of protective clothing, the inspectors observed several worket s frisk out using inoperable survey instruments. These observations were discussed with licensee managemen The licensee concluded that additional radiation protection technician oversight would be required for that contractor grou The inspectors witnessed maintenance activities conducted by noncontract personnel inside contaminated areas. Several poor radiological worker practices were observed. These observations included workers using their gloved hands to adjust safety glasses. A worker in a high-radiation / contamination area ms observed to unzip his protective clothing and place his gloved hand into his 11odesty clothing pocket to retrieve dosimetr The inspectors also noted that a worker placed his gloved hand under a leaky residual heat removal drain valve while attempting to reinstall the valv This was contrary to the directions of the radiation-protection technician overseeing the jo The inspectors also discussed the amount of oversight provided by supervisory personnel in contaminated areas with maintenance workers. In general, it was found that managements' expectation that supervisors observe workers was not being met for activities in contaminated area These issues were referred to the Region-based health physics inspectors for further evaluation.and followu These items are considered an unresolved item pending further review by Regional staff (445/9247-04).
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2.5 Conclusions -
The operators responded very well to the reduction-in feedwater flow to one1 steam generator. The operator's prompt actions mitigated the severity of the-event and, prevented _the plant from challenging safety setpoint The shift supervisor responded appropriately to the potential security .
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compromis The basis for declaring a NOVE was well founded and required .
compensatory posts be established. The event was assessed by licensee senior management from the technical support center. The conditions needed to assure the security of the facility were clearly established and met before the NOVE was exite The configuration control issues adversely impacted the design modification process and work control implementation. An unresolved item was identified ,
for further review by the resident office staf The radiological uptake events and several personnel contaminations resulted from contract personnel not being cognizant of radiation worker requirements and good practices. A minimal assurance that contract personnel could work safely in contaminated areas was provided through remedial training, supervisor awareness, and enhanced radiation protection technician oversigh Permanent plant personnel performance in contaminated areas was goo However, several poor work practices were notedLwhere personnel could easil become contaminated. An unresolved item was ident_ified'for further review by Regional staf REFUELING ACTIVITIES (60710)
3.1 Crane Operations On fleveiser 4 with the unit in Mode 6, the inspectors were observing the work associated with the Unit I reactor vessel head ' stud removal. Contractor-personnel were using the polar crane to lowyzand rais.e'the stud removal.-
holder' and track assembly. During the lowering of the track, the inspectors noted that the "C" clamp assemblies attaching the track to the_ polar crane .
wer_e allowed to swing freely and had. bumped. lightly into the reactor. vessel head-assembly. On one occasion, as the track assembly was being lowered into its final position, the inspectors noted that a "C" clamp assembly had hooked a power-cable attached to one.of the head's electric hoists. Neither the-crane operator' stationed at the 905-foot elevation, nor the rigger at the bottom of the refueling cavity directing the crane operator noticed the clearance problem. The inspectors made multiple attempts to gain the attention of the crane _ operator to stop lowering the lifting rig. Prior to any damage occurring, the crane operator noticed the inspectors and stopped the crane _from damaging the electric hois Later,_the lifting rig was repositioned to pick up a track containing removed-reactor studs. 'As the rig was being repositioned, one o_f the "C" clamps hung up below an I-beam. extending-from the cavity liner. The crane operator
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~10-continued repositioning the lif ting rig instead of lowering the load and clearing the obstacle. As the load was repositioned, it suddenly became detached.from the I-beam and swung about in an uncontrolled manne The-lifting rig again bumped into the upper head assembly at one of the electrical junction boxes. The inspectors also noted that, during the removal of the stud filled track, the guide ropes (tag lines) were not long enough to control the load while in the vicinity of the reactor head. Although no visible damage occurred, the inspector concluded that damage to the refueling cavity liner and the reactor vessel upper head co g onents could have occurred because of the weight and momentum of the uncorirolled loa The licensee was immediately informed of these observations and the licensee temporarily suspended lif ting operatior s in the reactor cavity. . ONE form 92-1149 was written by the licens6e to document and evaluate the crane and rigging problems. The licensee innediately initiated the following corrective actions:
- Installed longer tag lines, e Reviewed licensee's expectations with the contractors involved (i.e.,
adhere to onsite procedures), and
- Stationed a containment coordinator tc observe the remaining crane operations in the refueling cavit Maintenance Department procedure MDA-308. Revision 6 -" Crane and Electrical-Hoist Operator Certification Program," Attachment B requires that tag lines shall be. provided and used to control freely swinging loads whenever manual guidance is required to protect personnel or equipment. The procedure als requires care to ensure that rigging equipment clears all-obstacle Additionally, the procedure requires care to be taken while. lifting loads to ensure the load does not contact obstruction's or suddenly accelerate or decelerate. The inspectors identified the failure to follow the procedures-as a violation of Technical Specification 6.8.1 and procedure MDA-308 (445/9247-01).
