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| {{Adams | | {{Adams |
| | number = ML20216C743 | | | number = ML20236N496 |
| | issue date = 06/23/1987 | | | issue date = 08/04/1987 |
| | title = Safety Insp Repts 50-325/87-13 & 50-324/87-13 on 870505-31. Violations Noted:Failure to Follow Integrated Leak Rate Procedure.Fuses for High Drywell Pressure Instrument Not Removed | | | title = Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Repts 50-324/87-13 & 50-325/87-13 |
| | author name = Fredrickson P, Garner L, Ruland W | | | author name = Reyes L |
| | author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) | | | author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| | addressee name = | | | addressee name = Utley E |
| | addressee affiliation = | | | addressee affiliation = CAROLINA POWER & LIGHT CO. |
| | docket = 05000324, 05000325 | | | docket = 05000324, 05000325 |
| | license number = | | | license number = |
| | contact person = | | | contact person = |
| | document report number = 50-324-87-13, 50-325-87-13, NUDOCS 8706300350 | | | document report number = NUDOCS 8708110523 |
| | package number = ML20216C675 | | | title reference date = 07-24-1987 |
| | document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS | | | document type = CORRESPONDENCE-LETTERS, NRC TO UTILITY, OUTGOING CORRESPONDENCE |
| | page count = 11 | | | page count = 1 |
| }} | | }} |
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| | AU604 987 |
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| ' UNITED STATES -
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| / NUCLEAR REGULATORY COMMISSION.-
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| REGloN 11 :
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| 101 MARIETTA STREET.N.W.
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| Report'Nos. 50-325/13 and'50-324/87-13 n-
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| ? Licensee:
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| Carolina Power and: light Company--
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| P. O. Box 1551-4
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| .Raleigh,-NC.27602 l
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| IDocket'Nos. 50-325 and:50-324-License Nos. DPR-71 and.0PR-62
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| ' Facility Name: ' Brunswick 1 and 2-
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| Inspection Conducted: 'May.5.- 31,.1987-b-NN '
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| Inspecto
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| W. H. RVland Date Signed
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| L.: W. Gh+ er Date Signed..
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| 'ApprovedByh f )fW
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| - P. I., Fredridkson,' Secti.on. Chief.
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| .Date. Signed-l
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| ~' Division of-Reactor Projects
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| ~ SUMMARY
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| . This routine safety inspection involved the ~ areas - of maintenance Scope:
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| . observation, surveillance ' observation, operational safety;-l verification, ESF'
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| System walkdown,.and. Unit'2 forced outages.
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| 3, Results: One. violation was-identified: Failure to follow'i.ntegrated leak rate
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| procedure -' the. fus'es for a high.drywell pressure. instrument 'were not removed.
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| B706300350 870'624
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| . PDR-ADOCK 05000324 G-PDR
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| | ' Carolina Power and Light Company ATTN-9 r. E. E. Utley |
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| | Senior Executive Vice President j |
| | Power Supply and Engineering j |
| | and Construction-P. O. Box 1551 Raleigh, NC 27602 Gentiemen:- |
| | l SUBJECT: REPORT NOS. 50-325,324/87-13 Thank you for y' |
| | response of July 24, 1987 to our Notice of Violation,' issued on June 24, 1981, concerning activities conducted at your 8runswick. facility. |
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| | We have evaluated your response and found that it meets the requirements of 10 CFR 2.201. |
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| | We will examine the implementation of your corrective actions |
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| | during future inspections. |
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| DETAlLS
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| 1.
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| Persons Contacted
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| Licensee Employees
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| A._.
