IR 05000237/1987007
| ML17199G628 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 06/03/1987 |
| From: | Ring M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17199G625 | List: |
| References | |
| 50-237-87-07, 50-237-87-7, 50-249-87-06, 50-249-87-6, NUDOCS 8706120124 | |
| Download: ML17199G628 (17) | |
Text
U. S. NUCLEAR REGULATORY COMMISSION
REGION III
Report Nos. 50-237/87007(DRP); 50-249/87006(DRP)
Docket Nos. 50-237; 50-249 Licensee:
Commonwealth Edison Company P. 0. Box 767 Chicago, IL 60690 License Nos. DPR-19; DPR-25 Facility Name: *Dresden Nuclear Power Station, Units 2 and 3 Inspecti~n At:
Dresden Site, Morris, IL Inspection Conducted:
February 1 through May 14, 1987 Inspectors:
L. G. McGregor P. D. Kaufman S. G. DuPont D. E. Miller P. R. Rescheske R. A. Hasse Approved By:
M. A. Ring, ChiefJ.'1;.z.lo~~
Projects Section lC J
Inspection Summary Ins ection durin the eriod of Februar 1 throu h Ma 14, 1987 Re art Nos. 50-237/87007 DRP ; 50-249/87006 DRP
.
Areas Inspected: Routine unannounced resident and regional based inspection of previous inspection findings, allegation review, operational safety, followup of recent events, maintenance and surveillance observations, licensee event reports, personal contamination events, management meeting, commissioner's visit, and licensee's monthly operating repor Results:
Of the ten areas inspected, no violations were identified in nine areas. The remaining area had one identified violation - procedural deficiency (Section 2).
8706120124 870604 PDR ADOCK 05000237 G
I~
DETAILS Persons Contacted Conunonwealth Edison Company
- E. Eenigenburg; Station Manager
- J. Wujciga, Production Superintendent
- R. Flessner, Services Superintendent T. Ciesla, Assistant Superintendent - Planning R. Zentner, Assistant Superintendent* - Maintenanc_e J. Brunner, Assistant Superintendent - Technical Services J. Kotowski, Assistant Superintendent - Operations R. Christensen, Unit 1 Operating Engineer J. Almer, Unit 2 Operating Engineer W. Pietryga, Unit 3 Operating Engineer J. Achterberg, Technical Staff Supervisor D. Adam, Compliance Administrator J. Doyle, Q.C. Supervisor.
D. Sharper, W~ste Systems Engineer E. O'Connor, Radiation Chemistry Supervisor J. Mayer, Station Security Administrator W. Johnson, Chemistry Supervisor Saccomando~ Radiation Protection Supervisor M. Jeisy, Q.A. Superintendent Stols~ Q.A. Engineer The inspectors a1so talked with and interviewed several other licensee employees, including members of the technical and engineering staffs; reactor and auxiliary operators; shift engineers and foremen; electrica.l, mechanical and instrument personnel; and contract security personne *Denotes those attending one or more exit intervjews conducted on May 14, 1987, and informally at various times throughout the inspection perio Previous Inspection Findings (92701, 92702)
(Closed)
Violations (237/84026-01 and 249/84023-0l(DRP)):
A licensee identified discrepancy was not evaluated and promptly corrected as required by Quality Procedure, QP 15-51 and 10 CFR 50, Appendix B, Criterion XV In May 1983, licensee personnel identified a potential discrepancy involving the 10 CFR 50.59 evaluation format utilized in the Quality Assurance Manual and 21 months elapsed between identification of the discrepancy and identification of the violation by the NRC without any corrective actions. A review was performed and concluded that the event was an isolated occurrenc In addition, to prevent further discrepancies, the licensee added Qua.lity Assurance conunitments and NRC followup items, such as violations and open items, to the existing Dresden Commitment list to track and complete identified discrepancies in a timely manne The inspector found the licensee's actions to be adequat *
(Closed)
Unresolved Items (Inspection Reports No. 237/85039-01; 249/85036-01; and 237/84-15-05; 249/84-14-05):
Failure to impose 10 CFR 21 requirements on vendors supplying components subject to specifications unique to nuclear power plants. This item was identified during the procurement inspection conducted as part of the Dresden Safety System Outage Modification Inspection (Inspection Report Nos. 50-237/85039; 50-249/85036). A similar item has also been identified during the CECo generic procurement inspection conducted by Region III inspectors (Inspection Reports No. 50-237/84-15; 50-249/84-14).