L 3.2 Fuel Movement Observation The inspectors observed reactor core offload activities from the refueling bridge and the fuel transfer canal. It was noted that the offload -activities were properly coordinated with the control room. A licensed senior reactor operator was stationed .on the refueling bridge during all activities involving fuel movement. The individual at the transfer canal, operating the upender, maintained communications with the personnel at the spent fuel poo .3 fuel-Failures.
L The licensee estimated that between 6-8 fuel rodlets had failed during Cycle'2 operation. . To determine which fuel assemblies had experienced failed fuel rodlets and to assure they were not returned to the core during Cycle 3, the -
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-11-licensee performed IN-CAN and IN-MAST sipping inspections on each fuel assembl Based on the results of these inspections, five fuel assemblies were identified with a total of nine failed rodlet Four of these assemblies had been identified to be returned to the core for Cycle 3. Five of the failures appeared to have resulted from fretting, One upper end cap failure occurred below the weld. The cause for the end cap failure and remaining three rodlets had not been determined.
The core reload will consist of approximately 40 percent of the older generation Westinghouse fuel assemblies. The newer generation fuel assemblies contain debris filter bottom nozzles. A revised core reload analysis (Core Operating Limit Analysis Report) has been approved to permit the core reload with alternate fuel assemblies in place of the four damaged assemblie .4 Conclusions A violation was identified for the fai' lure to properly control heavy loads around the reactor vessel head. Personnel involved with the activity did not demonstrate the expected safety awareness and protect against possible damage to the reactor vessel and/or refueling cavity wal The licensee effectively monitored core off-load activitie Good communications were demonstrated between contract personnel, the control room, and the refueling bridge senior reactor operato The licensee took appropriate measures to determine which fuel assemblies contained failed rodlets. On the basis of this information, the licensee initiated a conservative approach and redesigned the core without the affected fuel assemblie OPERATIONAL SAFETY VERIFICATION (71707)
The objectives of this inspection were to ensure that this facility was being operated safely and in conformance with regulatory requirements, to ensure that the licensee's management controls were effectively discharging the licensee's responsibilities for continued safe operation, to assure that selected activities of the licensee's radiological protection programs were implemented in conformance with plant policies and procedures and in compliance with regulatory requirements, and to inspect the licensee's compliance with the approved physical security pla The inspectors conducted control room observations and plant inspection tours and reviewed logs and licensee documentation of equipment problems. Through in-plant observations and attendance of the licensee's plan-of-the-day meetings, the inspectors maintained cognizance over plant status and Technical Specifications action statements in effec i
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l-12-4.1 Reactor Startup On October 14, 1992, the inspectors observed the reactor startup until the reactor was critical and stable in the intermediate range. Plant recovery was initiated with the plant in Mode Procedure IPO-2A, Revision 8. " Plant Startup from Hot Standby," was followed throughout the observed portion of the startu Supervisory oversight was provided the entire time by a senior reactor operato Communications between the reactor operator and the reactor engineer performing the inverse count ratio were good, thereby allowing close coordination between the two to ensure criticality was achieved close to the calculated critical rod position. The inspectors noted that the reactor _
startup was conducted in an excellent manre .2 Plant Shutdow Control room observations were performed during the plant shutdown for Refueling Outage 2. The inspectors noted that interruptions of the control board operators and unit supervisors were limited to evolutions being conducted in accordance with the shutdown. Tag-out reviews were being ,
completed by an off-shift crew located in an office outside the control roo The inspectors observed the shutdown of Feedwater Pump A in conjunction with the plant shutdown for Refueling Outage 2. The pump was secured by the turbine building auxiliary operator and the field support superviso Procedure 50P-302A, Revision 5, "Feedwater System," was referenced during the shutdown. Step 5.2.1.5 required that motor-driven Feedwater Pump A Suction Valve 1-HS-2321 be closed; however, problems with the motor and manual operator prevented the operators from completing this activity, in accordance with operations department administration procedures, the field support supervisor obtained permission from the shift supervisor to close an upstream condensate supply valve (lCO-0254) in order to isolate the Feedwater Pump A suction. Good teamwork and coordination by the on-shift crew was observe The inspectors observed the addition of hydrogen peroxide to the primary system. Procedure CHM-536, Revision 5, " Chemistry Control of the Primary System," indicated that the initial hydrogen peroxide addition would normally be approximately 5 gallons; however, the chemistry shift orders had been modified to add 1/2 gallon of the chemical for the initial additio A chemistry technician stated that the quantity was changed since guidance supplied by Electrical Power Research Institute was based on solid plant operation. The addition to Unit 1 was performed prior to collapsing the pressurizer steam bubble. Chemistry personnel were concerned that, with a steam space in the pressurizer, hydrogen levels could be exceeded with the addition of a large quantity of hydrogen peroxide; therefore, a conservative approach was taken to ensure hydrogen levels were maintained within a safe rang During the addition of the hydrogen peroxide to the chemical mixing tank, the inspectors noted that the chemistry technicians spilled a small amount of chemical on the floor. Procedure SOP-104A, Revision 5, " Reactor Make-Up and
