IR 05000277/1987007
| ML20213G613 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 05/11/1987 |
| From: | Gallo R, Linville J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20213G534 | List: |
| References | |
| 50-277-87-07, 50-277-87-7, 50-278-87-07, 50-278-87-7, IEIN-86-106, IEIN-87-008, IEIN-87-8, NUDOCS 8705180368 | |
| Download: ML20213G613 (64) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Report Nos. 50-277/87-07 & 50-278/87-07 Docket Nos. 50-277 & 50-278 License Nos. DPR-44 & DPR-56 Licensee: Philadelphia Electric Company 2301 Market Street Philadelphia, Pennsylvania 19101 Facility Name: Peach Bottom Atomic Power Station Units 2 and 3 Inspection At: Delta, Pennsylvania Inspection Conducted: February 1 - March 13,1987 Inspectors: T. P. Johnson, Senior Resident Inspector J. H. Williams, Project Engineer R. J. Urban, Resident Inspector A. A. Weadock, Radiation Specialist A. G. Krasop los eactor Engineer Reviewed by: h/'. Ofrfville, E i 0, f
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' /Dat'e R actor Projects ection 2A ivision of Reactor Projects Approved by: b R. M. Gallo, Chief
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Reactor Projects Branch 2 Inspection Summary: Routine, on-site regular and backshift resident inspection U57hoursUnit2;298hoursUnit3)ofaccessibleportionsofUnit2and3, operational safety, radiation protection in-depth review, physical security, control room activities, licensee events, surveillance testing, refueling and outage preparation activities, maintenance, and outstanding item Results: An in-depth review of the licensee's radiological controls program was performed. A violation of posting requirements for high radiation areas was noted (Detail 9.3). In field implementation of radiological controls remains poor. Weaknesses were also noted in the areas of radiation work per-mits, performance of surveys, use of alarming dosimeters, quality assurance audit deficiency correction, and Operations Department response to Health Physics deficiency reports. A fire occurred in the Unit 3 turbine building on March 4. 1987. A review of the preparation for the Unit 2 refueling outage determined that the licensee has adequate planning activitie PDR A E K 05000277 ,
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DETAILS
1.. Persons Contacted
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J.' W. Austin, Superintendent, Construction Division
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8. L. Clark, Administrative Engineer-J. B. Cotton, Superintendent Plant Services J. K. Davenport, Supervising Engineer Maintenance G. F. Dawson, Maintenance Engineer
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A. B. Donell, Nuclear Operations QA
- R. S. Fleischmann, Manager, Peach Bottom Atomic Power Station A. A. Fulvio, Technical Engineer i *A. E. Hilsmeier, Senior Health Physicist
- J. P. McElwain, Nuclear Operations QC J. F. Mitman, Radwaste Engineer
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D. L Oltmans, Senior Chemist C. B. Patton, Supervising Engineer Field Engineers
- J. M. Pizzola, E&R QA F. W. Polaski, Outage Planning Engineer S. R. Roberts, Operations Engineer 4 *0. C. Smith, Superintendent Operations
- J. E. Winzenried, Staff Engineer Other. licensee employees were also contacted.
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- *Present at exit interview on site and for suraation of preliminary
' finding . Plant Status i 2.1 Unit 2 ;
Unit 2 began the inspection period at 80% reactor power with the
- system was repaired, and the pump was returned to service on i February 1,1987. Reactor power was returned to near full powe The A feedwater heater string was removed from service on February
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4,1987, due to an internal leak in the 1A heater (see detail 4.2.1). Reactor power was subsequently limited to 85%.
A load drop occurred to 34% reactor power on February 7, 1987, to perform rod adjustments, to leak check the condenser water boxes, and to rebrush the recirculation MG sets. Power was returned to l 85% on February 9, 1987. On February 16, 1987, load was reduced i to 66% to leak check the condenser. Reactor Core Isolation Cooling (RCIC) was declared inoperable on February 15, 1987, due to packing leak on M0-16 (see detail 8). Load drops occurred on February 17 and 18, 1987, for RCIC MO-16 repair and for condenser leak check l l
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.- 3 On February 23, 1987, a load drop to 28% occurred when the turbine mechanical trip valve failed to indicate reset during testing (see detail 4.2.2). Power was increased to 85% when it was determined that the trip valve had reset. On February 25,.1987, the static inverter / Uninterrupted Power Supply alarmed and a small reactor water level transient occurred (see detail'4.2.3). The unit remained at 85% reactor power until March 13, 1987, when Unit 2
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was shutdown for refuelin .2 Unit 3 Unit 3 began the inspection period at full powe Load drops to 60% power occurred on February 6 and 12,1987, to leak check condenser water boxes. A load drop also occurred on February 20, 1987, to change the control rod patter A fire occurred in a maintenance cage on March 4, 1987, on the 195 foot level of the turbine building (see detail 4.2.4). . Unit 3 was shut down on March 5, 1987, due to hydrogen leak in the main generator. The unit restarted on March 10, 1987, and was synchronized with the grid on March 11, 198 . Previous Inspection Item Update 3.1 (0 pen) Unresolved Item (277/86-13-02; 278/86-14-02). This item was unresolved pending review of maintenance on equipment to determine whether safety related equipment failures were caused by lack of
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lubrication. The lubrication program performed by the plant operations department had been stopped for about two years. The preventive maintenance program covers lubrication of equipment while the equipment or_ system is in operation. An additional issue was licensee action to follow-up on lubrication problems identi-fied while completing the lubrication round sheets. The inspector reviewed the maintenance history in the " Computerized History and Maintenance Planning Syster" file (CHAMPS) for systems 9 and 4 System 9 is secondary containment and system 40 includes all venti-lation systems. The CHAMPS listing contained approximately 370 Maintenance Request Forms (MRFs). After a review of equipment maintenance data, it was determined that no safety related equip-ment failures had occurred which could be attributed to lack of lubrication. Discussions with the licensee indicated the program was still not being fully implemented. Lubrication sheets were not being completed; at times they were not being issued, and follow-up on problems was not being pursued. After further discussions between the inspector and operations engineers, the licensee increased management attention to the lubrication program. The item remains open pending demonstration of satisfactory performance of the lubrication progra ..__ -
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. 4 3.2 (0 pen) Inspector Follow Item 277/86-12-11; 278/86-13-11. Formally establish corporate radiation protection assessment progra Proposed improvements and formalization for the corporate group's assessment of station HP activities have been made and are described in section 9.2. This item will remain open, however, pending approval of a formal procedure describing the progra .3 (0 pen) Inspector Follow Item 277/85-28-03; 278/85-26-03. Review professional training for Health Physics staf The licensee indi-cated that the proposed 60 topic professional training program was being reevaluated and revised. Training is being conducted during this period; however, a course in beta dosimetry has already been presented to the professional staff this year and the following additional courses have been scheduled: Radiation work in High Radiation Areas Health Physics Auditing Advanced Respiratory Protection This item will remain open pending review and formalization of the professional HP training progra .4 (Closed) Violations (278/86-09-01; 278/86-09-03). Failure to follow the correct rod pull program with the Rod Worth Minimizer (RWM) out of service and failure to ensure that a rod was in the correct position when the Rod Sequence Control System (RSCS) was bypassed. The licensee responded to the violations in a letter dated July 23, 1986, and also at an enforcement conference on March 27, 1986. The licensee response states that the violations were caused by a series of personnel errors involving failure of the reactor operator to withdraw the correct control rod, failure of a second operator to verify that the correct withdrawal sequence was being followed, and failure of a Shift Supervisor and Shift Superintendent to verify the proper position of a control rod prior to bypassing the RSCS input for that ro Licensee corrective action included: a manual scram of the Unit 3 reactor; disciplinary action for the licensed operators involved; licensee investigation and an NRC conference call; letters from plant management to operators detailing the event; revision to the affected RSCS, RWM, and startup operating procedures; management permission prior to bypassing the RWM; management meetings with operations per-sonnel; and assurance that when the RWM is bypassed that the second
, licensed operator has no concurrent duties. The inspector verified these corrective actions. Observations of subsequent startups with the RWM out of service have not noted any problems. The Peach Bottom Enhancement Program actions directed toward sensitizing operations personnel to procedure compliance. Based on the above actions, the violations are close . _ _
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. 5 3.5 (Closed) Unresolved Item (278/86-09-04). Control rod drop analysis (RDA) with an out of sequence rod on March 18, 1987. The Philadel-phia Electric Company Nuclear Fuel Management Section and General Electric (GE) Company reviewed the event and determined that the Unit 3 reload analysis RDA results bound this even The various RDA scenarios associated with this event were identified during the review. The worst case RDA results of the scenarios show a peak er.thalpy deposition of approximately 120 calories / gram compared to the RDA design criteria of 280 calories / gram. The 280 calories /
gram design criteria value bounds all currently licensed fuel bundle exposure values. The licensee concluded that the Unit 3 reactor operated in a condition such that RDA consequences would not have exceeded the design criteri The inspector reviewed the licensee's analysis and evaluation, and discussed the analysis with licensee engineers. Based on the above, this item is close .6 (Closed) Inspector Follow Item (277/85-28-01; 278/85-26-01). Review accessibility of the station's Senior Health Physicist to the Station Manager. In the current unit staff organization, the Senior Health Physicist reports to the Superintendent, Plant Services, who in turn reports to the Station Manager. The Senior Health Physicist indi-cated this reporting chain has proven effective for the majority of communications between the Radiation Protection and Station Manager levels. The Senior Health Physicist had the option of reporting directly to the Station Manager. The inspector also reviewed a licensee memo dated February 5, 1987, and titled, " Radiation Pro-tection Manager Access to the Plant Manager". The memo, signed by the Plant Manager, outlines a reporting structure and options con-sistent with that described by the Senior Health Physicis This item is close .7 (Closed) Inspector Follow Item (277/85-28-02; 278/85-26-02).
Licensee to develop position guides for reorganized Health Physics organization. Subsequent to this 1985 review the licensee has performed an additional reorganization within the Health Physics organization as part of the Enhancement Program. Development of position guides is specifically detailed as a task to be completed for the program, with a completion date requirement of July 198 New position guides for each position in the expanded organization have been drafted, reviewed, and in the majority of cases, approve The inspector reviewed selected position guides and noted they ade-quately described position accountability and responsibilitie Based on licensee effort in this area, this specific inspector folics item to track position guides can be considered close Effective licensee completion of milestones for the Enhancement Program, however, will continue to receive review.
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. 6 3.8 (Closed) Unresolved Items (277/85-11-10; 277/85-11-11). Review licensee assessment of external and internal dose received by workers involved in February 10, 1985 incident. On March 11, 1985, a special NRCinspection(No. 277/85-14) was conducted to review the licensee's assessment of external and internal dose received by the workers involved in the events surrounding the 81A valve maintenance activ-ities occurring on February 10, 1985. A complete description of those events is included in NRC Inspection Report No. 277/85-11. NRC review of the licensee's dose assessments indicated the assessments were reasonable, conservative and thorough. The licensee's assess-ments also indicated no regulatory exposure limits were exceeded. By apparent oversight, the unresolved items tracking these assessments were not formally closed out during the 277/85-14 inspection. These items are resolved and are close .9 (Closed) Inspector Follow Item (277/85-11-12). Review Health Physics Technician Staffing. NRC review of problems associated with 81A valve maintenance activities conducted on February 10, 1985, indicated a contributing factor may have been a shortage of HP technicians. During the current inspection, the inspector reviewed licensee staffing levels and proposed HP technician assignments for the Spr.. Unit 2 outage. Licensee technician levels appear adequate to support outage activities (see section 9.1). A draft Unit 2 outage schedule shows seven, eight, and four ANSI qualified technicians to support drywell operations on the
"X", "Y", and "Z" shifts, respectively. Additionally, a reserve pool of ANSI qualified technicians is being maintained each shift to work out of the HP office and supplement staff where require This item is close .10 (Closed) Inspector Follow Item (277/86-02-02). Train personnel in availability of annual exposure reports. The inspector reviewed part of the lesson plan used for General Employee Training (GET-0010) and noted that pen and ink additions were made to clarify how workers could obtain exposure histories. The inspector also verified that this clarification had been formally identified and assigned to an individual to insure its formal inclusion in the next revision of the lesson pla This item is close .11 (Closed) Inspector Follow Item (277/86-02-03). Revise Respiratory Training lesson plan to include equipment malfunction and procedural failure as reasons for relief from respirator use. The inspector reviewed the General Respiratory Training lesson plan and noted pen and ink additions addressing the above had been made. The inspector verified the above addition had been formally identified and assigned to an individual to insure inclusion in the next revision of the lesson plan. The inspector also verified by attendance in the General Respiratory Training class that all reasons for relief from respirator use as given in 10 CFR 20.103 were included in the class material. This item is close *
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'3.12 (Closed) Inspector Follow Item (277/86-12-15; 278/86-13-15).
Review adequacy of communications between HP technician staff and supervision. Communications between the HP technicians and supervision have been improved with the establishment of weekly technician meetings and weekly Communications and Problem Solving (CAPS) meetings (see section 9.1.3).