CECo Procedure No. Q.P. 4-51, "Procurement Document Control for Operation - Processing
'Purchase Documents," stated in Section 3, "If the manufacturers standard product; however, has been demonstrated to meet all applicable special
, design requirements (seismic, radiation, environmental, etc.) the item
\\_may be purchased corrmercial grade.
Also, Exhibit C of Procedure No-:--SNED/PE Q.12, "Classification and Listing of Safety-related Items and ASME Section III Components, 11 permitted the classification of "environ-mentally and/or seismically pre-qualified" componerits as safety~related, commercial grade (10 CFR Part 21 no_t applicable.)
This led to the procurement of components subject to seismic qualifica-tion by imposing the requirements of IEEE-344, but not the requirements of 10 CFR Part Zl on the vendo~ (Purchase Orders No~ 501912, and No. 283250).
The rationale for classifying these procurements as commercial grade was that the vendors literature had stated that the components were qualified in accordance with IEEE-344 and-thus met the definition of 11 commercial grade 11 as given in 10 CFR 21.3(a-1) in that it could be ordered on the basis of specifications set forth in the manufacturer's published product descriptio However, since IEEE-344 is a standard established specifically for nuclear applications, procurements imposing its requirements do not meet the 10 CFR 21.3(a-1)
criterion of 11not subject to design or specification requirements unique to nuclear facilities 11 which must also be met to be considered 11commercial grade".
The procedural deficiencies leading to these procurements are a violation*-of-19 CFR 2~ (237/87007-01; 249/87006-01).
Since the inspection on May 21, 1986, the licensee made the following corrective actions: Quality Assurance procedure Q.P. 4-51 has been revised to require verification of 11suitability of application 11 * Q.P. 4-51 was also revised to eliminate the words, 11by the supplier" and "on the basis of its manufacture to national standards 11 to ensure that special design requirements are determined and met in accordance with 10 CFR 21.3(a-1).
In addition to the above, "Safety-Related Corrmercial Grade" is no longer being used to identify components or parts. Only "Commercial Grade 11 is being utilized in order to not confuse procurements of commercial grade items. These actions are determined to be adequate and violations 237/87007-01 and 249/87006-01 are considered to be close One violation was identified in this are.
Allegation Review A 11 egati on:
On April 27, 1987, the NRC Region III office received the following anonymous allegation: "Please investigate the Dresden Nuclear Power Station, owned by Commonwealth Edison, in Morris, Illinois. The night shift is very neglegen They play cards and sleep in the control room."
Actions Taken:
The Resident Inspector Office routinely performs off-shift inspection, including control room observation None of the recent (within the last year) Resident Office inspections have noted any indications of sleepong or card playing in the control rob In order to deliberately investigate the allegation, three resident inspectors conducted enhanced inspections over a two week period observing conduct of operations during the back shift In addition, two NRC inspectors conducted interviews of two Nuclear Station Operators (NSO) and six Shift Control Room Engineers (SCRE).
This represents one third of the total NSO and SCRE complimen During inspections, it was determined that the control roo operators (NSOs) conducted plant operations in a professional and technical manner without any observed inattentivenes Tb~ interviews revealed that no inattentiveness or unprofessional conduct was observed by shift management (SCREs) during the previous yea However, it was revea 1 ed that in the past, greater than one year ago, the non-li-censed operators (B-operators) did play card~ in the lunch room (immediately outside of the control room area) during their meal periods.. This activity has since been restricted by Dresden managemen The inspectors reviewed this previous activity and determined that it did not have any effect upon safe operations in the control room; however, the inspectors do agree with Dresden management's actions to restrict any future unprofessional activit Conclusion:
Based upon the-inspections and observations above, the inspectors were unable to substantiate the allegation and consider the concern close No violations or deviations were identifie.