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-13-Chemical Control System,"- has a caution statement not to allow the funnel or vent to overflow while filling the chemical mixing tank. The inspectors found that the funnel arrangement was poorly designed.- The fill _ arrangement consisted of a tygon tube slipped over the tank fill piping and a-plastic funnel installed in the other end of the tubin Because of the proximity.of-the funnel to the overhead, the personnel adding the chemicals were required to partially bend the tubing and lift the addition bottle to shoulder height while pouring the chemicals. This arrangement made overflowing the funnel very likely while, at the same time, exposing one or more individuals to skin or facial contamination from hazardous chemicals. The chemistry technicians-indicated that previous requests for a design change were not approve The inspectors reviewed Material Safety Data Sheets 4766 (hydrogen peroxide)
and 5585 (lithium hydroxide monohydrate) which detail the health hazards -
associated with the chemicals. The exposure to lithium hydroxide and hydrogen peroxide poses health hazards to personnel if the chemicals were to contact skin or eyes or were taken internally. Acute exposure of the chemicals to the eyes may cause blindness. The inspectors identified this concern of the chemical addition arrangement to licensee senior managemen '
The addition of the hydrogen peroxide to the reactor coolant system was performed by an auxiliary operator. The operator performed the addition-in accordance with Procedure 50P-104A in_a cautious manner. The field support supervisor observed the job performanc .3 Observation of Licensed and Unlicensed Operators During the period of November 16-19, the inspectors observed licensed and nonlicensed operators during the conduct of shift turnovers, while performing assigned activities, and assessed communications between operators and with supervisory personnel, The following observations were made:
- :Several auxiliary operators-were observed performing various evolutions in the field. All evolutions were conducted' formal'.y with the appropriate procedure, and communication with the control room _was observed to be-timely, clear, and-effective. The field support supervisor was effective in ensuring safe operation of the systems-observed. The inspectors noted, however, that the operators did no always anticipate and verify that their actions resulted in the expected system performance. One instance involved establishing a purge-lineup on the Unit 1 Instrument Air Compressor 1-02. When the activity was finished, the auxiliary operator stated that he had completed the lineup and, therefore, was purging- When asked to verify the correct flow
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path, neither the' auxiliary operator nor the field support super _ visor
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could-successfully describe the flow path. A purge flow path was-
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- Crew turnover briefings, in general, were effective in presenting the oncoming crew with pertinent information; however, interviews with several operators revealed that the oncoming operator relies heavily on a thorough one-on-one turnover from the offgoing operato * Individual turnovers were observed, in general, to be very professional and thorough. The offgoing operators were very conscientious in providing a detailed relief and ensuring that the oncoming operator was ready to accept the shift. In most cases, the oncoming operator was attentive and questioned system statuses he did not fully understan However, one instance was noted, with the reactor core offloaded, where the oncoming licensed reactor operator was not attentive during the one-on-one shift turnover. The relieving reactor operator was observed to simultaneously perform annunciator lamp tests and replace bulbs while the offgoing reactor operator explained the main control board system's status. Immediately after turnover was complete, the inspector questioned the reactor operator about the status of several systems covered during the turnover, lhe reactor operator was not sure of the status of some systems which were not specifically relied on for "No Mode" operatio * In general, when questioned as to the meaning of lit annunciators in the control room, operators were knowledgeable and could explain the cause of the annunciator. There were some instances where the operator was not sure, but the safety significance of the alarm was very low and inconsequentia * The routine interaction between unit supervisors was observed to be frequent and sufficiently detailed to keep each other informed of evolutions affecting the other uni .4 As low As Reasonably Achievable (ALARA) Review The inspectors reviewed the licensee's approach to keeping radiation exposure ALARA. The review consisted of evaluating Procedures STA-651, Revision 4,
"ALARA Program,"; STA-657, Revision 4, "ALARA Job Planning / Debriefing";
and RPI-607, Revision 2, "ALARA Planning"; and comparing the requirements of these procedures to a refueling outage ALARA package. Radiation Work Permit (RWP) 92-1600, " Refueling"; and RWP 92-1400, "SG Eddy Current Testing,"
