IR 05000293/1987016
| ML20214S430 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 05/22/1987 |
| From: | Wiggins J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20214S407 | List: |
| References | |
| 50-293-87-16, NUDOCS 8706090267 | |
| Download: ML20214S430 (24) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
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Docket / Report No. 50-293/87-16 Licensee: Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199 Facility: Pilgrim Nuclear Power Station Location: Plymouth, Massachusetts Dates: February 21, 1987 - April 6, 1987 Inspectors: M. McBride, Senior Resident Inspector J. Lyash, Resident Inspection T. Kim, Resident Inspector L. Doerflein, Project Engineer L. Plisco, Senior Resident Inspector (Susquehanna)
Approved By: ~
44dV 7 J iggins,C$fGReactorProjects Dat6 S ion 18 Areas Inspected: Routine resident inspection of plant operations, radiation protection, physical security, plant events, maintenance, surveillance, outage activities, and reports to the NR Results: One violation was identified concerning failure to properly implement fire protection equipment surveillance procedures (section 3.c). This viola-tion indicates a lack of operator sensitivity to careful performance of sur-veillances and incomplete recognition of fire protection equipment status by the Fire Protection Grou Additional inspector concerns included the following:
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The absence of appropriate drawings and instructions for contractors per-forming MCC testing which may have caused an electrical circuit breaker to be miswired (Section 3.a).
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An apparent shortage of direct involvement in contractor work by the licensee maintenance organization is a potential weakness (Section 3.a).
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Poor mechanical equipment isolation and tagging practices (Section 3.a).
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Programmatic weaknesses in the processing of finger rings and in evalua-ting pocket dosimeter-thermoluminescent dosimeter discrepancies (section 3.b).
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Inadequate preplanning for electrical equipment isolations which led to the failure of bus B6 to reenergize following a loss of offsite power (section 4.b).
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TABLE OF CONTENTS-Page Summary of Facility Activities ........................ 2 Followup on Previous Inspection Findings .............. 2 Violations, Inspector Follow Items Routine Periodic Inspections .......................... 4- Plant Maintenance and Outage Activities- Radiation Protection Fire Protection Security-
' Review of Plant Events ................................ 17 . Unauthorized Refueling Bridge Movement Loss of Offsite Power Review of Licensee Event Reports (LERs) . . . . . . . . . . . . . . . 19 Allegation, Followup ................................... 20 Management Meetings ................................... 22
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Attachment I - Persons Contacted
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DETAILS 1.0 Summary of Facility Activities The plant was shutdown on April 12, 1986 for unscheduled maintenance. 'On July 25,1986, Boston Edison announced that the outage would be extended to include refueling and completion of certain modification Fuel was removed from the reactor in February 1987 to support these modification The reactor remained defueled throughout this inspection period. Major ongoing work included diesel generator, motor control center, valve motor operator and main turbine overhaul Source and intermediate range nuclear instrument dry tubes, local power range monitors, control rod blades, and reactor vessel instrument lines were also replaced during the perio NRC inspection activities during the period included a two week inspection by the NRC Nondestructive Examination Van and a radiation specialist inspection concerning an unexpected worker skin radiation dos NRC Commissioner Lando Zech toured the station on March 10, 198 .0 Followup on Previous Inspection Findings Violations (Closed) Violation (82-0?-04): Failure to use approved procedures to per-form safety-related maintenanc The inspector noted that inspection report 84-02 determined the licensee had established procedures which required all technical information received 'from vendors to be evaluated by the Nuclear Engineering Department for applicability and to be imple-mented via a Plant Design Change Request. As stated in the final response to the violation dated May 14, 1984, the licensee provided training to appropriate station maintenance personnel to re-emphasize the controls and requirements that apply to maintenance activities including the use of approved procedures for performance of maintenance and the subsequent documentation methods. The inspector reviewed procedures 1.5.3, " Main-tenance Requests", and 1.5.7, " Unplanned Maintenance", and noted that these procedures clearly require station maintenance be performed in accordance with approved procedure If an approved procedure does not exist, the Chief Maintenance Engineer will prepare one and have it ap-proved by the Operations Review Committee prior to approving the Main-tenance Pequest. The inspector had no further questions regarding this ite _ _ _ - _ _ _ _ _ _ _ - . . _ _ _ _ _ . _-. _ _ _ - _ _ _ ._
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(Closed) Violation (82-04-03): Failure 'to' include appropriate test, vent
.and drain valves in valve lineup procedures and on Piping and Instrument Diagrams (P&ID). The inspector noted that, as part of the completed pro-cess to update design documents (PUDD program), the licensee performed system walkdowns and updated P&ID's and procedures to reflect the as-built condition. This included the addition of appropriate test, vent and drain valves. The inspector noted that procedure 8.2.3, " Visual and Manual Inspection of Primary Containment Isolation Valves 1" and Smaller" has
'been retired. However, when the inspector compared the operating proced-ure valve lineups, the P&ID's and the Local Leak Rate Test diagrams, he found several inconsistencies with the numbering / identification of the test, vent and drain valve The licensee indicated a program was cur--
rently in progress to walkdown all plant systems to ensure that all valves are properly labeled and identified on P&ID's and system operating proced-ures. The licensee committed to this program in response to findings in NRC inspection 86-06 and is expected to be completed by December,198 The original violation is administrative 1y closed and the licensee's cor-rection of the remaining valve labeling / identification problems will be followed under existing item 86-06-0 (Update) Violation (85-03-11): Failure to use a current revision of test procedure. The inspector reviewed the licensee's corrective actions to preclude recurrenc During a previous inspection period as documented in the inspection report (IR87-03), the inspector witnessed performance of the surveillance test 8M2-2.10.8.6 where the test was per-formed utilizing Revision 5 of the procedure instead of the approved Revision 6. - This item will remain open pending licensee's effective long-term corrective action (Closed) Violation (85-20-02): Failure to maintain the Main Steam Line (MSL) Radiation Monitor trip setpoints less than seven times backgroun In response to this violation, the licensee developed and implemented procedure 7.4.3.1, " Daily Readings of Main Steam Process Radiation Mon-itors" which requires that a technician record daily each MSL radiation monitor reading and the seven times full power background valu This value is then compared to the pre-established MSL radiation monitor trip setpoints, annotated on the data sheet, for acceptability. The procedure also requires the chemical engineer review the results of the daily com-parisons. The inspector concluded that this procedure should prevent recurrence of this event and had no further question Inspector Follow Items (Closed) Inspector Followup Item (85-28-02): Review licensee corrective actions for incident involving water in the reactor building ventilation system. The licensee determined that the "E" condensate demineralizer was inadvertently pressurized to condensate pressure during a resin transfer