Operational Safety Verification (71710, 71814, 71846, 71707)
The inspectors observed control room operations, reviewed applicable logs and conducted discussions with control room operators during this report perio The inspectors verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected component Tours of Units 2 and 3 reactor buildings and turbine buildings were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and to verify that maintenance requests had been initiated for equipment in need of maintenanc ** *
The inspectors, by observation and direct interview, verified that the physical security plan was being implemented in accordance with the station security pla The inspectors observed plant housekeeping/
cleanliness conditions and verified implementation of radiation protection control The inspectors reviewed new procedures and changes to procedures that were implemented during the inspection perio The review consisted of a verification for accuracy, correctness, and compliance with regulatory requirement The inspectors also witnessed portions of the radioactive waste system controls associated with radwaste shipments and barrelin These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under technical specifications, 10 CFR, and administrative procedure During tours in the control room, the following was observed:
During a followup inspection of the unbolted control room panels, the inspectors noted that the metal facade which rests upon the upper-most portion of the control panels was not properly bolted in place. This facade is designed to meet the same longitudinal dimensions as each section of the control console for Units 2 and Sheet metal is used as an extert~ion of this facade u~-to the control room ceiling and is fastened to the facade with small sheet metal screw The heavy sections of this facade, approximately 10 feet long by 3 feet high by 1/4 to 5/16 inches thick, are held in place, to each other and to the control console panels, by friction clamp These clamps could vibrate loose or give way through a sliding movement under normal or earthquake conditions. Should these friction clamps fail to properly secure the facade in place, the facade would be free to fall down on the control console causing damage to operator controls. This item will be referred to Region III for disposition until the licensee can justify the "use as is
condition which is contrary to plant drawing (Open Item 50-237/87007-02; 50-249/87006-02).
No violations or deviations were identified in this are Followup of Events (92700)
During the inspection period, the licensee experienced several events, some of which required prompt notification of the NRC pursuant to 10 CFR 50.7 The inspectors pursued the events onsite with licensee and/or other NRC official In each case, the inspectors verified that the notification was correct and timely, if appropriate, that the licensee was taking prompt and appropriate actions, that activities were conducted within regulatory requirements and that corrective actions would prevent future recurrenc The specific events are as follows:
- Review of Control Rod Drive (CRD)
1102 11 Phenomeno An inspection was conducted by a Region-based inspector to review licensee activities and available information pertaining to the recurring 1102 11 CRD phenomeno The purpose for the inspection was to followup on the NRC concern that Dresden appears to have a history of significant numbers of CRDs that fail to fully insert following a reactor scram (FFIS events), settling at notch position 1102 11 (six inches withdrawn). It should be noted that other BWR plants have experienced similar FFIS problems with BWR/2-5 type drive mechanism The following documentation regarding this issue was reviewed by the inspector:
0
0
0
0
0
U.S. Atomic Energy Commission (AEC), Operating Experiences Report, ROE 72-19, 11 BWR Control Rod Drive Problems,
December 15, 197 AEC Operating Experiences Report, ROE 74-6, "Malfunctions of Control Rods and Control Rod Drives, 11 March 22, 197 General Electric Company (GE) Nuclear Services Information Letter (SIL No.-52),
11Control Rod Drive Inner Filter,
January 31, 197 GE SIL No. 52, Supplement 1, 11 Control Rod Drive Settling (Inner Filter Versus Blade Interference),
11 March 29, f9-7 Commonwealth Edison letter to Mr. Darrell G. Eisenhut, NRC 11Dresden Station Units 1, 2, and 3, Quad Cities Station 1 and 2, Failure of Control Rods to Fully Insert on Scram, 11 April 14, 198 GE NSE-47-1079, 11Control Rod Drive Failure to Fully Insert Evaluation for Dresden Nuclear Power Station Units 2 & 3,
April 198 US NRC Generic Letter No. 81-24, Multi-Plant Issue B-56, 11Control Rods Fail to Fully Insert, 11 June 15, 198 GE NSE0-35-0483, 11Control Rod Drive Maintenance Review and Evaluation for the Dresden Nuclear Power Station Units 2 & 3,
April 1983..
Daily Report RIII, August 12, 1986, Event Notification, Reactor trip at Dresden Unit 2 with 55 of the 177 control rods 11 bouncing
back to position 0 GE Maintenance Instruction, GEI-92809, "Control Rod Drive Model 7RDB144B, 11 May 20,* 1986.