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| P. Howe, Vice Precident - Brupswick Nuclear Project C. Dietz, General Manager - Brunswick Nuclear Project i
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| T. Wyllie, Manager'- Engineering and Construction" (
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| G. Olivar, Manager - Site Planning and Control J. Holder, Manager - Outages R. Eckstein, Mana r% Operatior.sr - Technical Support, t:
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| E. Bishop, Manager, L.-Jones, Director's, Quality Assuchnce (QA)/ Quality Control (QC)
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| R. Helme, Director - Onsite Nuclear Safety - BSEP J. O'Sullivan, Manager - Maintenance G. Cheatham, Manager - Env Wonmental & Radiation Control
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| J. Smith, Manager m Admir,istrative Support
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| K. Enzor, D1 rector -iRdgufatory Compliance
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| R. Groover, Manager 2 Project Construction
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| s V. Wagoner, Director xIPP,S/Long-Rangq Planning A. Hegler, Superintendent - Operations 1
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| W. Hogle, Engineering Supervisor
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| B. Wilson, Engineerinn Supervisor
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| B. Parks, Engineeringi$upervisor s-x
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| W. Biggs, Principal Engineer 7-R. Creech, I&C/ Electrical Maintenance Supervisor (Unit 2)
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| d R. Warden, I&C/ Electrical Maintenance Supervisor (Unit 1)
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| t t W. Dorman, Supervisor QA,
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| W. Hatcher, Supervisor Security
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| L R. Kitchen, MechanicalJ aintenance Supervisor (Unit 2)
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| C. Treubel, Mechanical faintenance Supervisor (Unit 1)
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| EJ R. Poulk, Senior NRC Regulatory Specialist w
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| W. Murray, Senior Engineer %y Specialist
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| D. Novotny, Senior Regulator
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| Nuclear Licensing Unit
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| Other licensee employees contacted included construction craftsmen, engineers, technicians, operators, office personnel, and security force members.
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| L 2.
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| ExitInterview(30703)
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| The inspection scope and findings were summarized on May 29, 1987, with the
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| general manager.
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| One Violation, failure to meet initial conditions
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| required by the Unit 1 integrated leak rate test (paragraph 6), was discussed in' detail.
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| An Unresolved Item (paragraph 8.b)), concerning an inadequate procedure, was also discussed.
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| The licensee acknowledged the findings without exception. The licensee did not identify as proprietcry anyt, of the materials provided to or reviewed by the inspectors during the inspection.
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| 3.
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| Followup on Previous Enforcement Matters (92702)
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| Not inspected.
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| F The inspectod observed maintenande1 activities and ' reviewed records to
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| :verifhthat twor.K ' 's conducted ih accordat:ce withiapproved procedures,
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| Techn11al Specifi$ 'tions, andr' applicable industry. codes and standards. The
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| "inspecWrs. also verifjedithat:. dedundanticomponents were operable;
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| dministrit}ye; c6htlols' were followedptagouts were adequate; personnel
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| y6 pre qualifded;' correct; replacement.partFw%re.lised;- radiological co'ntrols-J
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| .were proper; fire. protection was ade'qQte;- qdality control hold ; points
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| .were.:Ladequate,and. observed; adequat9 post-maintenance' testin'g; was
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| N performed; and' independent verification requiremsnts were L implemented.
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| The. inspectors independently verified'that selected equipment.was properly.
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| 1 returned toLservice.
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| Dutstandingiwork requests: wyre re'vieQd to knsureL that' the' licensee gave c-priority' to safe'ty-related maintenanc i,
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| 'The Linspectors observed / reviewed portions. of thifoHowingL maintenance-
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| activities:
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| ;M 3. A SG-BNNS1 h
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| ,Repla ement of High Press ~ure Coolant Injection (HPCI)
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| a : tiydraulic Actuator Gear Drive Gears,.
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| 87-ANDX1-Disassemble 1-E41-F006 to Determine Cduse of Low' Actuator-g
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| ' Repair ofz24821-TR-R614, Safety Relief Valve (SRV) Taiipipe.
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| ~87-AQIF1 #
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| Temperature Chart Recorder.
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| -MI-10-517C~
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| HPCI Hy'draulic Actuator Gear Drive and Speed Pickup Gear Assembly Inspection.
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| % nit 1 Mot 6r Operated Valve Program!
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| r The' licensee contracted w'ith Babcock and Wilcox '(B&W): to take valve
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| signatures and perform 1 diagnostics on 30 valves.