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to the cation regeneration tank because the demineralizer outlet valve (MO-2E) did not .ully close. The failure of the valve to fully.close was not detected due to a malfunction of the valve's position indication. The licensee replaced the faulty position indication limit switch and the valve operator torque assembly. In addition, the licensee verified the valve sequencing in the operating procedure was correct and inspected the reactor building contaminated area exhaust ventilation duct for resin intrusion. No resin was found indicating that modifications installed by PDC 82-27 during the 1984 outage to prevent resin intrusion into the reac-tor building ventilation system functioned as designe The inspector also noted that, due to past problems, the licensee routinely inspects the ventilation for resin intrusion in accordance with procedure 3.M.4-70,
" Resin Inspection of H&V Ductwork". The inspector found the licensee cor-rective actions adequate and had no further question (Closed) Inspector Followup Item (87-06-01): Follow up on licensee ac-tions in controlling Special Nuclear Materials (SNM) less than 1 gra During a previous inspection, the inspector's review determined that the SNM Inventory and Transfer Control Procedure 4.0 did not provide for the effective control and accounting for items containing less than 1 gram of SNM (i .e. TIP's, LPRMs, IRMs, and SRMs). The licensee stated that the procedures would be revised, consistent with 10 CFR 70.51 (b)(1)(c), by March .31, 1987 and that the licensee would account for all items. The inspector reviewed supporting records that showed physical inventories of SNMs that are less than 1 gram were conducted on February 13, 1987. The inspector also verified that the station procedure 4.0 SNM Inventory and Transfer Control has been revised to control and account for SNM less than 1 gra .0 Routine Periodic Inspections The inspectors routinely toured the facility to assess general plant and equipment conditions, housekeeping, and adherence to fire protection, security and radiological control measures. Ongoing work activities were monitored to verify that they were being conducted in accordance with approved administrative and technical procedures, and that proper commun-ications with the control room staff had been established. The inspector observed valve, instrument and electrical equipment lineups in the field to ensure that they were consistent with system operability requirements and operating procedure During tours of the control room, the inspectors verified proper staffing, access control and operator attentivenes Adherence to procedures and limiting conditions for operations were evaluate The inspectors exam-ined equipment lineup and operability, instrument traces and status of control room annunciators. Various control room logs and other available licensee documentation were reviewe .
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In addition to routine equipment operability confirmation, the inspectors performed independent walkdowns of selected safety systems. This included confirmation of the as-built system configuration, identification of any degraded conditions and evaluation of procedure adequac The inspector observed and reviewed the licensee's outage, maintenance and problem investigation activities to verify compliance with regulations, procedures, codes and standard Involvement of QA/QC, safety tag use, personnel qualifications, fire protection precautions, retest require-ments, and reportability were assesse The inspector observed tests to verify: test performance was in accord-ance with approved procedures and LCO's; collection of valid test results; removal and restoration of equipment; and oeficiency review and resolu-tio Radiological controls were observed on a routine basis during the report-ing period. Standard industry radiological work practices, conformance to radiological control procedures and 10 CFR Part 20 requirements were observed. Independent surveys of radiological boundaries and random sur-veys of nonradiological points throughout the facility were taken by the inspecto Checks were made to determine whether security conditions met regulatory requirements, the physical security plan, and approved procedures. Those areas checked included security staffing, protected and vital area bar-riers, personnel identification, access control, badging, and compensatory measures when require Plant Maintenance and Outage Activities
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Motor Control Center Testing During a tour of the reactor building on February 23, 1987, the inspector observed a contractor electrical technician working in MCC B-1 An uncontrolled wiring diagram of breaker B-1841 was taped to the inside of the cubicle door. Many cubicles have similar out-of-date drawings mounted on the door The tech-nician was hand-copying the drawing onto a sheet of paper. When questioned, the individual indicated that he had identified a wiring discrepancy and that he would use the drawing for trou-bleshooting and repair. The individual also stated that he had not been provided controlled drawings and did not know how to obtain them. The inspector accompanied the technician to the breaker test area and reviewed the maintenance package and con-dition of breaker B-184 r
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Breaker B-1841 was removed for testing and inspection on February 17, 1987 and reinstalled on February 18. On February 20, a QC inspector reviewing field terminations in the compart-ment, . identified that the overload block was cracked. The cub-icle was removed and the cracked block replaced on February 2 The cubicle remained on the bench unattended until February 2 On that date, the technician attempted to test the replaced overload block and noticed that sever:1 terminations were loose and that the wiring configuration was incorrect. At this point the technician proceeded to motor control center (MCC) B-18 and was observed by the inspecto Internal wiring at the overload heater block, control circuit terminal strip and the motor leads was incorrect. The effect of this miswiring would have been a failure of the valve motor operator to function and connection of the 120 VAC control cir-cuit to 1 pole of the 480 VAC power suppl Review of the maintenance package by the inspector identified several concerns. The initial cubicie inspection was not ade-quately conducted in that the cracked heater block was not iden-tified. The procedures contained only cubicle test information, although replacement of internal components was routinely per-formed. No wiring diagrams were included in the packages. The technician involved was unaware of how to obtain the needed drawings and instead utilized the uncontrolled drawing available on the cubicl'a door. While a lifted lead and jumper log was included in the package, it was applied only to external wirin Internal leads lifted during component replacement and testing were not reqJired to be labeled or listed on the lifted lead lo The inspector concluded that detailed step-by-step procedures for breaker component replacement may not be necessary provided adequate training, drawings and control measures are applie In this instance, the absence of appropriate drawings and re-quirements to label and record lifted leads may have contributed to the miswiring of the cubicle. While post work testing would have identified this problem it is not a substitute for correct performance of maintenance activitie In response to the inspector's concerns, the licensee counseled the contractors performing the testing on the need to use con-trolled drawings. Controlled drawings were provided for use during subsequent activitie The licensee also committed to
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remove the uncontrolled drawings from the cubicle doors and to i apply the lif ted lead log and labeling requirements to internal wirin In addition, measures were established to secure unat-tended equipmen .