were verified to conform to the requirements of the above procedure The ALARA package for the refueling activity was discussed with an ALARA technician to determine the effectiveness of the planning. The inspectors ascertained that the ALARA planning had been effectively accomplished as evidenced by the following:
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- Department ALARA contacts' meeting notes; e Review of Refueling Outage 1 exposure history;
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- Licensee's. discussion with the Callaway Nuclear Station personnel who recently removed a stuck reactor stud;
- An ALARA technician visited Babcock and Wilcox Company to observe the' ,
stud removal equipment,-interview the contractors involved in the operation, and determine placement of equipment and personnel; and
- Licensee's discussions with Westinghouse personnel and onsite_
representative to estimate time and equipment needed to complete each refueling tas The inspectors noted that no "ALARA Concerns / Suggestion" forms had been filled out during Refueling Outage 1 for the refueling activity. The ALARA technician did indicate that several suggestions from radiological protection technicians did factor into the present RWP (i.e., combine all-the refueling activities into one RWP versus several, due to past confusion on which task certain activities would be tracked).
Overall, the inspectors found the ALARA package preparation to be thorough, with excellent foresight and preplannin .5 Plant Housekeepina On November 21 while performing a walkdown of the radiological controlled area, the inspectors found a bag of potentially contaminated protective clothing inside of Unit 1 Room 1-088 that was not controlled in accordance with Procedure STA-652, Revision 4, " Radioactive Material Control." The used 3 clothing was contained in a radioactive material bag outside of any-contamination area and was unsealed and not marked as to the radioactive contents. Radiation protection personnel.were immediately notified and a survey of the contents indicated less than 100 counts above background. The bag was then sealed and removed-from the-area. -Conversations with the area I radiation protection technicians' indicated that no work was_ ongoing'in Room 1-088 that required protective clothing. The onshift radiation-protection supervisor was contacted and he-indicated that no other deficiencies with radioactive material control had been -identified and that the issue would be discussed with radiation protection technicians to verify that their areas were properly controlle .6 Conclusions
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The operators performed very well during plant startup and shutdown for the l refueling outage. Communications between the reactor operators was generally good. However, an instance was noted where the operator assuming licensed:
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L _ activities was not fully cognizant of plant statu It was also= noted that managements' expectations were not met for self-verification while placing an instrument' air system in purge.
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-16-The ALAP.A program was well implemented for refueling outage work activitie One instance of improper control of radioactive materials was noted. General ,
area housekeeping within the radiologically controlled area was goo .
5 MAINTENANCE OBSERVATION (62703) ;
5.1 Control of Tools Used on Stainless Steel Ouring a tour of the radiological controlled area on November 2, the inspectors observed work on two containment spray valves (CT-0026 and CT-0028). The maintenance technicians were removing body-to-bonnet studs and nuts to inspect for boron induced corrosion and to remove the deposits from the hardware. The technicians were using wire brushes that appeared to be carbon steel. A quality control inspector was observed allowing the technicians to use the carbon steel brushe The next day the inspector talked to the quality control inspector, involved in the previous day's work, about the use of carbon steel brushes on stainless steel components, lie stated that the use of the brushes was acceptable as long as they were color coated orange to indicate for use on stainless steel onl On November 4 the inspectors identified that the maintenance department-administrative procedures precluded the use of carbon steel brushes on austenttic and/or nickel alloy steels. . Procedure STA-612,. Revision 2, " System Cleanliness Control _and Cleaning, Section 6.7.2, requires austenitic
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stainless and/or nickel alloy steels be wire brushed with stainless steel brushes. The use of carbon steei tools can cause contamination of the stainless steel components by imbedding carbon steel in the stainless steel and, thereby, causing accelerated corrosion due to galvanic actio After the requirements were identified, the inspectors returned to the same e work area and found maintenance technicians using what appeared to be carbon-steel wire brushes. Conversations with the craft personnel and maintenance supervisor indicated that they understood-the use of carbon steel brushes was acceptable as long as appropriately color coded. An operations manager was immediately contacted and the licensee suspended work on those components pending resolution of the inspectors findings. The licensee initiated ONE Form 92-1140 to evaluate the affects of- stainless steel contamination. The licensee verified that the' brushes being used were carbon steel. The licensee immediately removed all carbon steel brushes color coded as stainless from thej work area and the Unit 1 radiological controlled area tool-issue room. The-inspectors identified the failure to control carbon steel brushes in accordance with Procedure STA-612 as a violation of Technical <
Specification 6.8.1 (445/9247-02).