This item is close . Plant Operations Review 4.1 Station Tours The inspector observed plant operations during daily facility tours. The following areas were inspected:
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Control Room
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Cable Spreading Room
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Switchgear and Battery Rooms
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Reactor Buildings
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Turbine Buildings
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Radwaste Building
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Recombiner Building
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Pump House
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Diesel Generator Building
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Protected and Vital Areas
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Security Facilities
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High Radiation and Contamination Control Areas
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Shift Turnover 4. Control Room and facility shift staffing was frequently checked for compliance with 10 CFR 50.54 and Technical Specifications. Presence of a senior licensed operator in the control room was verified frequentl . The inspector frequently observed that selected control room instrumentation confirmed that instruments were operable and indicated values were within Technical Specification requirements and normal operating limit ECCS switch positioning and valve lineups were verified based on control room indicators and plant observation Observations included flow setpoints, breaker positioning, PCIS status, and radiation monitoring instrument . Selected control room off-normal alarms (annunciators)
were discussed with control room operators and shift supervision to assure they were knowledgeable of alarm status, plant conditions, and that corrective action, if required, was being taken. In addition, the applicable alarm cards were checked for accuracy. The operators were knowledgeable of alarm status and plant conditions.
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4. The inspector checked for fluid leaks by observing symp status, alarms, and pump-out rates; and discussed reactor coolant system leakage with licensee personnel. The Unit 3 unidentified leak rate was frequently verified to be less than the 2.0 gpm limi . Shift relief and turnover activities were monitored daily, including backshift observations, to ensure compliance with administrative procedures and regulatory guidanc . The inspector observed the main stack and both reactor building ventilation stack radiation monitors and recorders, and periodically reviewed traces from backshift periods to verify that radioactive gas release rates were within limits and that unplanned releases had not occurre . The inspector observed control room indications of fire detection instrumentation and fire suppression systems, monitored use of fire watches and ignition source controls, checked a sampling of fire barriers for integrity, and observed fire-fighting equipment stations. A fire occurred in the Unit 3 turbine building on March 5, 1987 (see detail 4.2.4).
4. The inspector observed overall facility housekeeping conditions, including control of combustibles, loose trash and debris. Cleanup was spot-checked during and after maintenance. Plant housekeeping was generally acceptable. Overall improve =ents in plant appearance continue to be note . The inspector observed the nuclear instrumentation subsystems (source range, intermediate range and power range monitors) and the reactor protection system to verify that the required channels were operable. Core thermal limits for both units were verified to be within TS limits by reviewing the process computer P-1 and 00-3 printouts for February 5, 198 .1.10 The inspector frequently verified that the required
.l off-site electrical power startup sources and emergency on-site diesel generators were operabl .1.11 The inspector monitored the frequency of in plant and control room tours by plant and corporate managemen The tours were generally adequat _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ - _ _ _ _ _ _ _ _
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. 9 4.1.12 The inspector verified operability of selected safety related equipment and systems by in-plant checks of valve positioning, control of locked valves, power supply availability, operating procedures, plant drawings, instrumentation and breaker positioning. Selected major components were visually inspected for leakage, proper lubrication, cooling water supply, operating air supply, and general conditions. No significant piping vibration was detected. The inspector reviewed selected blocking permits (tagouts) for conformance to licensee procedure Systems. checked included the Standby Gas Treatment Syste . 4.1.13 The inspectors performed backshift and weekend tours of the facility on the following days during the inspection period: "Y" shift (3 p.m.-11 p.m.) - February 3,4,5,9,10,11,
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- March 3,4,5,6 "Z" shift (11 p.m.-7 a.m.) - February 2,10,11,12,13, 17,19,20,26
- March 2,3,4,6,9,12,13 "X" shift (7 a m -3 p m
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No violations were identifie .2 Follow-up On Events Occurring During the Inspection 4. Unit 2 Feedwater Heater Leak
i On February 2,1987, at about 5:00 a.m. , a high level alarm occurred on the 1A feedwater heater (FWH) shell side. The
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high level condition was confirmed by checking the local sight glass. The FWH shell side drains to the main con-J(q $ r denser through the "A", drain cooler. The drain cooler flow
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condensate occurred. The licensee concluded that an inter-nal leak from the 1A FWH tube side (condensate water) to
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the shell side had occurred. To isolate the leak, the
"A" FWH string was removed from service at 8:15 p.m. on February 4, 1987. Reactor power was subsequently limited to 85%. The resultant feedwater temperature was 351 degrees The 1A FWH had been scheduled for replacement during the Unit 2 refueling outage (Modification #1790) due to FWH degradatio .
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. 10 The inspector reviewed procedure S.7.1.E, "Taking a Heater String Out of Service", Rev. 1; Unit 2 Core Reload #6, June 1984; alarm card 206L#16; and FSAR section 14.5.2. The implementation of procedure S.7.1.E was verified by control room observations, operator interviews, and log review Operating with a heater string out of service was also discussed with licensee engineers and operators. The inspector determined that the 85% reactor power limit was imposed by plant management based on the following:
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GE turbine generator limit with heater string out of service (95%)
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condensate flow through two heater strings (90%)
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reactor feed pump suction pressure limit of 400 psi No unacceptable conditions were note .2.2 Unit 2 Turbine Mechanical Trip Valve Failure At 9:30 p.m. , on February 23, 1987, the main turbine electro-hydraulic control (EHC) system mechanical trip valve failed to indicate reset during routine test RT 5.4,
" Mechanical Trip Valve", Rev.1. RT 5.4 is performed weekly to verify the ability of the turbine mechanical overspeed mechanism to function. The test is not required by Technical Specifications; however, it is recommended by the GE turbine vendor manual. The test actuates the lock out valve and bypasses the mechanical trip valve. This allows a mechanical trip valve test without actually trip-ping the turbine generator. When the valve failed to indicate reset, the operator performing the test maintained the test push button depresse Releasing the test button with the trip valve in a tripped state, would cause a tur-bine trip and reactor scram (reactor power was at 85%).
The licensee reduced power to 28% and began troubleshooting the EHC circuitry. During the power reduction and trouble-shooting, the test push button was maintained depressed by operators. The licensee determined that the R3-D19 relay in panel 20C30 was not resetting, resulting in the indi-cation at the EHC panel that the trip valve was tripped (i.e., not reset). The trip valve was actually rese The relay was repaired, and EHC test indications returned to normal at 1:00 a.m. on February 24, 1987. RT 5.4 was subsequently performed three times with satisfactory result Reactor power was increased to 85%.
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. 11 The inspector reviewed RT 5.4, control room logs, and EHC electrical schematic drawings. Discussions were held with licensee engineers and the operators involved with the actual test. The inspector determined that the. turbine mechanical overspeed trip function was unavailable; how-ever, the electrical overspeed trip function was availabl Actuation of the local mechanical manual trip or the backup overspeed trip (electrical) would de-energize the lockout valve, making the mechanical trip valve available for turbine tri In addition, all other turbine generator trip functions remained functional. The turbine control valve fast closure and turbine stop valve closure scram functions were not affected. The inspector had no further questions. No violations were identifie .2.3 Unit 2 Static Inverter Fluctuations and Troubleshooting On February 25, 1987, at 12:55 p.m., with Unit 2 at 85%
power, a static inverter trouble alarm and B RFP low flow alarm were received in the control room, and spikes were recorded on the following instruments:
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LR-2-6-96, RPV level
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NR-2-7-45A, B, C, D, APRM
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FR-2-6-89, individual RFP flows
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FR-2-6-98, total steam flow / feed flow
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generator core monitor
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POR-2620, RFPT control valve position
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TR-2151, feedwater temperatures
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LR-2-3-110A/B, RPV level The only actual plant response was a decrease in reactor water level of about five inches. The cause of this tran-sient was an apparent power fluctuation in the static inverter uninterruptible power supply (UPS), panel 20Y5 Similar problems with the static inverter have occurred and were discussed in NRC Combined Inspection Report 50-277/86-19; 50-278/86-20. Of the eight instruments that experienced spikes, the last three recorders are not powered from the UPS. Therefore, actual reactor water level decrease was demonstrated by LR-2-3-110A/ .
.1 12 Normally, the UPS is powered from the 250 VDC station battery through the static inverte If static inverter output is lost, power from the 440 VAC emergency auxiliary power system (through a 120 VAC transformer)
is automatically supplied to panel 20Y50 through the auto transfer static switch. Retransfer to the static inverter is automatic. Apparently, auto transfer through the static switch did not occur because a relay that monitors the transfer did not orint out on the compute The licensee had a strip chart recorder monitoring the static inverter for several months. However, the recorder was removed several days before the even Over those months, the only abnormalities recorded were several slight voltage drops from the station batterie The Unit 2 static inverter is scheduled to be replaced during the 1987 refueling outage as MOD 1359. The licensee expects these unexplained transients to end once the new static inverter is in place. No violations were noted. The inspector will continue to follow this issu . Unit 3 Turbine Building Fire on March 4, 1987 At approximately 6:40 p.m., on March 4, 1987, smoke was reported on the 195 foot level of the Unit 3 turbine building by a plant worker. Unit 2 was at 85% power and Unit 3 at 100% power. The licensee's fire and damage team responded, and found a fire in a maintenance cage in the turbine building fan room. The fire was extin-guished by the fire and damage team using portable Anzul and carbon dioxide extinguishers. No off-site fire assistance was needed nor requested. The licensee declared an Unusual Event at 7:15 p.m., because of a fire in the protected area lasting more than ten minutes after attempts to extinguish i The licensee made an ENS call and notified the Senior Resident Inspecto The Unusual Event was terminated at 7:25 p.m., after the fire was extinguished. The fire appears to have been electrical in nature in a maintenance storage cage (" Fitters MSIV rebuild cage") causing combustibles to ignite and burn. The damage was confined to the cage area. The area included no safety related equipmen However, it was a posted contamination area (1000 DPM).
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The Senior Resident Inspector responded to the site at 8:25 p.m., on March 4, 1987. The inspector examined the fire area; walked down control room panels; discussed the fire, including its affect on plant operation, with licensed operators and fire team members; and reviewed station logs. No affect on Unit 2 (85% power) or Unit 3 (100% power) operation was noted. Based on reviewing control room radiation monitors, and local survey and air-sample data, no release of radioactivity occurre The inspector verified that no safety related equipment was involved directly or indirectly with the fire. The fire was in a metal cage which housed maintenance and test equipment for MSIV manifold repairs. The cage was divided and in the section adjacent to the fire, and the inspector noted an acetylene bottle marked flammabl A regional specialist reviewed the circumstances relating to.the fire on March 5, 1987. The inspector surveyed the scene of the fire to determine its origin and possible impact on safe shut down. The inspector determined that the most probable cause for this fire was a fan installed in the area for personnel comfor The fan was probably left "on" for an extended period of time, and the motor overheated. The ensuing fire or sparks set off a secondary fire in combustible materials left under the fa The review of the area and of the fire hazards analysis determined that there are no safe shutdown or safety related components in this area. However, the inspector observed a welders cart with an acetylene bottle about four feet from the fire origin. Neither the fire hazards analysis nor the pre-fire strategy plan procedure (F Pro-cedures) indicated the presence of this cart in the are The inspector stated that the reason for the pre-fire strategy plan procedure is to identify to the fire brigade the possible hazards in the fire area so that the proper fire attack is planned. Similarly, the fire hazard analysis should identify the combustibles in plant areas that may present a hazard to safety around the building and equipment. The licensee agreed that the acetylene cart should not have been placed ther The licensee also agreed to check whether any more fans of the type involved in the fire are in use at the facilit No violations were note .-. - -. _
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4.2.5 Unit 3 Shut Down Due to Generator Hydrogen Leakage Early on March 5, 1987, the licensee determined through operator rounds and control room indications that a hydrogen leak had developed ,in the Unit 3 main generator hydrogen cooling system. Although monitoring of the power block did not reveal any hydrogen accumulations, the licensee could not locate the leak and was not cer-tain where the hydrogen was going. Initial speculation was that the hydrogen was leaking into the generator seal'
oil system past a faulty hydrogen seal. This theory was based on foamy seal oil at the No.10 bearing (exciter side) and hydrogen levels at the turbine building roof vent from the air detaining tank twice as high as Unit 2.