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9 The inspector further noted that Quality Control Inspection Report (QCIR) 86-10-98A discusses post work testing of RHR valve 1001-168. This QCIR indicates that two motor leads had 'been swappea at the MCC compartment, o.: the internal side. When the valve open push button was depressed the' valve drove closed and tripped the overload The QCIR indicated that a condition adverse to quality had been identified but also indicated _that no nonconformance or deficiency report had been issued. The Maintenance Department manager responsible for the activity stated that the events described in the QCIR were inaccurate but that he had not discussed the incident with QC, Similarly, QC did not appear to have taken steps to resolve the disagreemen Subsequent investigation by the licensee verified that the breaker wiring was correct and that the QC inspector's observa-tions were a result of miscommunication. The inspector dis-cussed the situation at the Exit Meeting for this inspection, stressing the- need for prompt resolution of QC observations or conflict In addition to MCC work, the emergency diesel generator over-haul, motor operated valve preventive maintenance, 4160 VAC switchgear overhaul and HFA relay preventive maintenance are all performed and supervised by contractors. The Assistant Chief
- Maintenance Engineer (ACME) is the single licensee maintenance individual responsible for oversight of these activitie It appeared to the inspector that more direct licensee involvement in the MCC work may have identified the process weaknesses soone This aspect was also discussed with the license Diesel Generator Maintenance Activities The inspector observed portions of work activities associated with Maintenance Request (MR) 87-61-15,' Preventive Maintenance on B Diesel Generator, performed during February 24-27 and March 17-20, 1987, to determine that the work was conducted in a cord-ance with approved procedures, Regulatory Guides, Technical Specifications, and applicable industry codes and standard The following items were considered during this review: Limi t-ing Conditions for Operations were met while components or sys-tems were removed from service; required administrative approvals were obtained prior to initiating the work; activities were accomplished using approved procedures; QC hold points were established and performed where required; activities were accom-plished by qualified personnel; replacement parts being used ,
were properly certified; and fire prevention controls were properly implemente . _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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The preventive maintenance on the B diesel generator was per-formed in accordance with Procedure No. 3.M.4-36, Diesel Gener-ator Maintenance - General and Preventative. Review of the procedure found it thorough, consistent with the vendor Tech-nical Manual, and appropriate for the skills of those using i The procedure used was adequate to control the activity, and was properly approved. A controlled copy was not maintained at the job site, since a working copy was used, however, the controlled copy was kept by the contractor job supervisor in the mainten-ance office. The inspector reviewed the controlled copy and verified the completed activities had been properly documente No discrepancies were note The maintenance procedure reviewed did not contain any pre-designated QC hold points or QC notification point The MR's QC Control Requirements block was properly filled out and stated
" Notify QC Prior to Start, MRIR's REQ, Review Test Data". No other documentation defined the QC requirements for this complex -
activit The inspector discussed this with the QC supervisor and determined that the QC hold points were communicated ver-bally to the workers during daily activities. The QC_ inspectors went to the job site daily to determine the job status. There-fore, the inspector observed that the current system appears to place the burden on QC to monitor the job rather than the work force to notify QC. In addition, preplanned, detailed inspec-tion points were not clearly documented prior to the start of the work. QC Inspection Procedures were utilized to determine the extent of QC coverage during the MR review process, and the detailed QC hold points were communicated verbally to the job supervisors prior to commencement of the work. Although this practice appeared inefficient, the inspector observed good QC coverage of the work activities. Review of eight QC Inspection Reports associated with the maintenance activities verified that adequate QC inspectior;s were being performed. For example, cleanliness requirements were not clearly documented in the pro-cedure but thorough QC inspections for cleanliness occurred during reassembl The QC inspectors observed and interviewed during the mainten-ance activities were knowledgeable of the procedures and re-quirement Strong licensee management oversight of the contractor work activities was not evident during the inspector's observation of the above-described maintenance activities. The preventive maintenance was performed by a contractor, National Marine Ser-vices, and was supervised by another contractor, Multi-Aep. The
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inspector did not observe any licensee representatives directly monitoring the job activities during the periods of his observa-tions. In addition, three QC Inspection Reports for the main-tenance activity noted that supervision was not present at the job site during QC observation The contractors interviewed and observed, however, were very knowledgeabl The vendor Technical Manual at the job site was marked "For Information Only", but was actively utilized during the mainten-ance by the workers. The controlled copy was maintained in the maintenance office by the job supervisor No discrepancies were noted by the inspector between the manuals, and no defic-iencies resulted from this practic Several significant delays were encountered during the work due to replacement parts problems. For example, the radiator re-placement gaskets were found to be significantly thicker than those replaced. In addition, the replacement radiator cores also had different dimensions. The two radiators for the diesel each contain 3 core sections which are clamped together and mounted at the top and bottom to C-shaped channels. Accumula-tion of the dimensional differences resulted in the radiator assemblies being as much as 1-inch longer than it was with the original cores and gaskets and the bolt holes did not line up in the base plat This was not identified until the radiator installation was attempted. An engineering service