Dresden Procedure No. DMP 300-10, 11 Control Rod Drive Inspection and Maintenance, 11 Revision *
0
0
0
0 Dresden Procedure No. DMP 300-11, 11Control Rod Drive Leak-Test~
11 Revision Dresden Startup and Operations Letter Report for Unit 2, Cycle 1 Exxon Nuclear Company, 11 Interim Startup and Operations Report for Dresden 3, Cycle 10, 11 June 10, 198 Dresden On-Site Review Report OSR No. 85-34, review proposed changes in CRD rebuild criteria, February 18, 198 Dresden on~Site Review Report OSR No. 86-33, review criteria used to evaluate CRDs for overhaul, December 198 OSR No. 86-33, Revision 1, add six new CRDs to list of 74 drives previously selected for overhau GE letter (GEB0-7-18) to Commonwealth Edison, January 9, 1987, and attached CRD Modification Safety Evaluation, Stress Analysis Report, and Parts Lis Dresden Modification Package No. M12-2-86-49, January 23, 1987, for installation of BWR/6 type CRD Mechanisms (includes reviews, checklists, and 10CFR50.59 Safety Eval_u_ation).
The re~iew indicated that the apparent cause of the problem was excessive stop piston seal wea Leakage past the seals allows scram water to build up in the buffer area of the driv The leakage flow is too great to be accommodated by the bleed-off passages, and causes an abnormal pressure imbalance or hydraulic lock of the drives between notch positions 1102 11 and 1100.
In al 1 cases, subsequent manual insertion by the operator results in the rods being fully inserted. Technical Specification requirements regarding control rod operability and shutdown margin provide adequate assurance of the capability to place and maintain the plant in a safe shutdown conditio As evidenced by the Exxon, GE, and Dresden startup reports and core analysis (listed above), sufficient*
shutdown margin would still exist with the strongest worth control rod fully withdrawn and all other control rods at notch position 1102
- It was also concluded to be extremely improbable that control rods, following a scram, could settle to notch positions 1104 11 or beyon Consequently, NRR stated in Generic Letter No. 81-24 (June 15~ 1981), that the FFIS events do not represent a significant safety issu Extensive analysis was performed by GE for Dresden utilizing detailed records of control rod failures since 1976, and recommendations were proposed for CRD maintenance, testing, and replacemen In 1983, the licensee started to modify the stop pistons by drilling an additional bleed hole. This modification along with overhauling the 1102 11 CRDs at each refueling outage, has _apparently not
.
- alleviated the proble During the Unit 2 January 1987 outage,
.the licensee installed.12 new BWR/6 drive mechanism The BWR/6 design incorporates a new hydraulic buffer configuration, utilizes higher strength materials, and implements the latest design improvement In addition, the licensee is overhauling 68 CRD I~ acceptable performance with the BWR/6 CRDs is demonst~ated.
during the Unit 2 operating Cycle 11, the licensee plans to modify/
- convert the existing BWR/3 CRDs to the BWR/6 design. * A similar program will be implemented during the next Unit 3 refueling outag The inspector has determined that the licensee has taken appropriate actions to reduce or eliminate the 1102 11 drive proble Pending review of CRD performance during Unit 2's Cycle 11 operation, and subsequent corrective actions by the licensee, this issue will remain ope (Open Item 237/~7007-03; 249/87006-03). Unit 2 Reactor Scram During Refuelin On February 28, 1987, electrical power was momentarily lost to the 118 11 reactor protection bus, resulting in a scra Normally, in modes of operation other than refueling, the loss of one reactor protection bus would only result in half of the scram logic tripping; however, during refueling the loss of a reactor protection bus results in both channels of the "less than 600 psi vessel pressure with low condenser vacuum" logic tripping. _Jhe licensee determined that the power feed breaker (MCC 28-2) had malfunctioned, causing the loss of the protection bu The breaker was replaced and refueling was continue Manual Scram of Unit On April 9, 1987, Unit 3 was manually scrammed from 20% power as*
a result of a loss of main condenser vacuu The unit had been at 100% power when the condenser vacuum began to decreas The operators responded correctly to the loss of vacuum by reducing power and subsequently shutting the unit dow The licensee investigated the vacuum loss and made repairs~ The unit was returned to operation on April 21, 198 Reactor Scram on Low Reactor Water Level - Unit On May 3, 1987, Unit 2 automatically scrammed from 31% power on low water leve The scram occurred when an Instrument Mechanic (IM) was inserting a test module in the newly installed Feedwater Level Control Syste Because bf the small clearance between modules, the Channel 11 8