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After nine bottles of hydrogen were used in fifteen hours, the hydrogen usage rate approached 43,200 SCFD (normal r- rate is 1200-1400 SCFD), and the leak could not definitely be located, the licensee decided to shut down Unit At 4:35 p.m., on March 5, the licensee began reducing recirculation flow and inserting control rods using General Plant Procedure GP-9-3, " Fast Reactor Power Reduction". At 5:48 p.m., the Unit 3 operator manually scrammed the reactor (mode switch to shutdown). The operators began using Transient Response Implementation Plan Procedure T-100, " Scram", and T-99, " Post Scram Implementation". The inspector was in the control room to witness the shutdown, and no problems nor deficiencies were observe On March 6, 1987, the generator hydrogen cooling system was pressurized with helium in preparation for leak testin The licensee discovered that the hydrogen was leaking through a 1/4 inch " Whitey" vent valve; the hydrogen was escaping through a vent on top of the tur- !
bine building roof and not through the hydrogen seal as l previously thought. The valve was replaced and Unit 3 was started up on March 10, 1987. The inspector observed portions of the startup including turbine generator startup and synchronization in accordance with GP-2, " Normal Plant Startup" and S.6.3.1.A, " Main Turbine Generator Startup",
Rev. 19. No unacceptable conditions were note Post startup measurements of hydrogen usage in the Unit 3 generator of approximately 500 SCFD are much lower than seen by the licensee in a long time. The licensee attributes this reduction to the use of helium rather than Freon to locate leaks; many small leaks were found and repaired while looking for the large leak. The
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< 15 licensee is planning to use this same method on the Unit 2 generator before its startup after the 1987 refueling outage. No violations were note .3 Logs.and Records The inspector reviewed logs and records for accuracy, completeness, abnormal conditions, significant operating changes and trends, requi ed entries,. operating and night order propriety, correct equipuent and lock-out status, jumper log validity, conformance to Limiting Conditions for Operations, and proper reporting. The following logs and records were reviewed: Shift Supervision Log, Reactor Engineering Logs, Unit 2 Reactor Operator's Log, Unit 3 Reactor Operator's Log, Control Operator Log Book and STA Log Book, Night Orders, Radiation Work Permits, Locked Valve Log, Maintenance Request Forms, Temporary Circuit Modification Log, and Ignition Source Control Checklists. Control Room logs were compared against Administrative Procedure A-7, Shift Operations. Frequent initialing of entries by licensed operators, shift supervision, and licensee-on-site management constituted evidence of licensee review. No unacceptable conditions were identifie I 4.4 Unit 2 Refueling Outage Preparations On March 13, 1987, Unit 2 was shut down for a scheduled 71 day
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refueling outage. Major work items include refueling, 10 CFR 50 Appendix R modifications, 10 CFR 50 Appendix J testing, turbine maintenance, plant modifications, and other maintenance and testin . Refueling Outage Organization The licensee has organized the management of Unit 2 refueling outage similar to last Unit 3 refueling (1985-86) with some enhancements. The outage is divided into four areas each headed by a supervisor (area coordinator). These areas include: refueling floor, reactor systems, drywell, and balance of plant. These area coordinators are responsible for work planning and tracking, and for conducting daily status meetings. In addition, a shift outage director has been assigned to cover 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day and six days a week. The shift outage directors are responsible for directing outage activities on their assigned shifts. They are responsible for making decisions and coordinating resources in order to keep the outage on schedul The critical path for the outage is the refueling fuel floor activities. These jobs are to be man-loaded on a seven day n per week basis. Other jobs are being worked six days a week, with two ten hour shifts per da . ,_ __
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Operations has dedicated five SR0s to provide around the clock coverage for permit blocks, MRFs, and fire watch coordination. On the back shift, this SRO will be the Outside Shift Supervisor. During day shift, the Outside Shift Supervisor will be manned by a utility SRO. This allows the Inside Shift Supervisor (control room SRO) to concentrate on the operating unit. Operations has also dedicated ten shift blockers to assist in equipment i tagouts. These shift blockers are plant operators !
(non-licensed).
Periodic meetings to discuss outage progress are as follows:
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three times a day conducted by the shift outage directors
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daily conducted by the area coordinators
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daily conducted by plant management
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daily conducted by outage management
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twice a week among outage management, construction, maintenance, and plant management
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twice a week between plant manageri.-t and corporate vice presidents
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weekly with the NRC and outage managemen Each of the area coordinators has been assigned a HP tech-nician for radiological controls support. In addition, each maintenance craft (i.e., pipe fitters, electricians, machinists, etc.) has HP technicians assigned two shifts per day. These HP technicians are to assist the crafts in RWP and work coordination.
I
- The inspector reviewed the organizational structure and discussed it with licensee outage personnel. The inspector
! will continue to follow the effectiveness of this organi-zation including the changes made during the refueling outage perio .4.2 Work Scope The construction division has been assigned 66 plant modifications. The majority of these are electrical and implement 10 CFR 50 Appendix R requirements. The construc-tion division has contracted over 500 Catalytic employees
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.:. :
. 17-
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to assist in these modifications. Major modification work to be performed includes:
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1A feedwater heater replacement
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RFP minimum flow valves and piping changes
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permanent safety grade air supply installation
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install ESW header isolation valves
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upgrade Recirc MG sets sprinkler system
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torus attached piping support upgrade The inspector discussed Construction division work lists, equipment procedures, craft man-loading, schedules, and organization with the on-site Superintendent. The inspec-tor also reviewed the outage work list and selected implementing procedure The station maintenance division has been assigned all major corrective and preventive maintenance (PM). The work will be performed by the on-site maintenance crafts, supplemented by off-site mobile maintenance groups and over 500 Cata-lytic contractors. Fuel floor work is scheduled 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day, seven days . weck. Other maintenance work is scheduled to be covered 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> a day (two shifts), six days a week. Maintenance division work includes:
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reactor vessel disassembly and" assembly
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control rod drive changeout and maintenance
,
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snubber inspection and rebuild
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HPCI and RFP turbine inspections
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turbine generator inspections
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valve corrective maintenance
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electrical inspections, maintenance, and PMs
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main condenser inspection i
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_ . 18 The inspector discussed Maintenance division work lists, craft man-loading, schedules and organization with the on-site Supervising Engineer. The inspector also
. reviewed the outage work list and selected implementing procedure The Field Engineers (FE) have been assigned all modifica-tion electrical inspections and RCIC/HPCI MOVATs testin A group of 20 engineers reports to an on-site Supervising Engineer. Work activities include wiring checks and electrical construction hsting. The inspector reviewed FE work lists, schedules, organization, implementing pro-cedures and discussed these items with the Supervising Engineer, Field Engineer The Mobile Maintenance Section, In Service Inspection (ISI)-services is on-site to perform ASME Section XI ISI and balance of plant (B0P) erosion / corrosion inspections (see detail 5.1). This is the first cycle of.the second-ten year ISI interval for Unit There are 16 ISI technicians to perform UT measurements of B0P systems. GE will perform ASME Section XI IS The inspector reviewed the ISI work list, organization, man-loading, and discussed these items with ISI engineers. The inspector also reviewed the ISI plan for the Unit 2 outag . Quality Assurance and Quality Control (QA/QC)
The inspector reviewed the QA/QC organization, staffing and schedules to support the outage and related work activities. Nuclear Operations QA and QC organizations perform surveillances/ audits and monitoring /signoffs for work. The inspector reviewed the Nuclear Operations QA surveillance and audit schedule. QC coverage includes the following activities:
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reactor disassembly
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core alterations
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control rod drive exchange and rebuild
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control rod HCU maintenance i
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snubber inspections
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electrical breaker inspections
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recent inspections
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- 19
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housekeeping inspections
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modification acceptance testing (MAT)
Engineering and Research (E&R) QA performs surveillance and audits on modification activitie Scheduled audits include 10 CFR 50 Appendix R modification work and M0 VATS testing. Construction Division QC performs the quality inspection function for major modification ~
The inspector discussed the planned QA/QC activities with QA and QC supervision. The inspector determined that QA/QC coverage was in accordance with the QA Plan and implement-
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ing procedure No violations were identifie .5 Engineered Safeguards Features (ESF) System Walkdown 14. Core Spray System In combined Inspection Report 50-277/86-24 and 50-278/86-25, the inspector performed a detailed walk-down of portions of the core spray system in order to independently verify the operability of the Unit 2 and 3 systems. In that walkdown, the inspector identified two equipment deficiencies; one concerned a missing handwheel nut on a locked open valve, and the other concerned a detached conduit hose on an inoperable Pylet space heater for the 2 "C" core spray pump motor. In connection with the latter problem, the system engineer was questioned concerning the function of the motor space heaters, if both heaters are needed to perform their function, and the effect of heater inoperability on the operability of the core spray pump. These questions involved research by the system engineer and future follow-up by the inspector.
!
In this report period, the inspector had further discus-sions with the system engineer. The core spray pump motor manufacturer (General Electric) stated that only one of the two space heaters was necessary to fulfill their function during moist environment conditions. Their function is to keep internal electrical components dry to prevent " corona" from occurring during pump starts. Corona is a luminous discharge at the surface of a conductor or between two conductors of the same transmission line, causing insula-tion breakdown with a subsequent reduction in pump motor life. General Electric also stated that the pump would start in a moist environment even without both heater . -. .- - - - - - _ _ _ -
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The inspector toured the core spray pump cubicles to determine if the licensee had corrected the previous deficiencies. The handwheel nut on the locked open valve had been replaced and the conduit hose and space Feater were repaired. The cause of the failed space heater was due to a loose connection causing high resistance and sub- l sequent burn-out. Also, during this tour, the inspector f noted that the 3 "C" core spray pump motor space heater was inoperable. This problem was brought to the attention of the system engineer and a maintenance request form (MRF)
was written to repair the space heater. The failure mode of this space heater was identical to that of the 2 "C" core spray pump motor space heate Based upon the inspector's identified problems and the feedback from General Electric, the system engineer pro-posed adding a step to monthly surveillance tests ST- and ST-6.7 to check if heat is emitting from both heater The system engineer also proposed adding a step to the core spray preventive maintenance procedure to check the tight-ness of the heater lug nuts and to possibly add lock nuts to ensure. continued tightnes In light of the above, the inspector questioned licensee senior management as to extending the above practices to other safety related motors at Peach Bottom. Upon review of electrical schematic diagram E-321, " Miscellaneous Motor Space Heaters", the inspector found that the following j safety related motors have similar space heaters:
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residual heat removal (RHR) pumps,
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high pressure service water (HPSW) pumps,
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emergency service water (ESW) and emergency service water booster pumps,
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high pressure coolant injection (HPCI) auxiliary oil pump, ar.d gland seal condenser condensate and ( vacuum pumps,
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emergency cooling water (ECW) pumps,
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motor driven fire pumps,
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emergency cooling tower fans, and
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diesel generator building vent fan _ _ _ - _ _ - - - .
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. 21 The performance engineer is reviewing this area to deter-mine what actions will be taken for the above motors. The inspector will continue to follow this are . Reactor Core Isolation Cooling (RCIC) System
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The inspector performed a detailed walkdown of portions
'of the RCIC system in order to independently verify the operability of the Unit 2 and 3 systems. The RCIC system walkdown included verification of the following items:
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Review of the documents'11sted in Attachments 2 and Inspection of system equipment condition Confirmation that the system check-off-list (COL)
and operating procedures are consistent with plant drawing + --
Verification that system valves, breakers, and switches are properly aligne Verification that instrumentation is properly valved in and operabl Verification that valves required to be locked have appropriate locking device Verification that control room switches, indications and controls are satisfactor Verification that surveillance test procedures properly implement the Technical Specifications surveillance requirements (Attachment 3).
Overall, the inspector determined the RCIC systems in Units 2 and 3 to be operable. However, minor equipment, procedural, and drawing deficiencies were identifie The inspector determined that four deficiencies were more significant than the others: (1) the RCIC fill supply valve (HV 2-138-21208) on the condensate storage tank (CST) suction line for Unit 2 was found open, (2)
the lube oil level in the rear bearing cavity of the Unit 3 turbine was low, (3) the recirculation lube oil cooler supply pressure root valves (RTV 2(3)-13B-2(3)018)
were not on the COL, and (4) a closed and capped valve (HV-2-13C-52B) was found open with a pressure gauge attache ,
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,. 22 P&ID M-359 shows HV-2-138-21208 closed, and the COL states that this valve should be closed and capped; capping this valve is physically impossible. If the downstream valve leaked by or was inadvertently opened, the CST would begin draining to clean radwaste. The licensee closed this valve, but could not explain its misposition. The lube oil level was restored and root valves 2(3)-138-2(3)018 are to be added to the CO The inspector had no further concerns with the first three item The inspector had discussions with licensee senior management, an assistant operations engineer and the system engineer concerning the fourth item. Test valve (HV-2-13C-52B) is on the discharge line of the RCIC barometric condenser vacuum pump to the torus. Therefore, primary containment was provided by an upstream stop check valve (HV-2-13C-10) and the pressure gauge itself. The as-found configuration was apparently not in accordance with the FSAR since the downstream check valve's function could be considered defeate The pressure gauge is attached to support the performance of ST 6.22, "RCIC Pump Valve, Flow & Cooler". Apparently, the last time ST 6.11 was performed, the hand valve was not closed after reading the pressure gauge. The inspector questioned the usefulness of the gauge since it was only recording pressure supplied against the torus free space (due to attached siphon break). However, if the gauge was to be made permanent, the COL should be revised and a step should be added to ST 6.11 to ensure that the valve is closed following the test. The licensee stated that they were currently planning to remove the gauge on Unit 2 and would
, not add one to Unit 3. The inspector had no further questions nor additional concerns at this tim The inspector's remaining deficiency list was discussed with an assistant operations engineer for appropriate actions. The inspector will continue to follow licensee actions to correct the deficiencie Also, during the inspector's tour in the Unit 2 RCIC room, the inspector observed a 50% fluid level in the reservoir of hydraulic snubber 13-HB-S-1. The inspector
. questioned whether the level was normal, or if the snubber
'
was leakin The inspector reviewed ST 9.15-2A, " Accessible Only Seismic Hydraulic and Mechanical Snubber Inspection U/2 Only", Rev. 4, to determine if there was an acceptance criteria for fluid level or if the fluid port need only
-.+1- - , w w , , - - - - - %-ws>_- *. sw --e-my, w w-1 .4- +. we -, cm. -- - . +
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. 23 be covere ST 9.15-2A only requires the fluid port to be covered; the fluid level percent is recorded and is compared to past ST performances to detect leaks. The inspector reviewed the last two performances of ST 9.15-2A, dated June 28, 1985 and December 3, 1985, and compared those fluid levels to the observed level. Both STs had recorded levels of 45%, so apparently there is no leakag The inspector questioned the system engineer concerning acceptable fluid levels, such as how low the level would get before it is refilled. The system engineer stated that the ST is currently in revision to add a step to refill the reservoir when the fluid level is 50% or les The inspector had no further concerns or question No violations were identifie .6 Licensed Operator Requalification Exams The 1986 Licensed Operator Requalification Exams were given during the period January 9 to February 13, 1987. The exams were graded and the results were announced on February 20, 1987. There were six failures of the 44 exams administered. Three senior licensed operators and three licensed operators failEJ to achieve either an 80% overall or a 70% in each section. As required by procedure A-50, " Training Procedure", the six individuals were given oral exams. All of the individuals received passing grades; thus, they were allowed to continue on shift. These individuals were also placed in an accelerated retraining program as required by procedure A-50. Re-examinations were administered during the period February 26 to March 20, 1987. All six individuals received passing grades on the re-exa The inspector reviewed procedure A-50, exam and re-exam results, and discussed this item with licensee training personnel and opera-tions personnel. No violations were note . IE Information Notice Followup 5.1 IE Information Notice 86-106 In response to IE Information Notice No.86-106, Feedwater Line Break, the licensee has contracted for an independent evaluation of the feedwater system during the 1987 Unit 2 refueling outag PECo has had a program at Peach Bottom to inspect for pipe wall thinning. The program presently includes 16 areas where erosion /
corrosion was judged to be most likely. The main steam cross-around piping between the HP turbine exhaust and the moisture separator inlet has been monitored for the past ten years. A second area, the
s
. 24 RFP recirculation (min flow) piping between A0-2139/3939 A,B,C to the condenser inlet, has been monitored since 1981. The other 14 areas were identified in 198 The 16 " suspect" areas were based upon engineering review of past failure history, materials, and flow conditions. Additional areas will be added if conditions dictate. The 16 pipe inspection areas are defined as follows:
Area # Description 1 Main steam cross-around piping between the HP turbine exhaust and moisture separator inle RFP recirculation (min.-flow) piping from A0-2139/3139 A, B, C to the condenser inle Feedwater long path recirculation piping between R0-2663/3663 and the condenser inle RWCU piping between R0-106 and the condenser inle Main steam drains - main steam lead drains, MSV above seat drains, HP turbine inlet lead drain HPCI steam drains - HPCI turbine steam supply line drain from steam trap (ST-3) to condenser, HPCI turbine SV above seat drair from R0-70A to drain po RCIC steam drains - RCIC turbine steam supply line drain from steam trap (ST-122) to condenser, RCIC turbine SV above seat drain from R0-76 to barometric condense RFPT steam supply piping drains - steam lead, HP MSV above seat, HP MSV below seat, LP MSV above seat, LP MSV below sea RFPT first stage and shell drains downstream of A0-2557/3557 and A0-2685/368 RFP seal leakoff loop seal riser condenser side to A condenser shel Offgas recombiner preheater steam supply drai Extraction steam lines to condense Feedwater heater vent lines to condenser downstream of R0-2059/3059, 2062/3062, 2065/3065, 2068/3068, and 2071/307 _-.