request (ESR) was issued to allow el tion of the bolt holes. The inspector reviewed ESR Respo %morandum NED 87-273, dated March 18, 1987, which allowe o- bolt hole elongation, and observed the subsequent instali '.i The ESR also stated that a Potential Condition Adverse to Quality (PCAQ) would be devel-oped to address the fact that detailed drawings of the radiator assembly were not available. No discrepancies were identified during review of the ESR. Other dimensional differences between original and replacement parts were identified by the license For e> tmple, replacement cylinder heaus contained smaller di-ameter chaust manifold bolt holes than those on the previously installed heads. The licensee's Quality Assurance Engineering Section initiated an evaluation of the diesel vendor spare parts program in response to these problem On March 1, 1987, the licensee reported via ENS that nondestruc-tive examination of the B emergency diesel generator fan blades had identified several surface indications. Heat is removed from the diesel engine cooling water system by air cooled radia-tors. Cooling air is drawn across the radiators by a large fa During a maintenance inspection, indications were found at the
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base of two of eight blades on one fan which may be signs of cracking. These blades were replaced. The licensee is evalu-ating the significance of the indications. The licensee also identified that the fan blade hubs are rated for 531 RPM. Actual operating speed of the blades is 600 RP Nonconformance re-ports were initiated to document this discrepancy and track its resolution. No instances of fan blade or hub failure or impact on diesel performance were identified. The licensee has main-tained a procedure for periodic dye penetrant examination of the blades since 1976. This rariodic examination requirement was initiated in response to concerns regarding possible blade fatigue failure. The licensee initiated several nonconformance reports, failure and malfunction reports and an engineering service request to track followup of these problem Additional Observations
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On March 12, 1987, during a tour of the reactor building auxiliary bay, the inspector identified a valve which had been removed from the salt service water (SSW) system with two red tags attached. Valve MOV-3839 was included as part of the isolation boundary under a maintenance request (MR)
for overhaul of the reactor building closed loop cooling water (RBCCW) system heat exchanger. It had also been iso-lated under a separate MR for hydro-testing of system pip-ing. Two red tags were applied as requested by the above two MRs. Individuals working under a third MR physically removed the valve from the piping system without clearing the red tag Other instances of red-tagged components being removed have been noted by the inspector These other cases, however, involved only minor components such as vent and drain valves within the isolation boundar The inspector discussed this incident with plant manage-
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ment, stressing the possible impact on plant and personnel safety which could result from this practice. The Chief Operating Engineer, independent of the inspector, also noted the situation and took prompt action to review and revise the tagging requests and notify the involved group On March 19 and again on March 20, 1987, the licensee in-formed the NRC via ENS of piping supports which had failed visual examination The supports included snubbers, spring cans and rigid restraints on the residual heat re-moval system. The inspection failures have been documented on QC nonconformance reports and failure and malfunction report The licensee has not yet completed operability analysis of the applicable hanger The piping support visual inspection failures identificd during this outage is the subject of existing NRC open item 50-293/86-34-0 ,
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. Radiation Protection
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On March 16, 1987, a contractor worker received an unexpected skin radiation exposure from a small, highly radioactive par-ticle during work on the refueling floor of the reactor build-ing. The particle was detected when a portal monitor alarmed as the worker left the reactor buildin Subsequently, a health physics survey indicated that the worker had a small particle on his rib cage under his arm. The particle was not visible and had measured dose rates of 1 mrad /hr penetrating radiation and 50 mrad /hr nonpenetrating radiation with an R02A survey instru-men The licensee calculated a theoretical dose rate to the skin of 4.6 rads / hour from these measurements, assuming the radioactivity was Co-60. The particle was washed off the worker in a decontamination showe Previously, the worker had spent about two hours on the refuel floor and had helped pull a cable attached to a television camera out of the pool. The camera had been used to observe SRM and IRM dry tube cutting operations in the core area of the pool. The licensee believes that the particle may have been on the cable and penetrated the worker's protective clothing and undershirt. The licensee also investigated the offsite laundry facility that cleans protective clothing to determine if the particle could have been left in the worker's clothing from a previous laundry cycle. All work at the site using laundered protective clothing and all refueling floor work was stopped pending completion of a full evaluation, and implementation of corrective action This incident was reviewed in detail by a radiation protection specialist inspecto Results of this effort are documented in inspection report 50-293/87-17.
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The inspector reviewed documentation of multiple whole-body and extremity dosimeter values during the inspection perio Both thermoluminescent (TLD) and pocket ionization (P00) dosimeter records were examined. Workers are required to exchange their normal whole-body TLD and POD for multiple whole-body and l extremity dosimetry packs prior to work requiring the extra dosimetr Each dosimetry pack contains a TLD and a POD and is t dedicated to one monitoring location, e.g., head, chest, or
! Following the job (and before the end of a shif t), the wrist.
- worker is required to exchange the multiple dosimetry packs for their normal dosimetr Only health physics personnel are allowed to zero POD's. In addition, they read the P00's and control the dosimeter exchange Worker dosimetry records are updated with any radiation dose from the multiple pack POD's j prior to completing the exchanges.