control module wa~ jarred causing the narrow range water level indications to go to full downscal Because of the false low water level, the 118 11 feedwater Regulating Valve locked-up," increasing actual water level from 20 inches (normal level) to greater than 55 inches, thus tripping the main turbine and Reactor Feed Pumps which, in turn, resulted in the low level scra The licensee reviewed the event and implemented corrective actions to prevent induced scrams as a result of inserting the test module~ such as, training of the
!Ms on testing of the Feedwater Level Control Syste *
The unit was returned to power on May 4, 198 Reactor Scram on Low Reactor Water Level - Unit On May 12, 1987, Unit 3 scrammed on low reactor water level while operating at 99%
powe The scram occurred when the "C" Condensate Booster Pump tripped on an instantaneous overcurrent. This resulted in a decreasing feedwater suction pressure and an automatic starting of the "D" Condensate Booster Pump (which was in standby) at 160 psi However, suction pressure continued to decrease to 120 psig prior to the 11 0" Condensate Booster Pump reaching full flow and discharge pressure. This resulted in the two Operating Reactor Feedwater Pumps (A and B) tripping on low suction pressure and the standby Reactor Feedwater Pump (C) starting. Because of this configuration, reactor water level rapidly decreased from normal level to the low water level scram setpoint before the operator could reduce powe The event lasted for a duration of less than eight seconds (from the initiation of "C" Condensate Booster Pump trip to the reactor scram).
Adequate reactor vessel water level for safe operation was maintained througho~t the duration of the event and the unit was placed in hot shutdow High Pressure Core Injection (HPCI) Declared Inoperable - Units 2 and On March 13, 1987, while *the unit was in the startup mode (reactor pressure at 320 psig and temperature at 402°F), HPCI was declared inoperable due to the auxiliary oil pump fa_i_ling to operate as intende The auxiliary oil pump normally operates until the HPCI turbine reaches a minimum speed of 2000 rpm and subsequently trips. However, the auxiliary oil pump tripped prior to reaching 2000 rp Unusual Event - Unit The licensee declared an*Unusual Event at 0035 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> (CST) on January 21, 1987, and began shutting the unit down from 100% powe The event was declared due to 1 of 2 contain-ment cooling loops being out of service so maintenance could be performed on a heat exchanger and the No. 2/3 Emergency Diesel Generator (EOG) being declared inoperable at 0030 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> when it failed a surveillance test. The EOG cooling water pump had a low differential pressure, thus making the EOG inoperable. Technical Specifications required the plant to be in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of declaring 1 EOG inoperable with 1 containment cooling loop out of servic The State and local officials and the Senior Resident Inspector* were notified by the license The Unusual Event was terminated at 0510 hours0.0059 days <br />0.142 hours <br />8.43254e-4 weeks <br />1.94055e-4 months <br /> on January 21, 1987, when the contain-ment cooling loop was placed back in servic Scram in *cold Shutdown - Unit 3. Unit 3 had just completed a 13 day maintenance outage and was preparing for reactor startup when a reactor scram occurred at 0526 hours0.00609 days <br />0.146 hours <br />8.69709e-4 weeks <br />2.00143e-4 months <br /> (CST), on March 11, 198 The unit was in cold shutdown and no control rod movement occurre The scram was initiated as one of the Average Power Range Monitor (APRM)
channels was being removed from the bypass position. This type of operation normally produces a half scra The licensee has issued a work request to investigate the reason for receiving a full scra "