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- 25 14 Main turbine thirteenth stage shell drai Any carbon steel relief valve discharge piping where leakage through the relief valve is suspecte RFP suction side relief valve discharge piping down-stream of RV-2141A, B, Mechanical Engineering Division procedure entitled " Peach Bottom Atomic Power Station Units 2 and 3 Inspection of Piping Subject to Erosion and Cavitation Damage", dated November 1986, provides the schedule for inspections and identifies sample location. For example, Area 3 is to be done during the next refueling outage after March 1987 for Unit 2 and the upcoming refueling outage for Unit 3. For Area 3 there are 57 sample locations identified for the run of pipe at elbows and fittings for Unit Some areas are to be inspected each cycl Based upon previous inspections, pipe in the following areas have been replaced:
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Unit 2 - Areas 7, 8, 16 (MOD 1695 for pipe replacement and re-routing of pipe in Area 2 is scheduled for the 1987 refueling outage).
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Unit 3 - Areas 6, No unacceptable conditions were note .2 IE Information Notice 87-08 On February 4, 1987, IE Information Notice 87-08, " Degraded Motor Leads in Limitorque .DC Motor Operators", was issued to all nuclear power reactor facilities holding an operating license or construction permit. The purpose of the notice was to alert recipients of poten-tially defective DC motors installed in Limitorque motor operator Peerless-Winsmith motors manufactured between December 1984 and December 1985 may be fitted with Nomex-Kapton insulated leads that are susceptible to insulation degradation and subsequent short circuit failur After performing a review of maintenance records, the licensee found six suspect Limitorque motors that were in service in Units 2 and 3, and three more in the equipment storage warehouse. The six Limitorque motors were attached to the following valves:
1 , .
,e 26
' Unit 2
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M0-2-13-16, RCIC Steam Supply Outboard Isolation Valve
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M0-2-23-19, HPCI Pump Discharge Valve
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-M0-2-23-20, HPCI Pump. Discharge. Valve Unit 3
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MO-3-23-19, HPCI Pump Discharge Valve
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M0-3-23-20, HPCI Pump Discharge Valve
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M0-3-5245 (5244A), HPCI Turbine Exhaust. Vacuum Breaker Valv On February 23, 1987, the Electrical Engineering Division at PECo headquarters issued a Justification for Continued Operation (JCO)
for both units. The JC0 concluded that both units should continue operating until their next respective scheduled refueling outage:
Unit 2 is currently in an outage, and Unit 3 is scheduled for September 1987. Engineering justification included: (1) all the valves passed post-maintenance testing, (2) monthly surveillance testing proves their' operability, (3) there are no forces exerted on the motor leads that could lead to a. failure, and (4) the operating period remaining for each unit is relatively shor On February 25, 1987, the inspector attended the PORC meeting to discuss the JCO. The PORC concurred with the JC0 and proposed the following additional requirements: (1) change the monthly surveil-lance test period on M0-2-13-16 to weekly testing due to its limited '
service history (installed January 1987), and (2) determine if.this problem is reportable from an EQ standpoin On February 26, 1987, the inspector attended a conference call between NRC: Region I and the licensee. The NRC determined that the JC0 did not address several important topics. They were: (1)
relate the three spare motors found in the storage warehouse to tbm e insta11ea in the plant, (2) reemphasize the short amount of t.e. remaining for Unit 2 operation until shutdown, (3) discuss ECCS backup to HPCI if it became inoperable, (4) evaluate the hazards associated with removing safety related equipment from service, and (5) discuss how the results of the Unit 2 motor inspection will affect actions to be taken on Unit The PORC reconvened on February 27, 1987, to expand the JC0 and cover the topics requested by NRC: Region I. The inspector obtained the PORC meeting minutes and determined that the additional areas had been adequately addresse .. .. ~.
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On March 5,1987, Unit 3 shut down to repair a hydrogen leak on the main generator. During the outage, the licensee decided to inspect the three suspect valves on the HPCI system to determine what type of insulation is present on the motor leads. After removing the limit switch housing cover, the licensee determined that all three Limitorque motors were covered with EQ insulation (Nomex plus an epoxy impregnated braided fiberglass sleeve).
Therefore, no further action is required for these motor lead Unit 2 shut down for refueling on March 13, 1987. During the outage, the licensee will inspect the remaining three valves on the HPCI and RCIC systems to determine what type of insulation is used on the motor leads. The inspector will continue to follow this are No violations were note . Radwaste Dewatering Facility
In an effort'to minimize radwaste volume, the licensee is pursuing modifications (MOD) to the existing resin dewatering system. MOD 1750A provides for interfacing piping and an associated enclosure to support operation of a resin dewatering system behind the Unit 2 reactor building. The dewatering process involves pumping a resin slurry into a dewatering container and using an air-driven vacuum pump to extract free liqui The system is designed by NUPAC/ Gilbert Commonwealth and installation is in progres In addition, a GE designed AZTECH system has been procured for Peach Botto Installation is scheduled for September 1987. The AZTECH system uses vinyl toluene to dewater and solidify resin. GE will operate the transportable AZTECH syste The AZTECH system will only be used for condensate demineralizer resin The inspector attended a meeting on-site on February 20, 1987, to discuss the resin dewatering facilities and associated modification In addi-tion, the Hydro Nuclear dry active waste, GE super compactor and Radwaste storage facility were discussed. Attachment IB is a listing of attendee The inspector reviewed the MOD 1750A safety evaluation and toured the resin dewatering facility under construction. The licensee's radwaste organization has been changed. A new corporate Director, Radwate Manage-ment and new on-site Senior Engineer, Radwaste have been created. These organizations are responsible for the shipping, processing, and storage of radwaste, and for station housekeeping including decontaminatio Within the scope of the review of the radwaste dewatering facility, no unacceptable conditions were noted. The inspector will continue to follow these modifications, including startup testin No violations were note _ _ _ - ,_ _ - - - -. -- - - - - - - - - . .. -- ..
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. 28 7. Surveillance Testing The inspector observed surveillance tests to verify that testing had been properly scheduled, approved by shift supervision, control room operators were knowledgeable regarding testing in progress, approved !
procedures were being used, redundant systems or components wer available for service as required, test instrumentation was calibrated, work was performed by qualified personnel, and test acceptance criteria were met. Parts of the following tests were observed:
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ST 9.2, Control Rod Exercise, Rev. 15, performed on Unit 2 on 2/9/8 ST 2.3.6A, Functional Test of TE-TS 4933A, Rev. 3, performed on Unit 2 on 2/11/8 In addition, a review of completed surveillance (Attachment 4) tests was performe No inadequacies were identifie . Maintenance For the following maintenance activities the inspector spot-checked administrative controls, reviewed documentation, and observed portions of the actual maintenance:
Maintenance Procedure /
l Document Equipment Date Observed MRF 2-13-M87-1394 Unit 2 RCIC M0-16 Repack 2/15-17/87 MRF 2-13-M87-1421 Unit 2 RCIC M0-16 Motor 2/16-17/87 Replacement
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Administrative controls checked included maintenance request forms (MRFs),
blocking permits, fire watches and ignition source controls, item handling reports, QC involvement, plant conditions, TS LCOs, equipment turnover information, and post maintenance testing. Documents reviewed included maintenance procedures, material certifications RWPs, MRFs, and receipt inspection The above corrective maintenance was initiated due to a Unit 2 RCIC M0-16 packing leak. (The MO-16 valve is the outboard steam supply valve.) The valve.was repacked and attempts to stroke the valve were unsucce ssful . The motor was drawing high current during travel in the closed direction resulting in torque switch actuation. The licensee replaced the motor (Reliance, .33 hp, 250 VDC, Limitorque SMB-000) with a i motor manufactured by Peerless-Winsmith (see detail 5.2).
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. 29 The licensee's failure analysis preliminary report determined that the most probable cause was grounding of the A-2 motor lead (armature brushes)
caused by insulation breakdown and/or moisture in the Limitorque operato Inspections observed that the inner motor bell had rust deposits, the
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motor lead insulation was degraded and water (one pint) came out of the motor clutch housin The inspector examined the failed Limitorque motor, reviewed the licen-see's failure report, and discussed the failure mechanism with licensee engineers. Licensee corrective actions included: IE Notice 87-08 follow-up and justification for continued operation (see detail 5.2); and, an equipment history search for any previous failures. In addition, the licensee is evaluating the failed motor leads insulation and degradation; making recommendations to prevent future failures; determining if this failure is 10 CFR Part 21 reportable; and determining if the packing leak could affect the EQ of the valve. In conjunction with the open item in detail 5.2, these above items are considered unresolved (277/87-07-01).
No violations were identifie . Radiological Controls Program Review An in-depth review of the licensee's Radiological Controls Program was performed during the current inspection to evaluate program implementation in support of routine work activities and outage preparations. Areas inspected during this review included:
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Organization, staffing and management controls,
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Audits of radiological activities,
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Posting, labeling, and contamination control,
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External exposure controls,
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Internal exposure controls,
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ALARA practices,
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Instrumentation, and
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Health Physics responses to fire and shutdow In general, the licensee's program shows improvement in that problems with the organizational structure and existing procedures have been recognized and are being addressed. These changes are early in develop-ment, however, and in the field performance remains poor. One apparent violation, involving High Radiation Area posting was identifie ,
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Programmatic weaknesses relating to scope of Radiation Work Permits (RWPs), performance of surveys, and use of digital alarming dosimeters were also identifie __
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..- 30 9.1 Organization, Staffing, arid Management Controls Improvement of the Radiological Controls Program has been identified as one of the specific actior items included in the Peach Bottom Enhancement Progra Specific tasks identified include a review of organizational effectiveness, development and staffing of a revised health physics (HP) organization, and review and upgrade of health physics procedure . Staffing The licensee currently maintains a very large HP tech-nician staff. The permanent house staff includes 58 HP technicians, 33 of which meet ANSI qualification criteri Approximately 65 additional vendor personnel, two-thirds of which are ANSI qualified technicians, are routinely stationed on-sit Despite the large technician staff, the licensee's orga-nization features only one first-line supervisor (Senior Technical Assistant). Consequently, in-field oversight of the technician staff by the Senior Technical Assistant and his immediate supervisor, the Applied HP Supervisor, is extremely limited and supervisory time and effort is largely directed towards handling administrative concerns (overtime, scheduling,etc).