On November 4-the inspectors noted that tools used on stainless and carbon steels'were not segregated in the Unit 2 radiological controlled area tool-issue room. Maintenance Department Procedures MDA-405, Revision 3, " Control of Tools in the Radiological-Controlled Area"; and MDA-401, Revision 1,
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" Control of Tools," requires tools used on stainless steel shall be_ kept-segregated from tools used on carbon steel. .These_ tools include wire brushes, files, grinding discs, polishing wheels, et The licensee was'promptly >
informed of the tool segregation issue. However, on November 9,'the inspectors identified files improperly segregated in _the Unit I containment tool issue room. The inspectors identified the failure to properly segregate tools used on stainless steel and carbon steel, in-accordance with Procedures MDA-401 and MDA-405, as a second example of _ violation 445/9247-0 The inspectors were informed by the mechanical maintenance manager-that the contract supervisor for the Unit I containment tool room was aware of the tool issue and segregation requirements. However, the requirements were not know by the tool room attendants. The licensee immediately stopped issuance of tools and trained all contractor personnel on the requirements for tool control. The licensee removed all nonsegregated-tools from the tool room and restocked with new tools. The inspectors were concerned with the failure of numerous organizations to recognize the improper use and control of tools on stainless steel component .2 .,esel Generator Maintenance The inspectors observed two maintenance work activities on the Unit 1-Train A diesel generator. Contractor personnel and licenste quality control technicians were observed performing the disassembly of Cylinder Heads 7R and 8R using Work Order 3-92-318428-01 (nondestructive examination.of cylinder blocks and liner). The inspectors also observed the boroscopic inspection of various diesel generator and cylinder liners conducted in accordance with Work Order 3-92-322179-0 The work was conducted in accordance with the applicable maintenance -
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procedures. Contractor and licensee personnel;were observed using procedures throughout the maintenance.- Each work step was signed-off after completion of-the work activity. Good work practices were used by-the-craft personne .3 Chemical and Volume Control System Drain Line -
The inspectors observed the work activity associated with a-chemical and volume control system drain valve. This activity was conducted inside the reactor containment building, in an area designated as.a high radiation / contaminated area- The activity was performed in accordance with
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the work order instructions. During the performance of the maintenance activity, the inspectors noted several poor radiological work practice These included positioning the- safe _ty glasses with the palm _of the glove and
reaching inside the anticontamination clothing while inside the contaminated area. A review of the radiation work practice concerns is-provided in-paragraph , __ _, _ _ - _, . - - - _ , _ _ _ _ _ _ _ __
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-18-5.4 Conclusions One violation was identified for the use of carbon steel brushes on stainless steel components and the failure to segregate tools used on' stainless steel from those used on carbon steel. This reflected a lack of awareness by plant personnel to control materials used on safety-related component The work activities performed on the diesel generator and chemical and volume control system were conducted in accordance with the work instruction SURVEILLANCE OBSERVATIONS (61726)
The inspectors observed the surveillance testing of safety-related systems and components listed below to verify that the activities were being performed in accordance with the Technical Specifications. The applicable procedures were reviewed for adequacy, test instrumentation was verified to be in calibration, and test data was reviewed for accuracy and completenes The inspectors ascertained that any deficiencies identified were properly reviewed and resolve The inspector witnessed portions of the following surveillance test activities:
6.1 Diesel Generator Reverse Power Relav Calibration The inspectors observed two technicians performing the calibration of the Unit 1 Train A diesel generator reverse power relay. The calibration was authorized by Work Order 3-92-318811-01 and conducted using Procedure MSE-G0-0020, Revision 1, " Relay Calibration," Attachment 1 The inspectors noted that.the technician assisting _in the calibration was an instructor from the training department and was being trained on meter-and relay calibration techniques. Excellent coordination between the two technicians was exhibited throughout'the inspectors' observation. The supervising technician ensured that the other technician understood the importance of each check and adjustment. Good self-verification and work -
practices were employed by the craft ~ personne .2 Train B Centrifugal-Charging Pump Test The inspectors observed portions of surveillance testing of the Train B centrifugal charging pum The surveillance was conducted in accordance with Procedure OPT-201A, Section 8.2.2, and Work Order 5-92-501811-A The inspectors noted that the pressure gauge installed to measure the charging pump discharge pressure was _ leaking-water. A-swipe of the water was taken by a radiation protection technician and it indicated no activities above-background. The leaky gauge was investigated by instrumentation and control personnel. The investigation revealed the gauge was _ cleaned prior to the test and the water had not properly _ dried during the cleaning process. The 1icensee concluded that the fluid did not affect the -operability of the gaug The water was removed and the test was satisfactorily completed. No other T
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-19-discrepancies were noted duri_ng initial system lineup and procedural testing.