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Multiple pack TLD's may be repeatedly reissued to the same indi-vidual for the same monitoring location prior to processin Accumulated P0D values are tracked until the TLD's are pro-cessed. The TLD's (with the exception of TLD finger rings) are processed onsite. Wrist-P0D values are used to indicate extrem-ity dose until the finger ring is processe The inspector also reviewed recent licensee evaluations of in-consistencies in POD and TLD recorded doses. These evaluations are documented on licensee Exposure Evaluation Reports (EER).
The inspector found that the multipak POD and TLD system was an acceptable method of tracking worker radiation exposure. Al-though the multipak TLD's are not always promptly processed, the .use of the P00's coupled with licensee administrative dose limits provided an administrative system to control worker exposure No trend in POD-TLD discrepancies was observe However, the following weaknesses were noted:
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There is no administrative control to ensure that finger TLD rings of workers in high dose gradient environments are promptly processed. Currently, rings may be held for a month or more before being processed. Although the licen-see extremity dose limit (5 rems per quarter) coupled with the use of wrist dosimetry provide a level of assurance against overexposure, it is conceivable that workers in a high dose gradient environment could receive undetected large finger tip exposure The inspector reviewed dose records of workers with finger rings that are currently being processed. Most showed wrist doses similar to whole body doses, indicating low dose gradient However, the licensee indicated that extremity doses rarely exceeded 3 rad However, one case involving a high dose gradient was noted. The finger rings in this case were not processed until the end of the inspection period, over a month after the exposure. The lack of either an evaluation of the worker's extremity dose based on radiation surveys or prompt processing of the finger rings is a weaknes Inadequate exposure evaluations were sometimes used to resolve POD-TLD discrepancies. The licensee's administra-tive procedures require an evaluation if POD-TLD discrepan-cies exceed 25% and are greater than 75 mrad Specif-ically, dosimetry personnel occasionally used pre printed evaluations for incidents where a POD registered a signif-icantly higher dose than the corresponding TLD. The evalu-ations allowed up to 480 mrads per quarter to be subtracted from a POD dose for accuracy and drift reasons, without requiring any followu t
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In one case, this reasoning (without a followup) was used to justify a 330 mrad difference between daily P00's which read 570 mrad in total and a monthly TLD which read 240 mrads. Although POD's can sometimes discharge and drift upscale due to moisture effects or dropping the dosimeter, adjusting POD values for these effects without additional corroborating information is a poor practice. In addition, the inspector noted that the evaluation of the 330 mrad discrepancy included an arithmetic mistake of 75 mrad. At the Exit Meeting, the licensee stated that this worker's dose would be reevaluate Also, other records completed by the dosimetry clerk who made the math error will be reviewe Procedural control of the POD-TLD evaluations was incom-plete and not always followed. For example, the evaluation form in the exposure evaluation procedure, 6.2.011, did not match the format of the currently used forms. The licensee stated that the form in 6.2.011 was marked " Typical" to indicate that other forms could be substituted. Also, the pre printed evaluations were not included in procedures or department instruction The dosimetry issue procedure, SI-RP.2402 required that dose evaluations be filled out in accordance with procedure SI-RP.2400, which has not been issued. Instead, the licen-see used an older procedure, 6.2-011, but did not strictly follow i For example, procedure 6.2-011 required that an individual be contacted whenever a TLD-P00 discrepancy evaluation had to be performed. However the individual was not contacted for the 330 mrad discrepancy discussed abov Licensee management stated that the procedure was incom-plete, and should have allowed management to sign for the individual, if he could not be contacted. The inspector agreed, but noted that there was no evidence that the
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licensee had tried to contact the individua Licensee radiation protection management above the first line supervisor level was unaware of the pre printed evalu-ation process. This method had been developed by a pre-vious management team, and was being used by the clerical staff without the knowledge of the current dosimetry management. The inspector requested the licensee to evalu-ate its dosimetry program to determine whether other forms were used without managements knowledge. No additional unauthorized forms were identifie ,
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Cosimetry discrepancy evaluations were sometimes delaye The inspector noted that several discrepancy evaluations from mid-1984 were not completed until the end of 198 Licensee personnel indicated that staff limitations in the past had slowed the evaluation process. The licensee now has a contractor conducting the evaluations. Recent eval-uations appeared timely. At the exit meeting, the licensee stated that the evaluations will be required to be com-pleted within 30 days from the date the discrepancy is identified and that health physics procedures would be changed to reflect this requiremen In summary, multipak dosimetry records were promptly updated with P00 values. The use of POD values for extended periods is acceptable provided no significant, unmonitored dose gradients exist for the jobs in questio However, programmatic weaknesses were noted in processing of finger rings and in evaluating POD-TLD discrepancie Further NRC review of the ifcensee's dosimetry program will be con-ducted under NRC open item 86-19-06 and in NRC Inspection 87-1 c. Fire Protection
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On February 19, 1987 the licensee reported via ENS that the fixed dry chemical fire suppression system for the A Emergency Diesel Generator (EDG) was found to have been inoperable since December 23, 1986. A maintenance request (MR) written on that date documented that the pressure in one of the chemical storage bottles was below the required minimum. The inspector subse-quently reviewed chemical storage bottle pressure surveillances performed between December 23, 1986 and February 19, 1987. Sur-veillance 8.A.6, Fire Extinguisher Quick Checks and Maintenance Inspections, includes steps to record bottle pressures and ver-ify that they are within the gage green range, 275 psig to 350 psig. Completed surveillance 8.A.6 dated February 6, 1987 recorded a pressure of 270 psig and noted that requirements were not met, however no action was taken. Completed surveillance 8.A.6 dated January 5,1987 also noted a below minimum pressure of 270 psig, but this value was incorrectly designated as acceptabl Similar results were noted on other completed tests. Operations surveillance 2.1.12, Daily Diesel Generator Surveillance, checks equipment which supports diesel generator i operations but also includes steps to verify chemical storage
! bottle pressures are within the green range. Several completed surveillances examined by the inspector checked off the bottle
pressure as acceptable, others noted the outstanding MR number as a discrepancy, however, in either case, no action was take One test recorded the bottle pressure as 360 psig when it was in fact only 270 psig.