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| Twenty.seven of these are on the. HPCI and Reactor.. Core : Isolation-Cooling '(RCIC)-
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| l i3~.f systems. "The ' device used tis designated as a Motor Actuator Characte -
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| . rized ?(MAC).
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| The MAQ machine geasures and/or. calculates spring.
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| f(Qand'orqueswitchactuations>
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| orce,. stem thrust, worm shaft displacement, motor current, and limit
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| The raicro processor data is capable of.
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| R being' transferred into; graphic-form and/or hard copy.
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| The project
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| ] 'Nas completed at-the and of the report period.. The licensee has
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| ' repaired those valveshand/or actuators -which were - identified as
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| having ' deficiencies. The licensay has'. supplied to the.' i n spector,
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| copies of the MAC data 'for these' valves. The ~ inspector plans to i
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| review this data and document <that review as well as a summary of the
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| . problems' which required: correhtion in a future report.
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| This.is an
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| f Inspector Followup Item:
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| Revi s Valve. Diagnostic Test Results 325/87-13-01).
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| The inspector witnessed connection of sensors and/or testing of the HPCI suppression pool inboard suction isolation. valve, E41-F042,. the HPCI steam admission valve, E41-F001 and the HPCI inboard injection valve ' E41-F006.
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| During performance of the F006 valve test,.the measured stem thrust indicated. about one half of that specified by-the manufacturer (Limitorque).
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| Subsequent testing on other valves showed Lsimilar results.
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| The licensee, B&W and Limitorque have concluded that on large and/or fast actuating actuators, the correla-tion between the measured parameter, strain gauge output, and stem thrust is not correct. B&W, in cooperation with Limitorque, -.will-develop a new ccrrelation.
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| The licensee plans to have the stem thrust information on the affected actuators re-evaluated when the new correlation is available. The inspector will review the data as part of.the above IFI.
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| b.
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| Incorrect Spring Pack in Residual Heat' Removal (RHR) Valves ihe licensee declared both divisions of Suppression Pool Cooling.
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| inoperable for Unit 2 based on problems found in Unit'l Limitorque motor operators.
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| Units 1 and 2 were in operational conditions. 5 (refueling) and 1 (power operation), respectively.
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| The licensee had replaced the torque switch for valve 1-E11-F024A,
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| Unit 1, Division I, Suppression Pool Cooling Isolation Valve, on l
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| April 5, '1987. The licensee had found the problem with the torque switch while trying to cycle the valve after routine preventative maintenance on the motor control center.
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| The valve was stopping '
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| .immediately.after dual indication was seen on the. main control board
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| while shutting the valve. Main tenance, under work request 87-AKUP1, found the. torque switch contacts pitted and with high resistance.
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| -The valve was cycled and the motor-operator inspected. af ter the torque switch was replaced;:the results were satisfactory.
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| The 1-E11-F^24A valve then failed.to completely shut during perfor-mance of the quarterly RHR system operability test, PT-8.'2. 2. c, on
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| 'May 2, 1987.' The licensee initiallyLfound the torque out current too
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| low' with the required torque switch setting of 2.5.
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| The licensee
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| continued tn troubleshoot the problem the next 20 days' in between l
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| vesse t ~ hydrostatic test. Integrated Leak Rate Test (ILRT), replace-
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| ment of au/iliary contractor work, and work on the Unit 1 HPCI F006 i
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| valve. 'On May 15. 1987, maintenance found that the~ handwheel torque'
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| required to open1the torque' switch in the' shut direction was too low and they suspected a bad spring pack. After the ILRT, on May 22, the licensee found the spring pack to be incorrect: it had 9 believille washers instead of 11.
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| Since all ' four F024 valves '(' both units) were originally purchased
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| under 'the same order, the licensee inspected the Unit 1 F024B valve and found a liaht. spring pack also installed.
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| The licensee then i
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| i declared the vn.. 2 valves iroperable at 2:30 p.m. on May 22, 1987.