The inspectors reviewed all test data from the surveillance. The inspectors verified that the pump vibrational amplitude was acceptable and that the pump charging flow was above the pump curve.
The inspectors noted excellent communication between the control room operator and auxiliary operators. Overall, the surveillance was conducted with good coordination and control.
6.3 Conclusions
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The licensee effectively implemented the surveillance program in accordance with procedures. Briefings were conducted prior to the performance of each surveillance test. Excellent communications were noted between the control room and auxiliary operator ENGINEERED SAFETY FEATURES (ESF) SYSTEM WALKDOWN (71710) .
The inspectors conducted an ESF walkdown of the radiation monitoring system during the previous inspection period. - The results of the ESF walkdown are documented in NRC Inspection Report 50-445/92-40; 50-446/92-40, paragraph Based on the design modifications which are scheduled to be implemented, the
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inspectors determined that a review of these design modifications was neede .1 Design Modification DM 91-901: "Uporade-of Radiation Monitorina S_vstem Central Processor (CPU) and RM-ll Displav Generators" This design modification upgraded the radiation monitoring systems compute The license had determined that the present system was essentially obsolet This was based on replacement parts not being available. The system was also deemed unreliable causing loss of data and loss of communication with approximately 90 radiation monitors. Finally, the addition of the Unit 2 radiation monitors would have added to the communication proble The upgrade will require a change to the facility as described in the licensing basis document. A 10 CFR Part 50.59 safety evaluation was performed to evaluate this change. Safety Evaluation 92-144 concluded that no accidents and malfunctions of equipment important to safety, as described in the licensing basis document, would be affected by implementing this modificatio It also concluded that no new unreviewed safety questions exist due to the upgrade. The inspector reviewed the safety evaluation and concluded that the licensee had evaluated all concerns pertaining to the new computer system to define the credible potential failure modes.
-.The inspector reviewed the documentation provided to support the design _ _
modification. It was noted that the licensee had identified the applicable-station, operating, and abnormal procedures to be revised. The use of operator aids were also identified to clarify operation of the new operator console. Training was also to be given to the operators concerning the major
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changes to the system. The applicable sections within the Final Safety _
Analysis Report were identified that were required to be update .2 Desian Modification DH 91-148: " Containment PIG pumn Renlacement" ,
This design modification provides for replacing Roots AF-22 blowers on radiation monitors:
1-RE-5502/66/03 X-RE-5568A/75A/67A X-RE-5568B/75B/678 X-RE-5895A X-RE-58968 i The design modification authorizes the replacement of the Roots blowers with Thomas pump The licensee has concluded that the Roots blowers are obsolete and are no longer supported by the radiation monitor vendor. The Roots blowers also required frequent maintenanc Since the design modification changed the facility as described in the licensing basis document, a 10 CFR Part 50.59 evaluation was performe _
Safety Evaluation 92-153 concluded that the new equipment met the same qualification and installation requirements as the existing equipment and als would be' powered from the same source. The evaluation concluded that the operation of the monitors would remain the same and no-new failure modes were introduce l The inspectors reviewed the licensee's safety evaluation and associated-documentation to implement the design modification. The inspectors concluded _"
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that the licensee had conducted an extensive analysis and the conclusions were well supported. The inspectors verified that the licensee had identified the revisions that were required to' the Final Safety. Analysis Report, operatin procedures, and maintenance procedures, including the preventive maintenance-procedures. Based on this review, the inspectors concluded that the licensee had performed an appropriate review to replace the Roots blowers with Thomas-pump .3 Conclusions j
The design modification process was' appropriately implemented to enhance the _- 'j radiation monitoring system. The system engineer was fully cognizant of the i
design changes and was extensively involved with the design change. _The affected procedures. wore identified for revision, j