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The fire water supply piping for the B diesel generator preac-tion sprinkler system includes a portion exposed to outside air temperature This section of piping is supplied with heat tracing, a temperature controller and a remote temperature indi-cator (TI). Operations daily surveillance 2.1.12 includes a check when outside air temperature is below 32 degrees F to l insure that the temperature indicated by the TI is greater than 40 degrees Maintenance Request (MR) 85-33-4, written January 14, 1985, documented that this TI was broken. The TI gage has no indicator and the MR remains open. Completed copies of sur-veillance 2.1.12 however, consistently checked applicable steps as acceptable even though it was not possible to read the TI.
t When informed, the Operations Section Head processed a procedure change to verify that a local indicating light was lit signify-ing power to the circui The inspector pointed out that the light only indicates that voltage is available to the circuit; this light will not indicate if circuit fuses fail, a disconnect switch is open, the heat trace fails or any portion of the tem-perature control circuit does not operate. Discussion with the Maintenance Section Head revealed that the Maintenance Depart-ment was unaware of the existence of the heat-traced line, and had not included it in the preventive maintenance (PM) program applied to freeze protection equipmen The inspector ques-tioned the licensee regarding system operability given that no PM had been applied to the equipment and no positive indication of operation had been observed for over two years. The inspec-tor also questioned the possible existence of other unidentified freeze protected piping. In response to the inspector's con-cerns, the licensee performed a heat tracing circuit response test to verify its operability. The Maintenance Section Head informed the inspector that the fire protection piping heat tracing would be added to the preventive maintenance progra Also, a review to identify the existence of any other missed cases will be conducte Surveillance procedure 8.B.22 is a monthly inspection of the cable spreading room halon syste The procedure requires the operator to record as-found storage cylinder pressures, cylinder temperatures, and to calculate acceptable full charge pressures using an attached graph. As-found pressures are then compared to calculated full charge pressures to assess system statu Completed surveillance 8.B.22 dated January 29, 1987 was impro-perly performed in that full charge pressures were not calcula-ted. Instead, as-found pressures were recorded in both loca-tions. Because of this, proper assessment of equipment condi-tion was not performed. A subsequent surveillance was performed and its results were satisfactor _- - __________ _ _________________ __ _ _ _ _ _ _ _ .
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Technical Specification 6.8.D requires that written procedures to implement the fire protection program be established, imple-mented and maintained. The inspector informed the licensee that the above examples of failure to properly implement fire protec-tion equipment surveillance procedures is a violation of . Tech-nical Specification 6.8.0 (87-16-01).
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During walkdowns of general fire protection system equipment conducted during this inspection period, the inspector noted that six of the thirteen divisional valves on the fire main did not have tamper proof seals installed as required. In addition, similar seals on portable extinguishers were missing. Proper installation of these seals is one of the inspection points listed in the applicabie monthly surveillance procedure Re-cently completed procedures had not noted these discrepancie Given the high percentage of seals missing on the valves, it is likely that the condition existed during the surveillance and was overlooke The above examples indicate that operations personnel are not properly performing and reviewing periodic surveillances as these activities apply to fire protection equipmen This re-sulted in several cases where the operations staff failed to recognize' inoperable equipmen In addition, the Fire Protec-tion Group may not yet be fully knowledgeable of equipment status. Maintenance Requests on the EDG dry chemical system and fire protection piping heat tracing were outstanding for exten-ded periods of time without being recognized by the Fire Protec-tion Grou In response to the inspector's concern, the Operations Depart-ment management issued a memorandum on March 13,1987 to all Operations personnel detailing the firding Included were instructions to persons implementing surveillances stressing the importance of thorough and complete performance of all step Immediately following issuance of this memorandum, the Chief Operating Engineer conducted a series of on-shift briefings with the operators reviewing this situatio The licensee also instituted a program on March 16, 1987, requiring management tours and observations of work activitie These measures may enhance personnel performance in this are In addition, the licensee's Quality Assurance Department plans to perform a number of surveillance implementation audit The inspector noted that procedure 8.8.2, Rev. 19, Figure E-7 is incorrect. The figure does not accurately display the pipe and valve arrangement between the fire water storage tanks and the fire pump suctions. NRC special inspection 50-293/86-38 iden-tified a similar error on the system P&I The licensee stated that the discrepancy would be correcte i
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On March 31, 1987, the inspector observed a fire brigade prac-tice drill. The previously poor quality of fire brigade drills was identified as a violation during NRC inspection 86-37. The inspector found that the quality of the drill observed had im-proved, and indicated that licensee efforts in this area are having an effect. However, the inspector noted some areas which still appeared to require improvement. Security was not noti-fied prior to the start of the dril During the scenario, a vital area door was blocked open, resulting in a security alar A guard was dispatched in response to this alarm. Individuals should coordinate with security prior to the drill if similar situations are anticipate The inspector also noted that techniques to be used to extinguish the fire were not fully discussed or simulated during the dril Licensee progress in establishing and improving the fire brigade drill program will continue to be evaluated, Security On April 5, 1987 at 1:35 pm, a worker found a licensee security force weapon in the bathroom in the control roo The onsite licensee security supervisor was notified and he promptly identified the guard responsible for the weapon. The supervisor discussed the incident with the Senior Resident Inspector and subsequently notified the NRC via the ENS telephone line. The