The licensee's review of organization and staffing effec-tiveness has recognized this weakness and steps are being taken to address it. Seven " Plant HP" positions have been created; these individuals would act as first-line super-visors and would be responsible for in-field oversight of radiological activities. The positions are vacant pending licensee candidate review and apparently will not be filled until after the Unit 2 outag Staffing increases being implemented are the addition of an Assistant Applied HP, to provide administrative assistance to the Applied HP Supervisor, and the addition of one indi-vidual to take over responsibility for the Respiratory Protection program. Both these positions will be filled with vendor personne During the interim, the licensee has provided one contract first-line supervisor per shift
!
to address problems or concerns raised by the contract technicians, 9. Procedures The licensee's radiation protection procedures are under-going a major rewrite by vendor staff as part of the Peach Bottom Enhancement Program. The goals of this rewrite are
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.~ 31 to modernize procedures and to achieve commonality with Limerick Station procedures. The licensee's current estimated completion date for procedure upgrade is July 31, 198 . Communications NRC Combined Inspection 277/86-18; 278/86-19 identified conflicts in the routine interface between maintenance workers and HP technician These conflicts were characterized by the verbal abuse of technicians by the maintenance personnel. In one instance these conflicts resulted in a physical altercation in a contaminated, high radiation are Upper management attention has been directed towards resolving problems with the interface between HP and maintenance. Senior management personnel from the maintenance group briefed on-site maintenance personnel regarding problems that had occurred and the appropriate conduct of HP-maintenance relationships. HP technicians received instruction in methods for insuring effective communications. The inspector interviewed numerous station personnel and determined that positive changes in the HP-maintenance interface have occurre The licensee has also taken measures to improve communi-cations within the HP organization between the technicians and the upper level supervision. Weekly meetings, are held for both the house and vendor technicians; recent events,
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potential problem areas, etc., are discussed at these meet-ings. Additionally, weekly Communications and Problem Solving (CAPS) meetings are held and are attended by the Superintendent, Plant Services, the Applied Health Physicist, and representatives from the house technician staff. The inspector noted, however, that the large con-tractor HP technician group is not represented in the CAPS meetings. The inspector attended a CAPS meeting on March
5, 1987, and noted they represented a unique forum for the technicians to bring problems to a high level of management attentio .1.4 Health Physics Deficiency Reports (HPDRs)
Licensee procedure HP0-C0/600, " Health Physics Deficiency Report" describes the licensee's system for identifying, developing corrective measures for, and tracking radio-logical deficiencies. This system, devoted solely to radiological concerns, was developed when deficiencies with the implementation of the previous corrective action system were identifie In the current system, after a radiological concern is identified, the situation is
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investigated by an HP technician who acts as an "HPDR Investigator". The concern is also ranked from one to four as to severity level (one being most severe). The results of their investigation are passed on to the HPDR Coordinator, who recommends corrective actions and sends the HPDR to the responsible supervisor for dispositio Once dispositioned, the responsible supervisor or Senior Engineer has two days to act on the HPDR recommendation The inspector reviewed the implementation of the HPDR system by the following methods:
--
discussion with the HPDR Coordinator and several HPDR investigators,
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review of selected HPDR reports,
--
review of the HPDR status sheet The above review indicated that the new HPDR system is generally being effectively implemented. The popular belief among interviewed personnel is that this system represents an improvement over the old system; once identified, concerns cannot " drop out" of the system and information concerning the resolution of HPDRs can be easily accessed. The inspector identified two concerns with the implementation of the system:
--
Timely resolution of HPDRs sent to responsible Senior Engineers or work groups is not always apparent. In particular, the inspector noted that response from the Operations Group to dispositioned HPDRs is consistently delinquent. Review of the delinquent HPDRs indicated that several involved significant radiological concerns (i.e., failure to follow HP procedures, etc). Inter-views with HP technicians indicates there is a percep-tion that it is futile to generate an HPDR against the Operations group since it will be ignore The prioritization for review of HPDRs under investi-gation does not appear to reflect the significance of the HPDRs. Due to apparent time and personnel con-straints, the review and disposition process for less severe HPDRs is completed preferentially. The review of more complex or severe HPDRs is delayed since these HPDRs require more time for review, analysis, and determination of recommended corrective actions by the Health Physics group. For example, deficiencies concerning skin cor,tamination or lost dosimetry are sorted out and processed first; these can be processed
-.
33
relatively quickly and enable workers temporarily barred from radiological areas due to the investiga-tion to be returned to work. HPDRs dealing with multiple procedural violations or loss of controls are often set aside until appropriate time and resources can be assigned to the The inspector discussed the above concerns with licensea managemen Responsibility for the implementation of the HPDR program is currently assigned to an individual in the Support HP group. This individual also has responsibility for the Respiratory Protection program. The licensee indicated an additional professional is being brought in to take over Respiratory Protection responsibilities; allowing the i:urrent individual to devote his time to the HPDR syste The licensee also indicated that the concern with delin-quent HPDR response from the Operations group has been discussed between the Superintendent, Plant Services and the Operations Engineer. The Operations Engineer indicated that the delinquency in final disposition of an HPDR is a paperwork problem; corrective actions for identified deficiencies are taken immediately, but completion of the-paperwork to close out the HPDR is not timely. The licen-see indicated additional effort would be expended to insure timely closeout of HPDRs. Adequacy of licensee follow-up of HPDRs is unresolved (277/87-07-02, 278/87-07-01).
9.2 Audits of Radiological Activities Reviews of Peach Bottom radiation protection program implementation and performance are reviewed by three external groups: Quality Control (QC), Quality Assurance (QA), and the Corporate Radiation Protection staff. The inspector evaluated the effectiveness of the current auctit program by holding discussions with QC, QA, and cor-porate-personnel; and by review of the following documentation:
--
QC Detailed Monitoring Checklists (DMCs) associated with Health Physics,
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Corporate review dated February 6, 1986, entitled "Results of PBAPS Respiratory Protection Assessment",
--
Applicable portions of QAD-5, Rev. 13, " Procedure for Performance of QA Division Audits", i
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Applicable portions of QADP-6, " Quality Assurance Division Audit Program",
O
. 34
--
QA audits associated with Health Physics operations: AP 84-33 HPC, AP 84-48 HPC, AP 85-11 HPC, AP 85-14 HPC, AP 86-14 HPC, AP 86-63 HP Within the scope of the above review, no violations were note Although the audit program appears to be adequately implemented, specific areas for improvement in performance were identified and are described belo As noted above, six QA audits of the station Health Physics (HP)
program were reviewed. The inspector noted the current QA auditor with responsibility for the HP area has adequate experience for reviewing HP field operations. However, additional expertise is needed for in-depth reviews of more technical area Licensee QA procedures require that the HP and Chemistry program be audited to the 18 criteria of 10 CFR 50 Appendix B every two years. No formal system or procedure exists, however, which:
(1) identifies the significant program elements in the HP area, and (2) requires that each element be audited over some interva The inspector reviewed all QA HP audits performed during 1984-1985 (most recent completed two year cycle) and noted that the majority of HP program elements were in fact reviewed during this perio This scope, however,' appears to be at the discretion of and due to the efforts of the auditor, and not to the design of the progra The licensee indicated an evaluation would be performed to determine if an HP program element audit scope could be developed and effec-tively implemented. Results of this evaluation will be reviewed during subsequent inspection The inspector also noted a licensee weakness in response to internal audit identified concerns. QA audits performed over one year ago in the respiratory protection and ALARA areas (see section 9.10 and 9.12, respectively) identified concerns which had not been corrected at the completion of this inspectio Previous NRC review in the ongoing audit program of Station Radi-ation Protection activities identified the need for more formal involvement by the corporate Radiation Protection Section. The inspector reviewed a corporate review of the PBAPS respiratory protection program, dated February 1986. Corcerns identified in the review were valid, however findings were not clearly stated, identified with a number, or tracked. A forral response to con-cerns was not required. Consequently, overa'.1 value of the review suffere The Director of the corporate Radiation Protection (RP) Section indicated the following actions were to be implemented as part of the corporate review of station RP activities:
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O
. 35
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Corporate reviews will be performed on a quarterly basi Assistance will be obtained from the QA group for formalizing the audit syste l
--
The corporate staff will provide an individual to assist in QA reviews of the RP program. The Director also indicated the corporate staff will provide input to the QC group identifying indicators to review and assess RP program activitie The licensee indicated a procedure describing the corporate review program was still in development. The inspector did review a memo from the Director, Corporate Radiation Protection Section to the i Peach Bottom and Limerick Station Managers, dated March 5,1987.
l This memo indicates corporate assessments will be performed quarterly and identifies potential scope of the audits. The response to deficiencies identified in QA audits is unresolved (277/87-07-03,278/87-07-02).
9.3 Posting, Labeling, and Contamination Control The licensee's' program for the posting and labeling of radiological areas and radioactive materials was reviewed against the following criteria:
--
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HP0/CO-11, " Establishing and Posting Radiologically Controlled Areas" Conformance with the above criteria was reviewed by the following methods:
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tour of plant radiological areas,
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performance of independent survey measurements,
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discussion with licensee personne Within the scope of the above review, one apparent violation was noted, relating to failure to post a high radiation are CFR 20 defines a High Radiation Area (HRA) as any area where an accessible portion of the whole body could receive in excess of 100 millirem in one hour. 10 CFR 20 requires that each HRA be posted with a sign featuring the radiation caution symbol and the words " Caution" or " Danger" - High Radiation Area".
C
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On February 11, 1987, while performing a tour of the 135 foot ele-vation of the Radwaste building, the inspector noted that the upper platform area of the drum capping station was not posted as a HR Subsequent survey measurements identified general area dose rates inside the area of 150-220 mrem /h The inspector noted the licensee's routine method for posting the ,
, ,
' upper platform consisted of a barrier rope, with attached sign, I stretched between and fastened to the ladder handrails. A barrier I rope and attached sign was noted in the above instance; however, it had been untastened at one end, was trailing on the floor, and was
< not conspicuous. The inspector also noted the attached sign was inadequate, in that it did not specify "High Radiation Area."
Failure to post the high radiation area noted above is an apparent violation of 10 CFR 20.203 (277/87-07-04; 278/87-07-03). Upon iden-tification, the above area was immediately posted by the license )
An additional example of inadequate posting of a high radiation area was noted during this inspection. On February 24, 1987, the inspec-tor noted the access to the equipment pit on the Unit 3 fuel floor was not conspicuously posted as a high radiation area. A rope barrier, with attached "High Radiation Area" posting was noted at the ladder access; however, it had been unfastened at one end, was dangling down into the equipment pit, and was not visible. The licensee indicated the equipment pit is routinely kept posted as a high radiation area; a survey performed on February 24, 1987, indi-cated the presence of a filter inside the pit creating dose rates of 300 mrem /hr at 18" directly above the filter. Upon identification, the licensee immediately reattached the barrier rope, making the l posting conspicuous.
i The inspector noted a human-factors type problem which may have j contributed to the above posting deficiencies. At each ladder l access, a rope barrier with attached sign was routinely stretched between the ladder handrails to post the access. In each situation noted above, the rope had been unfastened at one end, apparently to allow personnel access, and then not reattached. The inspector noted the correct posting was present at the equipment pit access; it had simply not been replaced so that it was visible to personnel. The posting attached to the barrier rope at the drum capping station was inadequate however, in that it did not indicate "High Radiation Area".
The inspector also noted, during tours of various radiological areas, that the placement of boundaries for radiation areas was not always conservative. The licensee routinely posts radiation areas at mrem /hr. On several occasions, placement of radiation area bound-aries was adequate when measuring the field at waist level; however, dose rates of 4-6 mrem /hr were measured at the head level. The
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- licensee indicated technicians would'be instructed to be sensitive to posting' problems created when the dose is originating from. overhea Several strengths were noted in the' licensee's posting and contamination control programs. These included:
,
--
Use of tape on floor, as well as barrier ropes, to identify contaminated areas;
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good control and anticipation of posting requirements as new areas became accessible during scaffold construction on the
..
> . 165' elevation of the turbine building, Me,
'
,D ' extensive and effective use of contamination monitors, S
--
use of pictures outside HRA cubicles in the Reactor Building
- s-showing label plates of pumps located within,
--
use of clear, legible status boards posted on most elevations in the Reactor Building, showing current radiological
-
informatio In addition, the inspector noted ongoing improvement and frequent
.,
'
, self-correction of radiological postings during the inspection period. This specific improvement indicates a technician sensitivity v to potential posting concern .4 External Exposure Controls The licensee's program for minimizing and controlling exposure during routine work activities was reviewed against the following criteria:
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Technical Specification 6.13, "High Radiation Area",
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procedure HP0/CO-4, " Radiation Work Permits",
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procedure HP0/CO-68, " Field Use of Alarming Digital Dosimeters".
The licensee's performance relative to the above criteria was determined by:
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interview of HP technician and supervisory staff,
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observation of ongoing radiological work activities,
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review of numerous Radiation Work Permits (RWPs) and associated surveys and sign-in sheets,
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review of various HP control point logbooks,
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_
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- -.
f 0. .
4 38 s
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review of whole body exposure summary report, dated 12/31/86,
--
review of cumulative exposure history documentation for selected individuals,
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review of Health Physics Deficiency Reports.(HPDRs) 87-0130,87-013 Within the scope of the above review no violations were identifie However, one potential violation has been classified as licensee-identified and is being tracked as an unresolved item pending review of the licensee's corrective action (see section 9.4.1). Additionally, several weaknesses were identified relating to the scope of Radiation Work Permits (RWPs), use of digital alarming dosimeters, and per-
.formance of survey . Drum Capping Aisle (DCA) Incident On February 12, 1987, two Bechtel workers and two HP tech-nicians entered the Drum Capping Aisle (DCA) on the 135'
elevation of the radwaste building to overturn and flush resins contained in three High Integrity Containers (HICs)
down the floor drain. The DCA is a posted high radiation area, contaminated area, and airborne activity are Entry to the area was made on RWP #2-20-0146, titled
"Decon room, grease bearings, lube fittings and rollers, repair limit switch and pneumatic cylinder". This RWP had been suspended prior to the time of entry. The technicians were entering the DCA ahead of the workers, on an " escort survey" basis to reactivate the RW One of the HP technicians was a fully qualified techni-cian assigned to cover the work; the other was a junior technician along to gain familiarity with the are Responsibility for coverina the work had been turned over to the qualified technician shortly before the entr The two Bechtel workers and the junior HP technician all had requested and received exposure extensions up to 600 millirem. The qualified technician, however, had not received an extension and was limited to the normal station limit of 300 millirem / day. The ALARA group estimated that the decontamination work could be completed with a total exposure of 300 millire Upon entry to the DCA the technicians identified general area dose rates up to 12 R/hr near the HIC No effort was made to stop the operation despite the excessive dose rate After progressing past the HICs to survey the entire aisle,
. . _ -
0-f.- 39 the technicians turned and found the workers had knocked over the HICs-and were using water lances to sluice the resin to the floor. drain. Consequently, the technician's exit route was blocked by the overturned HICs and spilled resin. The technicians then retreated to a relatively low dose area (approximately 200 mr/hr).