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-21-8 REVIEW 0F TEMPORARY INSTRUCTION 2515/113 (Closed) 11 2515/113: * Reliable Decay Heat Removal" The licensee implemented measures to assure reliable decay heat removal during-outage This assurance was provided through outage risk assessment, procedure controls, and operator trainin During the development of the outage plan, the licensee identified critical systems and power sources which were required to be maintained operable. A risk assessment was performed to evaluate the overall affect equipment outage windows would have on required safety-related systems and power sources. This assessment accounted for decay heat, plant mode, and reactor coolant system inver tory . Specific criteria were established, including a formal documented review by the Independent Safety Engineering Group, to modify the-equipment-outane window ties. This was to assure that the required safety-related components and power sources remained operable. This also assured that-equipment which was to be relied on, but was not required to be operable by
- the lechnical Specifications, remained availabl The licensee had implemented Integrated Plant Operating Procedure IPO-010A,
" Reactor Coolant System Reduced Inventory Operations." This procedure provided the steps necessary to drain and fill the reactor coolant system for maintenance or maintain steady conditions during reduced reactor coolant system inventory operations. The inspectors reviewed the procedure and found that it provided specific requirements for reduced inventory operations. The procedure also provided appropriate procedure references to transition to l other refueling or operating procedures.
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I The inspectors discussed the procedure implementation with licensed personnel.
l The individuals were provided specific training on reduced inventory operations and were cognizant of the procedural requirements. The operators demonstrated that they were knowledgeable of the equipment which was required to be operable and the instrumentation to be utilized. They also were cognizant of which abnormal procedures may be require .2 Conclusions j The licensee has implemented an effective program for assessing outage. risks.
l This program was well supported by plant management. Procedural controls for i reduced inventory operation were very good. Personnel responsible for implementing the procedure were knowledgeable of the procedure requirements l and had received appropriate training to properly control reduced inventory
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operations.
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p-22-9 FOLLOWUP (92701) (Closed) Inspection followup item 445/9019-05: 446/9019-05: Development of acceptable proceduralized strike plans The licensee has evaluated the need to develop a written strike contingency plan. Based on their review, the licensee does not intend to develop a detailed written contingenc The inspectors will continue to monitor for a future need for the licensee to develop proceduralized strike plan .2 (Closed) Unresolved item 445/9220-07: 446/9220-07: Sufficient shift-manning of licensed senior operators for two-unit operation The licensee has implemented neveral permanent steps to provide reasonable confidence that shift manning for dual-unit operation is appropriat The inspector noted that minimum shift crew composition as defined in Procedure 00A-102, Revision 14, "Cnnduct of Operation," Attachment 8.A, page 1 of 3, required four senior reactor operators and the field + pport supervisor when both units are in Mode 4 and abov The additional s;nior reactor operator was being utilized to reduce the administrative burden on the assigned unit supervisors, especially in the area of work control. The inspectors noted that maintenance administrative activities-have been lessened on the day shift by increased maintenance suppo,t on-the night shift.: These-initiatives have collectively-reduced the administrative burden on the unit supervisor and allowed 9+ aata.' supervisory interfact. and oversight on the part of the unit supeiviso .3 (Closed) Unresolved Item 445/9240-01: Steam generatot atmospheric block valve unauthorized repair The licensee formed an evaluation team to address ONE Form-92-965 This ONE- ;
form was written to evaluate and assess the issues = involved with the repair to.-
the steam generator atmospheric block valve reach _ rod as detailed in NRC-Inspection Report 50-445/92-40; 50-446/92-40. The reach rod repair _was determined not to have been authorized and was in violation of-Procedure STA-606, Revision 18, " Work Requests and Work Orders." Section requires that a work order-be completed prior to commencing maintenance except for emergency or-urgent work. No work order was initiated before or after the
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work was conducted. In response to the NRC concerns, the licensee action were as follows:
- Technical Evaluation 92-2176 documented that the valve was operable .;
without the duct- tape installe The operations department will. issue'a lessons learned to address management's expectations for acceptable work