- Scram in Cold Shutdown - Unit With the unit in cold-shutdown, a reactor scram signal was received at 0134 hours0.00155 days <br />0.0372 hours <br />2.215608e-4 weeks <br />5.0987e-5 months <br /> (CST), on March 10, 198 No rod movement occurre The scram signal occurred during the performance of an ECCS undervoltage test which requires two scram ~ignals (reactor pressure and low condenser vacuum) to be bypasse When electricians installed the first jumper a half scram occurred as expected; The half scram signal was not reset as required due to miscommunications between the electricians and the control room operator. Thus, when a second jumper was installed, another half scram signal was generated, completing the logic for a full scram signal. The scram signal was reset and no other systems actuate Reactor Scram - Unit On March 21, 1987, at 1125 hours0.013 days <br />0.313 hours <br />0.00186 weeks <br />4.280625e-4 months <br /> (CST),
with the unit at 98.5% power, a Generator Load Reject scram signal, followed by Group II and III isolations, was receive The signal was generated when a contractor jarred a panel in the 2/3 Auxiliary Electric Room while pulling cables through a cable tray. The impact caused SGC 12A main generator neg~tive sequence relay to trip, thus causing the scra The root cause of the event has *been attributed to personnel error; Reactor Scram While Shutdown - Unit On March 21, 1987, at 1152 hours0.0133 days <br />0.32 hours <br />0.0019 weeks <br />4.38336e-4 months <br /> (CST), while recovering from an earlier reactor scram at 1125 hours0.013 days <br />0.313 hours <br />0.00186 weeks <br />4.280625e-4 months <br />, Unit 3 reactor scrammed on low reactor water level. Reactor feedwater pumps (RFP)
11A 11 and 11 8 11 tripped on high reactor water level during the initial scram at 1125 hour0.013 days <br />0.313 hours <br />0.00186 weeks <br />4.280625e-4 months <br /> Once the high level trip signal cleared the NSO prepared to start the 11 8 11 RFP but was unable to open the 118 11 discharge valve due to the high differential pressure across the valve disc created by the running pump's dis-charge pressure. Thus, operation of the 11 8 11 RFP was immediately terminated and just as the 11A 11 RFP was started, the reactor scrammed on low reactor water level and Primary Containment Group II and III isolations were receive The scram was immediately reset and reactor water level restored to norma The total time evolution for restoring feedwater pump operation was approximately 5 to 10 minute No violations or deviations were identified in this are.
Monthly Maintenance Observation (62703, 71710)
Station maintenance activ1ties of safety related systems and components listed below were observed/reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with technical specification The following items were considered during this review:
the limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were
- performed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and, fire prevention controls were implemente Work requests were reviewed to determine status of outstanding jobs and to assure that priority is assigned to safety related equipment maintenance which may affect system performanc The following maintenance activities were observed/reviewed:
D 42569 2-2499-3A D 42573 2-2499-lA D 60895 D2 X area D 58926 Control Rod Drive D 53242 D 53243 D 53244 Rebuild Solenoid and inspect for Environmental Qualification Rebuild Solenoid and inspect for Environmental Qualification Remove insulation in X area as ~equeste Inspection of electrical control and instrument cables and general cleanliriess of area.. Some liquid tight conduit was damaged and licensee is replacing sam Overhaul of the Control Rod drives on list (D58925).
Witnessed maintenance and rebuild of seven control rod Replace Steam Separator Replace Reactor Steam Dryer Replace Reactor Head D 59113 HPCI 2301-35 Tested motor operated valv Required per IE Bulletin 85-03 11Motor-Operated Valve Common Mode Failure During Plant Transients Due to Improper Switch Setti_ngs
- D' 62971 2/3 A Rad Electrical Motor fire. Replaced Moto Waste Sparging Air Compressor D 63578 Accumulator Repair leaking 126 Valve (Packing).
26-55 D63588 Accumulator Repair leaking 126 Valve (Packing).
18-83 D 60264 Reactor Water Replacement of inboard sea Clean Up Pump 2B
D 62196 Torus Vacuum Inspector visually inspected the torus vacuum Breakers breakers l-1601-32C and 32 The inspectors also inspected Unit 2 torus for internal cleanliness and touch-up painting. All areas were clean with additional attention given to the suction heade D 58083 Unit 3 Diesel Inspection of repaired soak back oil pump fitting which was leakin The inspectors reviewed the repair of Unit 2 Isolation Condenser support grids within the condense A final cleanliness inspection was also conducted by the resident inspector No violations or deviations were identified in this are.