After a short. period the qualified technician exited the area past the work in progress in an effort to prevent exceeding his administrative exposure limit of 300 mrem per day. This effort was unsuccessful, in that the technician received a total of 340 millirem during the entry. The two workers and junior HP technician exited shortly after the alarming of their digital dosimeters; these individuals'
exposure was within the limits of their exposure extension The licensee generated two HPDRs (87-0130, 87-0131) con-cerning the above event. The major deficiencies identified with the above evolution and current RWP system included the following:
--
An inadequate survey method was used to cover the entry. An inadequate ALARA review was performe Station use of the " escort survey" method to reinstate suspended RWPs appears to be abused. The RWP_used for the work was too broad in scope to control the wor The technician controlling the work was inadequately briefed and did not adequately review paperwork associated with the jo The technician should have stopped the work based on area dose rates (10-12 Rem /hr).
Proposed corrective actions developed by the licensee as-
'
part of the investigation are contained in the HPDR and include the following:
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provide detailed instruction and training to HP tech-nicians on when to stop work, based on changing radiological conditions;
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provide training to ALARA personnel in performing
,
pre-job evaluations;
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emphasize the HP technicians mandate to conservatively follow procedures, require job specific RWPs, and stop the job if required.
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...
L 1e 40 Inspector review of the above incident indicated several potential violations may be associated with the above work,-
including failure to perform an adequate survey, and failure to follow the station RWP procedure. However, in an effort to encourage licensee self-identification and correction of problems, NRC Enforcement Policy (10 CFR 2, Appendix C) allows the mitigation of violations if they
< were identified by the licensee and certain criteria are met. The licensee's identification of deficiencies asso-ciated with the above event will therefore not be cited as a violation at this time but will remain unresolved pending review of the implementation of their proposed corrective actions during a. subsequent inspection (277/87-07-05;
>
278/87-07-04).
H9. Scope of Radiation Work Permits (RWPs)-
As part of their response to the Drum Capping Aisle (see above) incident described above, HP supervision assigned a team of HP technicians to review all currently active RWPs in use at Unit 2 and 3 to identify all deficiencies or pro--
cedural inconsistencies. -The Applied Health Physicist indicated that as a result of this review, deficiencies '
were noted with a significant percentage of active RWP Problems noted included inconsistencies between the RWP and associated ALARA reviews and recent surveys, inclusion of multiple maintenance request forms (MRFs) with one RWP, and -
RWPs in which the scope of work authorized was too broad to control the work, i.e., the RWPs were not " job-specific" enough. This latter weakness with RWP scope was noted to have contributed significantly to problems.noted with the DCA incident. The inspector attended a meeting conducted during this review process in which the Applied Health Physicist emphasized to the technicians the need to insure RWPs are written for specific work evolution The magnitude of the licensee's review in this area reflects positively on the licensee's corrective action program. Additional licensee corrective actions to resolve this weakness will be reviewed in future inspections in association with the unresolved item identified abov . High Radiation Area (HRA) Control Technical Specification (TS) 6.13 and the licensee's RWP procedure allow personnel entry to a HRA if the individual or group is provided with:
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a dose rate instrument, or h
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a digital alarming dosimeter, or
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- 41
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an individual, qualified in radiation protection procedures, who is equipped with a radiation moni-toring instrument and who provides positive contro Implementation of these requirements was reviewed by observation of several jobs performed in HRAs, and the associated RWPs. The inspector noted that all reviewed RWPs appropriately required either a digital alarming dosimeter or ion chamber for entries to a HRA. Implemen-tation of these requirements indicates weaknesses, as noted below:
9.4. On February 9, 1987, the inspector noted an individual working on scaffolding inside the respirator cleaning room on the 116' elevation, radwaste building. The scaffolding was posted as a HRA and the applicable RWP (#2-20-0035A)
required an ion-chamber or alarming dosimete The worker did not have either the required ion chamber or alarming dosimeter. Inspector review identified that no HRA existe Further investigation revealed that the worker had received permission to enter the area without an alarming dosimeter. This was based on the technician's knowledge of an undocumented survey performed earlier that day indicating a hot spot in the area had been flushed. The technician had started to revise the RWP to reflect this but was called away to cover another job. A contributing factor to this problem was the lack of available alarming dosimeters at the instrument cag .4. On February 18, 1987, three workers were noted working inside a posted HRA around the "C" fuel pool heat exchanger (HX), on the 165' elevation of the Unit 3 reactor building. The controlling RWP (3-19-0126) required either an ion chamber or alarming dosimeter. The inspector noted one of the workers was wearing an alarming dosimeter, however this worker was physically separated by approximately 30 feet from the other two who were not wearing alarraing dosimeters. A survey meter was located approximately 50 feet outside the are The inspector noted that no administrative or regulatory exposure limits were exceede The inspector questioned the workers and deter-mined that although they were physically separated, area dose rates in the two areas were approximately the same and an HP technician had
F u .-- 42 r-I' come by to review the work approximately every half hour. The workers indicated, however, that
,
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no direction had been given to them concerning the use of one alarming dosimeter to cover a group; i.e., the need to stay together, et .4.3.3 On February 24, 1987, two workers entered the Unit 3 fuel floor equipment pit to rig in the bridge mast for repair. The controlling RWP (3-94-0138) required an ion chamber or alarming dosimeter. Posting of the area is discussed in section 9.3. The technician controlling the operation indicated that although they were under the technician's direct scrutiny, neither of the workers entering the equipment pit wore alarming dosimeters or carried a dose rate instrumen The inspector also determined that no formal survey of the equipment pit was made prior to the workers' entry to the pit on February 24, 198 Instead, the HP technician relied on his knowledge of conditions from the last survey (performed on February 11, 1987) and a quick check of dose rates with a meter held out over the pit. The workers were instructed to remain in the pit only long enough to land the mast in the pit and, in fact, were in the area approximately three minute The workers were under the direct visual control of the technician during the operation and their movements did not take them into the small high radiation area generated by the filter (see section 9.3). No significant exposure was received by either of the worker .4.3.4 The inspector determined that in each instance above, no violation of TS 6.13 had occurred, in that some combination of the three controls required by the TS was in place during each situation. A weakness was noted, however, in that technicians appear to be individually interpreting requirements for HRA area controls and alarming dosimeters based on situational consideration The inspector reviewed procedure HP0/CO-68 and noted it provides insufficient guidance on the use of the alarming dosimeter (i.e., how many workers can perform an entry with one dosimeter, what are the appropriate restrictions on dose rates or
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uniformity of the radiation field). The inspec-tor interviewed several house and contract technicians and determined,'due to the lack of guidance, technicians were often implementing requirements in use at the last site they had worked at. No clear majority consensus existed as to how many workers could use cne alarming dosimeter or what restrictions should appl The licensee recognized this as a problem and indicated that a station policy would be devel-oped detailing limitations on the use of digital
, alarming dosimeter as an option for controlling HRAs. Developed station' policy, along with development of an implementing procedure and lesson plans, will be reviewed during a subse-quent inspection. Use of digital alarming dosimeters is an unresolved item (277/87-07-06, 278/87-07-05).
9. Radiological Surveys The inconsistencies and deficiencies in performance described above indicate a weakness in the radiological survey program, in that a casual attitude and performance was noted in the performance and documentation of survey Licensee investigation into the DCA incident indicated that
" escort surveys" (in which the technician accompanies the work group and surveys on the way in to an area) are fre-quently used to reactivate suspended RWPs. Use of this survey method puts additional pressures on the technician which may result in inappropriate responses to changing radiological conditions.
"
Inconsistencies described in sections 9.4.3.1, 9.4.3.2,
, 9.4.3.3, and 9.4.3.4, above, highlight additional problems;
'
and reflect technician reliance on undocumented data or
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older surveys to update radiological control .5 Internal Exposure Controls The licensee's program for the sampling and analysis of airborne radioactivity and the provision, maintenance and use of respiratory protection was reviewed against the following criteria:
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10 CFR 20.103, " exposure of individuals to concentrations of radioactive materials in air in restricted areas",
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NUREG 0041, " Manual of Respiratory Protection Against Airborne Radioactive Materials",
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. 44
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HP0/CO-9, Respiratory Protection Program,
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HP0/CO-9a, Respiratory Protection Training & Fitting,
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HP0/C0-9c, Respiratory Protective Equipment Maintenance &
Quality Assurance,
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PECo Radiation Protection Manua Licensee implementation of the above criteria was reviewed by the following methods:
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attendance in a General Respiratory Training (GRT) session conducted on January 30, 1987;
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tour of respirator issue, cleaning, and maintenance facilities,
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review of training records and qualifications for respiratory issue and maintenance personnel,
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observation of routine air sampling practices and review of job-specific air sampling,
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discussion with licensee and contractor personnel,
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review of the following audits and assessments:
AP 86-14 HPC, "PBAPS Dosimetry / Bioassay and Respiratory Protection Programs",
"Results of Peach Bottom Atomic Power Station Respiratory Protection Assessment", dated February 6, 198 . Respiratory Protection Program Within the scope of the above review, no violations were identified. The licensee's program is adequate to insure respirators are appropriately inspected, maintained, and j worn by qualified individual However, certain deficien-cies in program oversight and implementation were noted and are described belo Overall responsibility for the respiratory protection program is held by the PECo Health Physicist - Technical Support, who functions as the Respiratory Protection Administrator (RPA). The RPA or his designee is pro-cedurally required to provide the following levels of program oversight:
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. 45
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review GRT lesson plan and qualify instructors teach-
.ing GRT (HP0/CO-9a),
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approve all program modifications (HP0/CO-9),
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evaluate program effectiveness, on an annual basis, with submission of a report to the Senior Health Physicis Respirator issue, cleaning, maintenance and inspecting is performed by contractor personnel (Bechtel). Bechtel also provides specific training for their respirator issue and maintenance personnel. General respiratory protection training (GRT) is provided to radiation workers by the station training departmen During the course of the above review, the inspector noted the following deficiencies:
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RPA or his designee did not approve the lesson plan for GRT and did not approve or was not aware of several lesson plans developed by Bechtel in October 1986 for-instruction in respirator maintenance, inspection and repai Instead, lesson plans were noted to have been reviewed and approved by the ALARA Health Physicis The RPA did not document reviews of program records, etc., performed to evaluate program effectivenes Written reports summarizing these reviews were not submitted to the Senior Health Physicis The instructor teaching GRT on January 30, 1987, did not reference a lesson plan and consequently omitted several items included in the lesson plan. The instructor indicated the lesson plan was in a format s king it hard to teach fro Several individuals issuing respirators did not appear familiar with all information included on the Respi-rator Qualification List. This list is checked to insure respirators are issued to qualified individual Procedural requirements had not been determined to identify required training for individuals performing respirator maintenance and repai Procedure HP0/C0-9F, " Issuance of Respiratory Pro-tection Equipment", did not require that the issuer check the potential wearer to evaluate facial hair as specified in NUREG 004 a e 46
.
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Individual serialization of respirators had just begun and was not yet proceduralize Exams associated with Bechtel lesson plans for respi-rator maintenance and filter testing were not specific,
~
and the majority of questions dealt with general or
" motherhood" type information. During a review of-completed exams the inspector noted one exam that had been incorrectly graded; the individual passing the exam should have failed. The licensee immediately removed this individual from respirator inspection duties pending re-examina?. io The inspector noted that similar deficiencies to those identified above had been identified by the licensee. A corporate assessment performed in February 1986 identified that PECo oversight of the Bechtel group may be lacking. A QA audit performed in February and March 1986 identified as an " observation" that the RPA was not performing a written evaluation of program effectiveness on an annual basis. No apparent action was taken_by the licensee to review or address these concern The licensee indicated the following corrective actions would be taken:
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The GRT lesson plan will be revised to make it easier to teach from. The RPA will be included as part of the approval cycl The RPA reviewed all lesson plans developed by Bechtel pertaining to respirator issue, maintenance, inspec-tion, and filter testing during the course of this inspection. A memo, dated February 9, 1987, was sent to the Bechtel training coordinator from the RP This memo detailed improvements and needed revisions to be incorporated into the lesson plans. These improvements included making the exams more procedure and task specifi A surveillance test procedure will be developed requir-ing an annual review of records by the RPA to evaluate and report on program effectiveness. A draft copy of this procedure was produced during the inspection perio The lesson plan for respirator issue will be rev" *',
to emphasize familiarity with the respirator quali-fication lis , . . - . . - ..
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e; 47
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The prucedure for respirator. issue will be revised to require the issuer to check the wearer for facial.
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hai A new procedure will be developed identifying training and qualification requirements for individuals per-forming respirator issue, maintenance, and inspectio Procedural revisions will be made t'o formalize the
- licensee's program to serialize respirator The licensee indicated that the QA audit and corporate assessment performed in 1986 did not identify the above mentioned concerns as formal audit findings and conse-quently did not require a response from the HP group. The i inspector noted the QA audit identified the concern as an
" observation"; the-corporate assessment was in an it'armal format and did not characterize findings-at all. The
"
licensee indicated that audit system has been recently revised such that a formal response is now required for
. audit recommendations as well as finding Licensee implementation of the above corrective actions in the respiratory protection program is unresolved
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(277/87-07-07,278/87-07-06). . Air Sampling The licensee's program for sampling and analysis of air-
'
borne radioactivity was reviewed by observation of in-field sampling, tour of the counting facilities, and discussion with HP technicians and supervisory personnel. No apparent violations were identified during this revie The inspec-tor observed that:
' --
air samples were taken as required by the RWP to
support work evaluations,
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air sampler positioning was appropriate for sampling the work area,
.