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practice * . Based upon Technical Evaluat' ion'92-2176, the priority assigned was consistent with management's expectations; however, a review of the p:w.ess for reassigning work requests _into open work orders will _be conducted to ensure correct prioritizatio t I
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e Reemphasize that each manager who is responsible for verifying programmatic concerns which effect their department is identified on a ONE For This violation was of minor safety significance and the licensee has performed an extensive evaluation and formulated correctiva actions. The violation is-not being cited becau;e the criteria specified in Section Vll.B.1 of the NRC Enforcement Manual were satisfie .4 (Closed) Unresolved item 445/9231-02: Testino of the centrifugal ,
chargin._q pump alternate miniflow relief valve
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The inspectors, along with a Region-based inspector and an NRR-representative, witnessed a test of the centrifugal charging pump alternate miniflow relief valve for Unit 2. The test was performed in response to the problems identified during the Shearon-Harris testing of the charging pump' alternate miniflow relief valve. Shearon-Harris had conducted a test of the alternate miniflow system in September _1992 to examine relief valve performance _due to concerns that the NRC had expressed .with the reliability of the safety injection alternate miniflow system. The Shearon-Harris "B" charging pump miniflow system exhibited significant relief valve chatt.or which resulted in rupture of the relief valve bellows, it appeared that the chattering was caused by the throttling of the_ relief valve inlet flow by the-initial opening / closing of the associated isolation valv The licensee developed a Unit 2 Startup Test Procedure 2Cp-ST-49-09, Revision 0, " Alternate Miniflow Verification," This test performed a-functional test of the centrifugal charging pump alternate miniflow relief
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valve (2-8510A). The procedure specified that the acceptance criteria for-the-relief valve were that it would not demonstrate any significant vibration or chattering under full-flow condition The centrifugal' charging pump suction flow path for this test was established from the refueling water storage tank.with the volc'e control tank path
! isolated. 'The charging pump miniflow valves were kept open throughout-ths
- tes The pump discharge valve was closed and the pump was' started on normal l
mi n i fl ow.- - After the normal miniflow was stabilized,- the alternate miniflow
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i isolation valve was. opened from the control room. -This provided pump-discharge pressure to the-alternate miniflow relief valve. During the test both the normal miniflow itne and the. alternate miniflow line were open.-
L The inspectors noted that no significant vibration or chattering was observed i
l at the valve or piping throughout the duration-of the test. The thr'ottling of l
the relief valve flow by the opening / closing of- the associated isolation valve
- did not' appear to cause any valve chatte Instrumentation including strain
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gages and accelerometers were installed on the valve to measure any movemen The inspectors reviewed the Alternate Mini Flow Test Report PTR-45, Revision 0, dated November 23, 1992, The report concluded that no significant l
valve chatter or water hammer effects occurred. There were no visual changes
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to the ' relief valve or the alternate minimum flow line during- the tes Posttest- analysis showed that there were no significant-pipe strains or vibrations measured at any point. Based on the: inspectors' observation of the test and a review of-the test data, the inspectors concluded that th i licensee's conclusion regarding system performance was well founde j
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ATTACMMENT 1 1 PERSONS CONTACTED 1.1 TO ELECTRIC M. R. Blevins, Director of Nuclear Overview D. Buschbaum, Supervisor, Compliance W. J. Cahill, Group Vice President, Nuclear Engineering and Operations R. Carter, Assistant to Manager, Maintenance D. L. Davis, Manager of Technical Support N. Harris, Licensing Engineer T. Hope, Site licensing J. J. Kelley, Plant Manager D. McAfee, Manager, Quality Assurance J. W. Muffett, Manager of Design Engineering- -
S. S. Palmer, Stipulation Manager B. Prince, Radiation Protection Manager S. Smith, Unit 1 Work Control Center Manager J. E. Thompson, Senior Engineer, Licensing 1,2 CITIZENS ASSOCIATION FOR SOUND ENERGY (CASE)
O. L. Thero, Consultant The personnel listed above attended the exit meeting. In addition to the <
personnel listed above, the inspectors contacted other personnel during this inspection perio EXIT MEETING An exit meeting was conducted on November 23, 1992. During this meeting, the inspectors reviewed the scope and findings of the report. The licensee did not identify as proprietary any information provided to, or reviewed by, the -
inspectors.