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| The licensee concluded that a lighter than'specified spring pack was-installed in the Unit 2 F024A valve and adjusted the torque switch from 2-1/2 to 3-3/4 to compensate for it. The licensee has had the concurrence of Limitorque for the adjustment. The licensee concluded
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| .that ' the F0248 valve in Unit 2 has' a heavier spring pack installed
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| with characteristics similar to the specified' spring pack.
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| The licensee determined the spring pack characteristics of the Unit 2 valves by-observing valve handwheel torque when the torque switch operated while closing the valve.
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| The licensee had returned the Unit 2 F024A to operable status'at 9:00 p.m. on May 22 and the F024B at 12 midnight on May 22, complying with
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| .the applicable sections of technical specifications.
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| New spring packs will be installed when available.
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| The licensee has reported that they have no readily accessible record l
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| of. ever replacing the spring packs, indicating that the valves'may have been delivered with spring packs not matching Limitorque's own documentation.
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| The vendor program branch will followup on the generic issues rair,ed by the event. The inspectors will continue to j
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| follow the licensee's actions in the Motor Operated Valve'(MOV) area-as described in paragraph 4.a.
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| No violations or deviations were identified.
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| 5.
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| Surveillance Observation (61726)
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| The inspectors observed surveillance testing required by Technical'
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| Specifications.
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| Through observation and record review, the inspectors
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| , verified that:
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| tests conformed to Technical Specification ' requirements; administrative controls were followed; personne1' were qualified; instru-mentation was calibrated; and data. was accurate and complete.
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| The inspectors independently verified selected test results and proper return to' service of equipment.
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| The inspectors witnessed / reviewed portions of the following test activi-I ties:
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| IMST-ADS 23R Automatic Depressurization System (ADS) Safety Relief Valve Primary Position Channel Calibration.
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| IMST-APRM29Q - Average Power Range Honitor (APRM) Flow Bias Flow Units C&D l
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| Channel Calibration.
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| l 1MST-RPS34R Reactor Protection System (RPS) Main Steam Line Isolation Valve Closure Circuit Response Time.
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| 2MST-HPCI27M HPCI and RCIC Condensate Storage Tank Low Water Level
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| Instrument Channel Calibration.
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| PT-20.6 Drywell to Torus Leak Rate Test.
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| - _ - - -_-_ _ _____-__. _ - _ _ _ - _ - _ _ _ _ _ - _ _ - - _
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| 6'
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| PT-78.2 Liquid Radwaste Radioactivity Effluent Monitor Channel Calibration (012-RM-K604).
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| During performance of IMST-RPS34R on May 14,.1987, the inspector observed
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| that two procedural steps were incorrectly performed. Step 7.4.7 requires l
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| test leads to be connected to test panel points TPB-31 and 33; instead, I
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| they were attached to.the adjacent column of test points.
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| This resulted -
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| in a failure, on the first attempt, to obtain the necessary data to -
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| determine the response time. During troubleshooting of the failure to-obtain data,' the system.was partially restored to-normal by reinstalling I
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| fuse C71A-F3F. Upon detection of the improperly connected test leads, a
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| technician resumed the procedure at the beginning of section 7.4.
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| However, step '7.4.10,. remove fuse C71-F3F, was overlooked.
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| Thus, 'a subsequent run of the test again failed to measure the response time.
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| This deficiency was corrected.
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| In addition, the inspector also observed a technician correct a step he i
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| had just indicated as'not applicable.
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| Step 7.4.35 is not applicable if j
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| the reactor mode switch is in RUN. - The technician initially indicated the j
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| step as not applicable, and then realized his mistake and performed the
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| step (the mode switch was in REFUEL at the time).
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| l None of.the above items had a potential for resulting in an incorrect L
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| recnonse time test.
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| Each one of the above items'was associated with a
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| .j L
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| different technician.
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| These items were discussed with maintenance l
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| l supervision.
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| Perhaps, because the unit was shutdown, the technicians may i
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| l-have been less attentive.
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| This concern was expressed to management.
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| L Maintenance management stated 'that the items would be discussed with
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| .j members of the surveillance staff to ensure that a " shutdown mentality"
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| was not being developed.
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| No violations or deviations were identified.