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all air samplers observed in the field were appropri-
,
ately calibrated.
l The inspector did note the following potential concerns:
,
Discussion with HP technicians and review of logbooks indi-l cates a limited availability of low-volume air sampler Logbook entries indicate jobs have been stopped on several
'
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occasions due to the lack of a low volume air sampler. The
- " low-vol" sampler is the licensee's routine method for i
1-
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e
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,
sampling for iodine and consequently must be used for work in the reactor building The inspector reviewed licensee instrument inventory and determined that an adequate number of low-volume samplers is available (200 on-site, 80 on order). Limited avail-ability appears to be due to delay time in calibration or failure to return samplers to inventory once issued to the field. The licensee intends to permanently mount air samplers at selected locations inside the plant to insure iodine-sampling capability in specific area During tours of the sample counting facilities the inspec-tor noted a large backup of air samples awaiting gamma spectroscopy counting (gamma scan). Air samples sent in from the field are first counted for gross beta activity (10 minute count) and then gamma scanned (10 minute count).
The backup occurs between the beta and gamma counts. On one occasion, the inspector noted approximately 50 air-samples taken on the "X" shift awaiting a gamma scan at the beginning of the "Y" shift. The inspector noted that larger numbers of samples could be expected during an outage situation, aggravating the delay between collection of the air sample and notification of sample result The licensee indicated that all air samples are initially screened by frisker count in the field and the results acted upon by the technicians in the field, prior to delivery of samples to the count room. Appropriate response measures, if required, are therefore taken immediately. The licensee conceded that a large sample backup problem does occur, particularly during outage The following actions are being planned for the upcoming outage:
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opening and use of the Unit 1 counting facility,
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staggering of air sample collection times by area so the count room is not overwhelmed at the end of the shift,
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consideration for reducing count time for known " hot" air samples,
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close scrutiny of air sample results in specific areas to reduce the frequency of unnecessary air samples (i.e., samples showing continuing negative results).
On February 11, 1987, the continuous air monitor (CAM) on the 135' elevation of the radwaste building alarmed due to high counts of the alpha channel. This activity was finally determined to be due to naturally occurring radioactive
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material (radon). Entries in the Senior Technician logbook indicate-that this determination was not performed in a timely manner; it took over five hours and required con-tacting the Applied Health Physicist at home before the HP group decided, based on isotope half-life, that the air-borne activity was from natural source During review of the above event, the inspector determined clear station policy or instruction did not exist concern-ing evaluation of air samples for naturally occurring activity. The licensee indicated clarification and instruction in this area would be provided to the tech-nicians in a Health Physics Instruction (HPI).
Timeliness of the licensee's air sampling program and adequacy of procedural instructions in responding to naturally occurring airborne activity events is unre-solved (277/87-07-08, 278/87-07-07).
9.6 ALARA Program Review The licensee's implementation of an "As Low As Is Reasonably Achievable" (ALARA) Program was reviewed against guidance contained in Regulatory Guide 8.8 and the following criteria:
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HP0/CO-501, " Request for an ALARA Review",
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HP0/CO-502, " Station ALARA Review Committee" (SARC),
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HP-303, " Station ALARA Review Committee", dated March 11, 1987,
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HP0/C0-503, "ALARA Pre-Job Review",
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HP0/C0-504, "ALARA Post-Job Review".
Performance in this area was evaluated by discussion with ALARA personnel and review of the following' documentation:
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exposure estimates / goals for the upcoming Unit 2 outage,
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SARC meeting minutes, 1986-1987,
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selected RWPs and associated ALARA pre and post job reviews,
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QA audit #AP 85-14 HPC, "PBAPS ALARA Program."
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Peach Bottom exposure for 1986 showed a substantial decrease in relation to previous years. The original goal for 1986 was 1600 man-rem; this was later revised to 1000 man-rem. Actual exposure for 1986 totaled approximately 1020 man-rem. Exposure for 1984 and 1985 equaled 2364 and 3354 man-rem, respectivel The above decrease appears to be most directly attributable to an extended operational history for both units in 1986; no significant outage work was performed during this period. The inspector ques-tioned whether 1986 exposure had been compared with exposure accrual during a solely operational period during previous years. Such a comparison would indicate whether the rate of exposure accrual during operational periods had decreased over time. The ALARA Health Physicist indicated this comparison had not been performe Within the scope of the above review, several weaknesses in the implementation of the ALARA program were noted and include the followin Procedure HP0/CO-502, effective during 1986, required that meetings of a quorum of the Station ALARA Review Committee (SARC) be held monthly. The inspector reviewed monthly meeting minutes and determined that only three SARC meetings were held during 1986 which met the full procedural quorum requirements. Meetings held during other months did not have the specified quorum or, in several cases, were not hel QA audit AP 85-14 HPC, performed in 1985, contained a noncom-pliance finding identifying that the SARC was not meeting at the required frequency; however, licensee corrective action appeared inappropriate to prevent recurrenc The ALARA Health Physicist indicated that the organization of the SARC as required by procedure HP0/C0-502 was inadequate in that it required too low an organizational level of partici-pants to be effectiv Recognition of this ineffectiveness led to consequent problems in meeting attendance. Several licensee actions have been taken to correct this concern, however. The SARC was reorganized for 1987 with required attendance from each major group at the Results Engineer level. Procedure HP-303, detailing the new organizational requirements, was approved during the inspection period. The inspector reviewed SARC meeting minutes for January and February 1987, and noted an improvement in attendance and content over meetings held in 198 Documentation of work histories, exposures, etc., by the ALARA group does not appear to be centralized or readily available for future ALARA review. For example, no central file is kept containing all historical post-job reviews; they are maintained by the individual ALARA reviewers in their personnel files and u-___-__-__-__-_____--___-____-______-____-_-_____
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. 51 are consequently not easily available for review. The post-job reviews themselves do not always contain all required informa-tion; exposure goal revisions made during the progress of the job are not included, and consequently the post-job review documentation often is not logical. The inspector determined that the required ALARA post-job review for piping ISI per-formed in the Unit 3 drywell during 1985 was so incomplete as to effectively not have been done; however, this appears to have been an isolated case as all other post-job reviews checked were in plac Procedure HP0/CO-4 requires post-job ALARA reviews for specific jobs based only on the magnitude of the original exposure esti-mate for the job; i.e., as job estimates and exposure categories get higher, post-job reviews become required. The inspector noted no post-job review is required for a low-dose estimated job, even if actual exposure ultimately differs greatly from the
, estimat The licensee indicated a revision would be made to the procedure requiring a post-job review when the original estimate differs from the actual exposure by 25%. This will be reviewed during a subsequent inspectio Outage exposure goals for the upcoming Unit 2 outage were developed based solely on historical information. Although calculated, individual exposure estimates were not totaled for the approximately 2800 Maintenance Request Forms reviewed by the ALARA group. Consequently, it is not known how the historically derived outage goal relates to exposure calculated by a review of estimated man-hours and dose rates. The inspection also noted the licensee's outage exposure goals were derived for broad categories of work, rather than specific work evolutions (with-one exception - Control Rod Drive work). Specifically, the
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licensee's goal of 325 man-rem for the outage was divided up into the following categories: Modifications, CRDs, Support, Outage (routine), Outage (special). Pre-outage exposure goals typically observed in the industry are more job specific.
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An effective method for eliminating identified hotspots (localized areas of high radioactivity) does not appear to be in place. On February 11, 1987, the inspector reviewed an area radiological survey taken on February 5, 1987, on the 135' elevation of the radwaste building. The survey indicated a hotspot (5 R/hr contact,1 R/hr at 18") nn an overhead drain line passing over the new HIC storage are The inspector noted a portion of the line passed over an elevated platform and was consequently accessible to per-sonnel on the platform. Since the survey did not indicate the exact location of the hotspot, the inspector questioned the HP staff as to whether the hotspot and associated 1 R/hr
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- . 52 general area doserate was accessible to personnel. The issue of concern was whether access to the drain line snould be con-trolled as a " lockable" high radiation area, as specified in Technical Specification 6.13. On February 12, 1987, the licensee posted a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> guard in the area to restrict access to the drain line. This guard was maintained for over three weeks while the station and corporate radiation protection staff deliberated as to whether the hotspot in the drain line was accessible to personnel. The inspector took survey measurements in the immediate area of the posted guard; dose rates ranged from one to two mr/hr. A guard in the area indicated typical exposure received at this post was five -
mrem per two hour shift. The inspector expressed concern that -
the licensec would post a guard for a lengthy time rather than change the postin The ALARA Health Physicist indicated that the hotspot was successfully flushed on March 11, 1987. No problems were encountered during the flush. The inspector noted that the delays incurred while final disposition of the easily-flushable hot spot was achieved created unnecessary exposure and reflects poor ALARA review. The ALARA HP indicated work was currently underway on a system to identify and quickly resolve hotspot Additional areas for improvement in the ALARA area included:
more timely allocation of contracts for significant exposure work. The contract for CRD work was not awarded as of one month prior to the Unit 2 outage, thereby allowing less time for mock-up trainin *
specific training in performing ALARA reviews has not been presented to the station ALARA staf *
ALARA considerations are currently not factored into the bidding process when the station contracts out significant exposure wor The licensee indicated that specific training for ALARA coordinators would be performed as part of their corrective actions identified in response to the resin decontamination incident described in section 9.4.1. This training will be reviewed as part of the close-out of that item. The licensee indicated that inclusion of ALARA into the contract / bidding process would be evaluate . 53 Program strengths noted during this review included the following:
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An ALARA Technical Assistant has been detailed to the maintenance group to perform reviews for planned maintenance activities. This should improve communications between the groups and insure ALARA is aware of the actual scope of the wor The ALARA Health Physicist stated on March 13, 1987, that authorization was given for substantial staffing increase These included additional clerical support, the temporary detail of three Senior HP technicians to ALARA, and author-ization for procuring additional professional suppor Implementation of the ALARA program will continue to be reviewed during subsequent inspection .7 Portable HP Instrumentation Adequacy of the licensee's stock of portable HP instrumentation was reviewed by the following methods:
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discussion with the Instrument Control Physicist and the Health Physicist - Support,
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review of licensee instrument inventory logs,
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review of various HP control point logs,
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observation of available instrument inventories in the instrument issue cag The above review indicated the licensee owns satisfactory numbers of instruments to support the RP program; however, significant portions of this stock may be unavailable for in the field use, creating instrument shortage problem Numerous and repeated log entries indicate recurring shortages in digital alarming dosimeters (digidoses) and low volume air samplers. Teletectors and neutron detectors also occasionally appear to be unavailabl Health Physics supervision acknowledged that shortages of certain instruments, specifically low-volume air-samplers and digidoses, do exist. The HP staff has determined " desired" numbers of each instrument type for both operational and outage conditions; review of instrument inventory numbers indicates sufficient instrumentation has been procured to maintain these levels. HP supervision attri-butes the shortages to two causes: (1) a failure to return issued instruments from the field, and (2) slow turn around time of instru-ments from the calibration lab. Instrument repair and calibration is performed by the I&C grou .
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- Licensee planned corrective actions include the following:
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the licensee has ordered an additional 350 digidoses and 80
--low-volume air samplers,
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- the licensee intends to permanently mount low-volume air tsamplers in designated plant area to maintain control over location,
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the HP: staff will continue efforts to communicate instrument needs and priorities with I& .8 Health Physics Response to Fire on March 4,1987, and Unit'
Shutdown on March 5, 1987 On March 4, 1987, an' electrical fire was discovered in Unit 3 on the 195 foot elevation of the turbine building (section 4.2.4).
On March 5, 1987,. Unit 3 was shut down due to an unidentified hydrogen. leakage problem (section-4.2.5). HP response to the-above events was evaluated by the following methods:
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review of Sr. HP Technician log,
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review of radiological surveys taken in response to fire,
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direct observation of actions of the Sr. HP Technician and others in the HP office during response to shutdown on March 5, 198 HP response for the above events appeared decisive and appropriat A sufficient number of surveys was taken in response to the fire to identify that no radiological concerns existed. Technicians were appropriately dispatched to potentially affected areas during the March 5, 1987, shutdown to insure radiological conditions had not changed. Effective sample / survey prioritization and overall direc-tion was note .9 Conclusion Based on extensive infield observations, discussion with station personnel, and review of licensee records, the following conclusions are provide In general, the licensee's program shows some improvement in that problems with the organizational structure and existing procedures have been recognized and are being addressed. These changes are early in development, however, and field performance remains poor.
, One apparent violation, involving High Radiation Area posting, was
! identified. Programmatic weaknesses relating to the scope of Radi-
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ation Work Permits (RWPs), performance of surveys, use of digital
! alarming dosimeters, response to HPDRS, and response to QA concerns l were identified.