guard was relieved of his duties, pending completion of a licensee review. The initial licensee evalu-ation indicated that a security compensatory measure was weakened for about 16 minutes because the guard did not have a weapon during that pericd. The inspector reviewed the area where the compensatory meas-ure had been require The licensee evaluation will be reviewed during a future routine NRC specialist inspectio .0 Review of Plant Events The inspectors followed-up on events occurring during the period to deter-mine if licensee response was thorough and effective. Independent reviews of the events were conducted to verify the accuracy and completeness of licensee informatio Unauthorized Refueling Bridge Movement On February 23, 1987, at approximately 0840, the licensee identified that the reactor refueling bridge may have been operated without authorization. At the time of discovery, the reactor had been de-fueled for over one week, the reactor cavity was flooded and the fuel pool gates were remove According to interviews conducted by the licensee, licensee personnel utilizing the bridge for control rod blade changeout terminated activities on the morning of February 2 The bridge was left positioned between the reactor cavity and spent fuel pool with the control rod blade uncoupling tool and fuel support casting removal tool suspended from two auxiliary hoists. When crews
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returned on the morning of February 23, they noted tha't the control rod blade uncoupling tool pneumatic lines were pulled under the bridge and into the reactor cavity, with -the tool lying over the cattle shoot side panel. They did not recall leaving the equipment in this condition the previous Saturda In response to the inci-dent, the licensee instituted measures enhancing controls on admit-tance to the refueling floor and on monitoring of floor activit similar incident of unauthorized bridge use occurred in September, 1985. Subsequently, the licensee's review concluded that the bridge had not been moved, but had probably been left in that condition by the crew on February 21. The bridge crew was counseled on the inci-dent. The inspector had no further question Loss of Offsite power At 8:45 am on March 31, 1987, the station lost offsite power when one of two .345 KV lines feeding the station was deenergized. Prior to the event, both 345 KV lines (the Bridgewater and Canal Station lines) were tied to the station's switchyard ring bus. However, the startup auxiliary transformer, which supplies the station's elec-trical buses when shutdown, was only being fed through breaker 352- Redundant feeder breaker 352-2 was out of service for maintenance. A fault away from the site, on the 345 KV Canal Station line, tripped the ring bus 352-3 and -4 breakers, as designed, and consequently caused the startup transformer to deenergize. The fault was later determined to be caused when a transmission tower static line was downed by a stor The A Emergency Diesel Generator (EDG) started and energized 4160 VAC safety-related bus A5 and its associated 480 VAC load center 8 The B EDG and bus A6 were out of service prior to the event for pre-ventive maintenance. After opening the Canal Station line manual disconnects, the licensee shut breakers 352-3 and 4 and restored offsite power at 9:30 a During the event, safety-related 480 VAC bus 86 failed to reenergize after the "A" EDG picked up the A6 and B1 buses. Bus 86 normally receives power from either bus B1 or B2 through series-connected cir-cult breakers52-102 and 601 or through 52-202 and 602. One set of series breakers is normally closed and the other is normally ope Upon loss of power to B6, the closed breakers trip and the open breakers are automatically closed to reenergize this bus. The in-spector noted that B6 failed to reenergize because breaker 52-202 was removed for maintenance. Specifically, interlocks are provided to prevent crossconnecting buses B1 and B2 through B6. Breaker 52-102 will not shut if breaker 52-202 is shut, as sensed by a "b" auxiliary contact on the breaker. With breaker 52-202 removed, the closing logic for breaker 52-102 was not satisfied and therefore 86 would not automatically reenergize. At 10:27 am, the breaker 52-202 auxiliary contacts were jumpered to complete the transfer logic and B6 was
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The inspector noted that operations and maintenance personnel troubleshooting the failure of 86 to reenergize were unaware that breaker 52-202 had been remove Further, the licensee found that the breaker logic functions and the automatic switching system were not considered in the development and implementation of the mainten-ance procedure used to remove the breaker, indicating a weakness in the preplanning for the breaker removal activity. A similar problem with maintenance preplanning was identified in inspection 86-4 Because the plant was in a shutdown condition and defueled, the in-spector concluded that the loss of bus 86 resulted in no significant safety impact. The licensee later replaced breaker 52-202 with a spare breaker in the test position. The inspector noted that during the evening of March 31, when restoring the Canal Station line to normal (i.e. shutting the disconnects), the B6 bus was intentionally deenergized. The bus automatically reenergized as designed with the spare breaker completing the logi In response to this event, the licensee performed a review of other automatic transfer schemes to identify similar problem None were found. The licensee also formed a root cause analysis team to review the even The licensee notified the NRC Operations Center via the Emergency Notification System as required by 10 CFR 50.72. The notification was made because the "A" Standby Gas Treatment System automatically started when the "A" EDG picked up the buses due to the initial momentary loss of reactor building radiation monitors and secondary containment isolatio .0 Review of LER's LER's submitted to NRC:RI were reviewed to verify that the details were clearly reported, including accuracy of the description of cause and ade-quacy of corrective actio The inspector determined whether further information was required from the licensee, whether generic implications were indicated, and whether the event warranted onsite followu The
following LER's were reviewed:
LER N Event Date Report Date Subject 86-025-01 11/11/86 2/23/87 Misalignment of the Fire Suppression Water System and Inoperable Fire Pumps (Update)87-001 1/12/87 2/11/87 Non-Seismically Qualified HGA Relays i
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LER N Event Date Report Date Subject l