Monthly Surveillance Observation (61726)
The inspectors observed surveillance testing required by technical specifications for the items listed below and verified that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that limiting conditions for operation were met, that removal and restoration of the affected components were accomplished, that test results conformed with technical specifications and procedure requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies Tdentified during the testing were properly reviewed and resolved by appropriate management personne Th~ inspectors witnessed portions of the following test activities: Unit 2 HPCI The inspectors witnessed portions of HPCI surveillances including, DOS 2300-1 11 HPCI Motor Operated Valves and Pump Operability," DOS 2300-3 11 HPCI System Pump Test," and a special reliability test in which the HPCI turbine was quick started from cold condition DOS 250-5 "Automatic Bl owdown System at Low Pressure and Rated Pressure" for Units 2 and DOS 1100-1 "Standby Liquid System Pump Test" Unit 2 DOS-7500-2 DIS 2300-1 Moisture Removal from Standby Gas Treatment Charcoal Absorbers (completed 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> test) "A
Fram HPCI Steam Line Hi Flow Isolation Master Trip Calibration.
- DIS 1500-5 DIS 1400-1 DIS 1500-6 DOS 6900-6 DOS 700-1 DIS 1100-3 DIS 260-4 1 DOP 1000-1 1 DMP 300-16 1 DIS 1700-2 1 DIS 5600-5 Unit 3 DIS 700-4 DIS 2300-2 DIS 1500-3 DIS 600-4 DIS 2400-1 DIS 500-1 DIS 700-6 DIS 1200-1 DOS 500-2 1 DOS 500-8 1 DIS 700-8 LPCI Containment Cooling System Logic Test Procedur Core Spray Header Differential Pressure Instrumentation Calibratio LPCI System Recirculation Loop Break Detectio Volt Station Battery Capacity Tes SRM In-Op Rod Block Tes Standby Liquid Control Pump Suction Temperature Calibratio *
ATWS Transmitter Calibration and Logic Tes Filled and Vented "2A" Shutdown Cooling Heat Exchange Scram Pilot Testin Off-gas Rad Monitor Function Tes Turbine Trip Function Tes IRM Rod Block/Scram Calibration Tes HPCI Turbine Permission Tes Reactor Water Level 3 Core Height Indicator Calibration and Master Trip Uni Reactor Wide Range Pressure Transmitter Calibratio Drywell Hydrogen Monitor Calibratio Reactor High Pressure Scram Calibratio APRM Flow Bias Scram, Rod Block Calibratio RWCU "B" Leak Detection Survei 11 anc Rod Block Monitor Function Tes MSIVs Not Full Open Scram Tes Rod Block Monitor Calibratio *
1 DIS 1700-7 Reactor Building Vent and Refueling Room Floor Rad Monitor Function Tes No violations or deviations were identified in this are.
Licensee Event Reports Followur (93702)
Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with technical specifications:
Unit 2 (Closed) 87002-00:
Unit 2/3 Radwaste Off Stream Liquid Effluent Monitor Isolated During Radwaste Discharge Due to Personnel Erro Review of this event is documented in Region III Inspection Reports 50-237/87003 and 50-249/8700 (Open) 87004-00:
Unit 2 Primary Containment Type 11 B 11 and 11 C 11 Local Leak Rate Test Limit Exceeded Due to Excessive Leakage Through Primary Containment Isolation Valv Review of this event is being conducted by a Region III Specialis (Closed) 87006-00:
Reactor Scram During Refueling Due to*Motor Control Center (MCC) 28-2 Main Feed Breaker Trip at Bus 2 Review of this event is documented under Paragraph 5.b of this repor Unit 3 (Closed) 87002-00:
High Pressure Coolant Injection (HPCI) System Inoperative Due to Oil Pressure Regulation Valve Failure. Review of this event is documented under Paragraph 5.f of this repor The preceding LERs have been reviewed against the criteria of 10 CFR 2, Appendix C, and the incidents described meet all of the following requirement Thus no Notice of Violation is being issued for these item The event was identified by the licensee, The event was an incident that, according to the current enforcement policy, met the criteria for Severity levels IV or V violations, The event was appropriately reported, The event was or will be corrected {including measures to prevent recurrence within a reasonable amount of time), and The event was not a violation that could have been prevented by the licensee's corrective actions for a previous violation.