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1 Physical Security The inspector routinely monitored security activities for compliance with the accepted Security Plan and associated implementing procedures, including: operations of the CAS and SAS, checks of vehicles on-site to verify proper control, observation of protected area access control and badging procedures on each shift, inspection of physical barriers, checks on control of vital area access and escort procedure On February 11, 1987, the inspector toured the protected area with security supervision. Items checked included: guard tours, detection aids, protected area fencing and lighting. The inspector questioned the security personnel and determined them to be knowledgeabl On March 5 and 11, 1987, the inspector observed SAS and CAS operation, respectively. Items checked included: alarm station manning, alarm response, communications, alarm station operator knowledge, and detection aids. The inspector determined that the CAS and SAS operators were knowledgeable and operations were conducted in a professional manne On March 12, 1987, the inspector observed security badge dissemination; protected area search and entry procedures from the auxiliary SA No inadequacies were identifie . Design Drawing Control Because the licensee had experienced problems with converting drawings to a computer aided drawing (CAD) system at Limerick, the inspector reviewed the drawing status at Peach Bottom. Unlike Limerick, where drawings are under Bechtel control, Peach Bottom's drawings are controlled by the license The licensee issued about 600 electrical drawings in 1984-85 which were drawn by computer. They are in the process of issuing about 180 P& ids drawn by computer and have issued computer drawn P& ids in 1984-85 for a few selected safety systems. A contractor (UE&C) is performing the drawing conversion. The licensee was initially planning to spot check the computer drawings, however after a few drawings were reviewed they decided to make a line by line check of 100% of the drawings. For the P& ids, the licensee described the review effort (which was independent of the contractor's own quality control program) as seven men reviewing drawings for approximately six months. The licensee estimated that fewer than ten request for drawing changes (RDCs) were associated with computer conversion errors when the electrical drawings were issue The inspector discussed the drawings with site and Engineering Design persor.nel responsible for the drawings. Review was made of procedures, and controlled drawings in the station files and control roo In addition, a review was made of RDCs. For the period 1978 through 1984, the number of RDCs were between 30 and 80 per yea .
p o- 56 In 1985 and 1986 the number of RDCs increased to approximately 400 and 300 respectively. The inspector reviewed selected drawing corrections for 1985 and 1986. Most drawing changes were associated with modifica-tions or errors found in the "as-built drawings". Very few of the errors were associated with computer drawn electrical drawings or P& ids. This observation was in agreement with the licensee estimate Based upon this review, the inspector' concluded that the licensee has adequate procedures and took appropriate and effective measures to ensure that drawings converted to the CAD system were error free at Peach Botto The NRB discussed design drawing changes at a meeting on March 12, 1986. The inspector attended a portion of the meeting. The licensee has implemented a formal audit by QA, ISEG, and engineering personne This audit selected seven safety systems for revie The results of the audit will be reviewed in a future inspectio The inspector had no further questions at thi.; time. No violations were note . Unresolved Items Unresolved items are items about which more information is required to ascertain whether they are acceptable violations or deviations. Unre-solved items are discussed in details 8, 9.1, 9.2, 9.3, 9.4.1, 9.4.3.1, 9.5.1 and 9. . Management Meetings
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13.1 Preliminary Inspection Findings A verbal summary of preliminary findings was provided to the Manager, Peach Bottom Station at the conclusion of the inspectio During the inspection, licensee management was periodically notified verbally of the preliminary findings by the resident inspectors. No written inspection material was provided to the licensee during the inspection. No proprietary information is included in this repor .2 Attendance at Management Meetings Conducted by Region Based Inspectors Inspection Reporting Date Subject Report N Inspector 2/16-18/87 Licensing Exams 87-04 Lange 2/2-6/87 Chemistry Van 87-05 Struckmeyer
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- 57 Inspection Reporting Date Subject Report N Inspector 2/2-5/87- Fitness for Duty 87-06 Bailey 2/23-3/6/87 Maintenance 87-08 Chaudhary j 13.3 NRC Region 1/PECo Management Meeting on February 20, 1987
. On February 20, 1987, a management meeting was held at the Peach Bottom Station. At this meeting, PEco discussed the status of the
" Peach Bottom Enhancement Program (PBEP)". The PBEP was developed in response to the June 1986 SALP Report and the Region I Diagnostic Team Inspection Report 277/86-12 and 278/86-13. The PBEP is designed to improve the short and long term safety, reliability, and operating effectiveness at Peach Bottom. Specific goals, objectives, action
' items and tasks of the PBEP were discussed. The licensee presented the current status of the PBEP implementation. In addition, the following items were discussed: operating status, operator training, Appendix R status, chemistry status, outage planning and health physics status. A list of meeting attendees is included in Attach-ment 1A to this inspection repor : The inspector will continue to follow the implementation of the
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PBEP. Another management meeting will be scheduled in six to
, eight weeks.
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ATTACHMENT 1A PECo/NRC Meeting February 20, 1987 NRC T. P. Johnson, Senior Resident Inspector, PBAPS R. J. Urban, Resident Inspector, PBAPS R. M. Gallo, Chief, Projects Section 2A, DRP R. R. Bellamy, Chief, Radiation Protection Branch, DRSS R. J. Clark, NRR/ DBL, Project Manager L. H. Bettenhausen, Chief, Operations Branch, DRS J. H. Williams, Resident Inspector, PBAPS W. F. Kane, Director, Division of Reactor Projects T. E. Murley, Regional Administrator A. A. Weadock, Radiation Specialist PECo R. H. Logue, Assistant to Manager - Nuclear Support B. L. Clark, PBAPS Administrative Engineer W. M. Alden, Nuclear Support, Licensing Section D. L. Oltmans, Senior Chemist, PBAPS A. E. Hilsmeier, Senior Health Physicist, PBAPS J. B. Cotton, Superintendent Plant Services, PBAPS M. Cassada, Director of Radiation Protection J. F. Mitman, PBAPS Maintenance Engineer A. A. Fulvio, Technical Engineer, PBAPS S. R. Roberts, Operations Engineer, PBAPS R. H. Moore, Superintendent, QA Division F. W. Polaski, Outage Planning Engineer, PBAPS F. J. Coyle, PB Mechanical Project Engineer R. S. Fleischmann, Manager, PBAPS G. M. Leitch, Manager, Nuclear Generation Department J. W. Gallagher, Vice President, Nuclear Operations J. Johanson, PBEP Team J. E. Winzenried, PBEP Team D. C. Smith, Superintendent Operations, PBAPS A. J. Weigand, Vice President Electrical Production W. F. Casey, Superintendent Station Maintenance Division R. J. Costagliola, Superintendent Nuclear Maintenance J. K. Davenport, Supervising Engineer Maintenance Commonwealth of Pennsylvania S. Maingi, Principal Nuclear Engineer, Bureau of Radiation Protection (BRP)
W. P. Dornsife, Chief, Division of Nuclear Safety, BRP
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ATTACHMENT 18
.PECo/NRC Radwaste Meeting, February 20. 1987
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T. P. Johnson, Senior Resident Inspector, PBAPS :
R. J. Urban, Resident Inspector, PBAPS A. A. Weadock, Radiatioi Specialist R. J. Clark, NRR/ DBL, Project Manager PECo W. J. Knapp, Electrical Production J. A. McHenamin, Nuclear Services
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J. F. Mitman, PBAPS Maintenance Engineer T. E. Shannon, Mechanical Engineering W. A. Alden, Nuclear Support, Licensing Section S. J. Mannix, Assistant Operations Engineer D. O'Brien, Consultant M. E. Hyslop, Mechanical Engineering C. M. Cooney, Mechanical Engineering Commonwealth of Pennsylvania
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i S. Maingi, Principal Nuclear Engineer, Bureau of Radiation Protection (BRP)
W. P. Dornsife, Chief, Division of Nuclear Safety, BRP
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ATTACHMENT 2 Documents Reviewed for RCIC System ESF Walkdown 1. Peach Bottom Atomic Power Station Units No. 2 & 3, Technical Specifications 2. S.3.5.A, " Normal Set-Up of the RCIC System for Automatic Operation, Rev. 13 3. COL S.3.5.A, " Reactor Core Isolation Cooling System (Unit 2 and 3),
Rev. 1 4. P&ID M-359, "RCIC System", Sheets 1 and 2 5. P&ID M-360, "RCIC Pump Turbine Details", Sheets 1, 2, 3, and 4 ;
6. Peach Bottom Atomic Power Station Units No. 2 and 3, Updated Final Safety Analysis Report
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J ATTACHMENT 3 Surveillance Tests Reviewed for RCIC System ESF Walkdown ST 1.2, "RCIC Logic System Functional Test", Rev. 6 (TS 4.5.D.la, Table 4.2.B.10) ST 2.2.4A/8, 5A/B; ST 2.7.4A/B, 5A/B, " Calibration Check / Functional
, Test of DPIS 2-13-83/84 (Unit 2) and DPIS 3-13-83/84 (Unit 3)", (TS Table 4.2.B.9) ST 2.3.8; ST 2.8.8, " Functional Test of TE/TS 4936, 37, 38, 39 A-D (Unit 2) and TE/TS 5936, 37, 38, 39 A-D (Unit 3)", (TS Table 4.2.B.11) ST 2.3.8, 9, 10, 11 A/B/C/D; ST 2.8.8, 9, 10, 11 A/B/C/D, " Calibration Check of TE/TS 4936, 37, 38, 39 A-D (Unit 2) and TE/TS 5936, 37, 38, 39 A-D (Unit 3)", (TS Table 4.2.B.11) ST 2.1.10 A-A/B, B-A/B, C-A/B, D-A/B; ST 2.6.10 A-A/B, B-A/B, C-A/B, D-A/B, " Calibration Check / Functional Test of PS-2-13-87 A-D (Unit 2)
and PS-3-13-87 A-D (Unit 3)", (TS Table 4.2.8.13) ST 2.4.27, 28; ST 2.9.27, 28, " Calibration Check of LT/LISL 2-13-170, 171 (Unit 2) and LT/LISL 3-13-170, 171 (Unit 3)", (TS Table 4.2.B.18) ST 2.5.26, 27; ST 2.10.26, 21, " Functional Check of ECCS A/C-2, B/D-2 Card File (Unit 2) and A/C-2, B/D-2 Card File (Unit 3)", (TS Table 4.1.B.18) ST 6.11, "RCIC Pump, Valve, Flow & Cooler", Rev. 30 (TS 4.5.D.lb, c, d; 4.5.H; Table 3.7.1; Table 4.2.8) ST 10.2, "RCIC Flow Rate at 150 # Steam Pressure", Rev. 6 (TS 4.5.0.le; 4.11.C2)
1 ST 5.2, " Venting of RCIC Pump Discharge Lines", Rev.1 (TS 4.5.G)
1 ST 11.4, "RCIC Simulated Automatic Actuation Test", Rev. 0 (TS 4.5.0.la; Table 4.2.B)
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J ATTACHMENT 4 Surveillance Tests Reviewed
, ST 1.2, "RCIC Logic System Functional Test", Unit 2 (12/13/86); Unit 3 (10/14/86) ST 5.2, " Venting of RCIC Pump Discharge Lines", Unit 2 (7/11/85) ST 6.11, "RCIC Pump, Valve, Flow & Cooler", Unit 2 (1/4/87); Unit 3 (1/11/87) ST 6.11.8, "RCIC Torus Suction Check Valve Operability", Unit 2 (7/11/85); Unit 3 (3/16/86) ST 10.2, "RCIC Flow Rate at 150 # Steam Pressure", Unit 2 (7/9/85); Unit 3(3/5/85) ST 12.13, "RCIC Vacuum Relief Valve VRV-139A-D Functional", Unit 2
.(12/9/86); Unit 3 (12/11/86)
L ST 12.15.5, "RCIC Pump Contaminated Piping Inspection", Unit 2 (11/20/86); Unit 3 (10/7/86)
l ST 12.15.6, "RCIC Turbine Contaminated Piping Inspection", Unit 2 (12/3/86); Unit 3(10/14/86)
- ST 2.3.8, " Functional Test of TE/TS 4936, 37, 38, 39 A thru D", Unit 2 .
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10. . ST 2.3.80, " Calibration Check of TE/TS 49360", Unit 2 (11/12/86)
11. ST 2.3.10B, " Calibration Check of TE/TS 4938B", Unit 2 (11/10/86)
12. S7 2.8.8, " Functional Test of TE/TS 5936, 37, 38, 39 A thru D", Unit 3 (12/30/86)
13. ST 2.8.8A, " Calibration Check of TE/TS 5936A", Unit 3 (11/13/86)
1 ST 2.8.11A, " Calibration Check of TE/TS 5939A", Unit 3 (10/25/84)
1 ST 2.1.10A-A, " Calibration Check of PS-2-13-87A", Unit 2 (11/4/86)
1 ST 2.4.27, " Calibration Check of LT/LISL 2-13-170", Unit 2 (4/19/85)
1 ST 2.5.26, " Functional Check of ECCS A/C-2 Card File:, Unit 2 (1/5/87)
18. ST 2.9.28, " Calibration Check of LT/LISL 3-13-171", Unit 3 (2/16/86)
19. ST 2.9.27, " Calibration Check of LT/LISL 3-13-170", Unit 3 (2/16/86)
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b.' Attachment 4 2 2 ST 2.1.10C-8, " Functional Test of PS-2-13-87C", Unit 2 (12/29/86)
21. ST 2.6.10B-8, " Functional Test of PS-3-13-87B", Unit 3 (12/30/86)
22. ST 2.6.100-A, " Calibration Check of PS-3-13-87D", Unit 3 (11/7/86)
23. ST 2.2.48, " Functional Test of DPIS 2-13-83", Unit 2 (12/29/86)
24. ST 2.2.5A, " Calibration Check of PS-2-13-84", Unit 2 (11/6/86)
25. ST 2.7.4A, " Calibration Check of PS-3-13-83", Unit 3 (11/11/86)
26. ST 2.7.58, " Functional Test of PS-3-13-84", Unit 3 (12/30/86)
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