87-002- 1/23/87 2/23/87 Inadequate Logic System Functional Testing 87-003 2/11/87 3/13/87 Two Target Rock Relief Valves Set-point Slightly Out of Tolerance 1 87-004 2/18/87 3/20/87 Improper work pri-i oritization result-ing in condition prohibited by Tech-nical Specifica-tions LER 87-004 documents the licensee's failure to recognize the effect of outstanding maintenance on the operability of B diesel generator dry chem-ical fire suppression system. The LER however, does not identify the group or groups responsible for the problem, or the reason why the over-sight occurred. This omission was discussed with liensee management during the inspector's exit intervie LER 86-025-001 supplements the licensee's initial report on loss of the fire water supply system. The inspector noted that, in addition to pro-viding further information on that event, the update also discussed a second event in which the diesel driven fire pump was declared inoperabl These two events, although somewhat similar in nature, are not directly related. During the exit interview, the inspector discussed with the licensee the inclusion of multiple events in a single LE LER 86-025-001 was also used to satisfy special reporting requirements of the Technical Specifications (TS). The LER, however, did not specify l which reporting requirements were being addressed, what information was
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required, and did not clearly satisfy the subject requirement During I the' inspector's exit interview, licensee management committed to review l the LER, establish which TS were being satisfied, ensure that all required i information was -included and to submit an updated LER if necessar . Allegation Followup: (No RI-87-A-016) Radiological problems During LpRM j Replacement On February 28, 1987, the NRC received an anonymous telephone call alleg-l ing that during Local Power Range Monitor (LpRM) replacement the backflush
! lines were rigged up backwards and the backflush was not performed. In addition, personnel under the reactor vessel were not evacuated when a
. radioactive chip reading 30 Rem / hour " fell down" during one LPRM replace-
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men Finally, the alleger stated that personnel were told not to say l anything about the incident.
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In response to these concerns, the inspector reviewed the adequacy of the licensee controls during the LPRM replacement work. The inspector noted that during removal of an LPRM on February 27, a valve manifold in a line supplying demineralized flushing water to the LPRM guide tube, was instal-led backwards. The flush is used during LPRM removal to minimize corros-ion particles from entering the instrument guide tub The manifold, which is located in the drywell, contains a check valve and manual isola-tion valv Instead of performing a flush, this configuration allowed contaminated water to reach a second valve manifold located outside the drywell in a radiologically clean area near the equipment hatc The improper configuration was identified when contamination was discovered in the area around the second valve manifold and on the clothing of a indi-viduae working with the demineralized water hoses. The inspector noted the contamination incident was documented on a Radiological Occurrence Report (ROR). Also, the manifold was properly oriented prior to reinstal-lation of the LPRM and, to prevent recurrence, was checked prior to each subsequent LPRM replacemen Just prior to installing the new instrument in place of the LPRM removed above, a radioactive chip fell into the LPRM guide tube. The particle was observed on the underwater television camera by personnel on the refuel floor and the health physics (HP) technician under the reactor vessel was told to expect it. The contact radiation level on the LPRM seal tube under the vessel was 30 Rem /hr due to this radioactive chip. However, the radiation levels rapidly dropped off with distance from the tube such that the levd was 800 mrem /hr at eighteen inches from the seal tube. The workers under the vessel were kept in a 100 mrem /hr area. The radioactive chip was quickly drained to the drywell floor drain sump (as required in the procedure) and had no effect on background radiation level The inspector did not find the actions taken inconsistent with the formal LPRM replacement training or with the pre-shif t briefing. In addition, the inspector noted that the workers under the vessel had extremity dos-imetry and the doses received on February 27, as recorded on the Radiation Work Permit using self-reading dosimeters, were small and well below regu-latory limit The inspector concluded that, while a portion of the information in the allegation was correct, the problems were not significant and the licensee took adequate corrective action in response to them. The inspector also found the licensee controls of the LPRM work, in particular, the training and HP coverage, were adequate. The inspector also noted that the prob-lems encountered during the activity were properly documented (HP log and ROR) and HP and plant management were aware of them. There was no evi-dence of any attempt to cover up the problem The inspector had no further question .
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7.0 Management Meetings At periodic intervals during the course of the inspection period, meetings were held with senior facility management to discuss the inspection scope and preliminary findings of the resident inspectors. No written material was given to the licensee that was not previously available to the publi On March 9, 1987 management representatives of the NRC Region I appeared before a special Massachusetts legislative committee on the Pilgrin Nuclear Power Plan The committee, formed by a group of state legisla-tors, conducted hearings throughout the month on issues ranging from emergency planning to the health effects of radiation. Boston Edison management was also represente Representatives of state and local governments, civil defense agencies and citizens groups have also testi-fied before the committe NRC Chairman Lando Zech toured Pilgrim on March 10, 1987. After a short presentation by Boston Edison, Chairman Zech, accompanied by Region I Administrator Thomas Murley, participated in a station tour conducted by the licensee. Following the tour, Chairman Zech discussed the status of licensee improvements with senior Boston Edison management.
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Attachment I to Inspection Report 50-293/87-16
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Persons Contacted R. Bird, Senior Vice President - Nuclear L. Oxsen, Vice President, Nuclear Operations
- K. Roberts, Station Manager 1 0. Swanson, Nuclear Engineering Department Manager N. Brosee, Maintenance Section Head T. Sowdon, Radiological Section Head N. Gannon, Chief Radiological Engineer i E. Gordon, ERHS Group Leader J. Seery, Technical Section Head R. Grazio, Field Engineering Section Manager P. Mastrangelo, Chief Operating Engineer R. Sherry, Chief Maintenance Engineer C. Higgins, Security Group Leader F. Wozniak, Fire Protection Group Leader f
- 3enior licensee representative present at the exit meetin t
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