- Unit 3 (Closed) 87003-00:
Reactor Water Temperature Exceeds the Technical Specification Limit of 212°F With Primary Containment Not Established Due to Personnel Erro Review of this event is documented in Region III Special Inspection Report No. 50-237/87011 and 50-249/8701 One violation associated with LER 87003-00 was identified in the review of this are.
Personnel Contamination Events Personnel contamination events have been tracked quantitatively by the licensee over the last three years and reviewed during several NRC inspections (237/86008; 249/86010; 237/86023; 249/86028; 237/87003; 249/87003). *The reported rate (time specific) of personnel contaminations at the Dresden Station appears to be inordinately hig During 1986 the licensee reported 1760 personnel contamination incidents using the INPO criteria of skin or personal clothing contaminated such that when monitored with a pancake probe frisker a response of 100 counts per minute is detecte (The licensee had originally reported about 2500 personnel contamination incidents using a more conservative criteria than specified by INPO.)
The licensee recently formed a committee to review the personnel contamination incidents and develop an action plan aimed at reducing the number of incident The committee was composed of several station managers and supervisors from various departments including rad/che The committee compiled their recommendations into a draft report which was recently submitted to the Station Manager for consideratio The draft report addressed several subjects including the following summarized recommendations:
0
0
0 Greater participation/coordination of foremen in job preparation/
planning and oversigh Foreman would accompany rad/chem technician during the radiological survey of job site; foreman would then discuss radiological controls with work cre Increased use of tailgate sessions to enhance worker awareness of contamination control Increased routine smear surveys, better RWP surveys, and increased routine cleanin Increased management tours/oversigh Review need for and use of "gang boxes", and alternate method Review need for better tool and equipment decontamination facilities/
equipment, and centralized storage/control of decontaminated reusable material,-
0
0
Review need for increased use of glove boxes and other containment Revi,ew need to post survey information at job sites and 1 imit numbers of workers in an are Increase use of personal contamination incident trending by department and tas Make greater us~ of information gained from interviews of persons contaminate. Increase training provided to workers concerning contamination and
.contamination control Establish/assign ALARA coordination responsibilities to selected members of individual department The above recommendations are being considered by station managemen The licensee is also considering alternate methods of personal contamination event trendin The alternate methods would compare number of events to the quantity of radiological work performed during an established reporting perio The licensee's actions in response to committee recommendations will be reviewed during future routine inspections of the licensee's contamination control progra (Open Item 50-237/87007~0~(DRSS);
50-249/87006-04(DRSS))
No violations or deviations were identifie.
Management Meeting An enforcement conference was held on February 26, 1987, at the NRC, Region III Office in Glen Ellyn, Illinois. The enforcement meeting was to discuss the Safety Systems Outage Modification Inspection which identified a significant weakness in the control of plant modification Details of the inspection are addressed in Inspection Report N /86009 and 50-249/8601 il. Commissioner's Visit On March 25, 1987, Commissioner James K. Asselstine visited the Dresden Nuclear Station. While on site, the Commissioner held meetings with the NRC resident staff, as well as with licensee management and supervisory personne The Commissioner toured the facility, including the radiation waste facility, reactor buildings for Units 2 and 3 and the control room, observing the performance of the shift operating cre.
Report Review During the inspection period, the inspectors reviewed the licensee's Monthly Operating Report for March and Apri The inspectors confirmed that the information provided met the requirements of Technical Specification 6.6.A.3 and Regulatory Guide 1.1 The licensee announced the following Dresden site management changes:
Dave VanPelt transferred in as Assistant Superintendent Maintenance, effective March 30, 198 Randy Zentner, Assistant Superintendent Maintenance transferred to Office of Executive Vice President, effective April 17, 198 Bob Geier, Senior Engineer, transferred to Quality Control Supervisor, effective April 6, 198.
Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or bot Open items disclosed during the inspection are discussed in Paragraphs 4, 5.a, and Exit Interview (30703)
The inspectors met with licensee representatives (denoted in Paragraph 1)
informally throughout the inspection period and at the conclusion of the inspection on May 14, 1987, and summarized the scope and findings of the inspection activitie The inspector also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspectio The licensee did not identify any such documents/processes as proprietar The licensee acknowledged the findings of the inspectio