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| 6 '.
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| Operational Safety Verification (71707)
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| j The inspectors verified conformance with regulatory requirements by direct observations of activities, facility tours, discussions with personnel, reviewing of records and independent verification of safety system status.
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| l The inspectors verified that control room manning requirements of 10 CFR l
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| 50.54 and the technical specifications were met.
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| Control room, shift-
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| supervisor and clearance logs were reviewed to obtain information
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| concerning operating trends and out of service safety systems to ensure that there were no conflicts with Technical Specifications Limiting-l Conditions for Operations. Direct observations were conducted of control
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| room panels, ' instrumentation and recorder traces important to safety to l'
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| verify operability and that parameters were within Technical Specification limits. The inspectors observed shift turnovers to verify that cnntinuity of system status was maintained.
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| The inspectors verified the status of selected control room annunciators.
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| .....
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| _ _ _ _ _ - _.
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| | q We appreciate your cooperation in this matter. |
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| | Sincerely, Original Signed by l |
| | Luis A. Reyes |
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| | Luis A. Reyes, Director Division of Reactor Projects cc: @.W. Howe,VicePresident |
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| | Brunswick Nuclear Project |
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| ;0perability' of. selected Engineered Safety Feature (ESF) trains were
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| , | | p R. Dietz, Plant General Manager bec: |
| verified by. insuring that: each accessible valve in the flow path was in
| | BBCResidentInspector |
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| | ' Document Control Desk |
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| | .i State of North Carolina J |
| its: correct: position; 'each power supply and breaker, including control-
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| ! | | ! |
| | RII RI! |
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| l room, fuses, were. aligned;for components that must activate upon initiation signal;.removallof power from those ESF motor-operated valves, so identi-fied by Technical Spec.ifications, was completed; there was no leakage of
| | RII RII SVi'as:vyg PFr dr'ckson DVe'rrelli hWBrownlee |
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| major components; there =was proper lubrication and ' cooling water available; and a condition did not exist which might prevent fulfillment-
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| .of? the' system's L functional requirements.
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| Instrumentation essential.to
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| ~ system actuati.on or performance was verified operable: by observing l
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| on-scale : indication and proper instrument' valve lineup, if accessible.
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| Thel inspectors verified that the licensee's health physics policies /
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| procedures were followed. This included.a review of area surveys, radia--
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| tion work permits, posting, and instrument calibration.
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| LThe inspectors verified' that:
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| the security' organization was properly
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| ~.
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| manned and security personnel were capable of performing their assigned functions;. persons :and packages were checked prior to entry into the-
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| protected area. (PA); vehicles were properly authorized, searched and-
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| escorted.within the PA; persons within the PA displayed photo identifica-
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| :)
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| tion badges;; personnel in vital areas were authorized; and effective q
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| compensatory ' measures were employed when required, n
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| The ' inspectors also observed plant housekeeping control s,. verified.
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| position of certain: containment isolation valves, checked a clearance, and verified the operability ~of onsite and offsite' emergency: power sources, a..
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| 'High Drywell Pressure Instrument Left Energized During ILRT
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| 'The inspector found that Unit ' 1 Drywell Pressure'.High Instrument:
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| 1-E11-N011A was reading greater than 5 psig while N011B, C & D -were less than 0 psig during an ILRT with containment pressure at about 48 psig at 3:00 p.m. on May 18, 1987.
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| The inspector questioned the.
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| d licensee concerning the discrepancy. The licensee reported that..the
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| 'l
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| fuses should have been pulled for -all those instruments during the
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| ILRT lineup. On May 19, 1987, the licensee informed the inspector that due to a labeling problem, the wrong fuses had been removed.
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| The licensee had removed the fuses for the Core Spray and Residual i
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| Heat Removal Injection Permissive Reactor Steam Dome Pressure Instru-i ment 1-B21-N021A instead of the E11-PT-N011A fuses. :The fuse labels were reversed in the back of the Emergency Core Cooling System (ECCS)
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| cabinet 1-XU-63 that contained both instruments.
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| The ILRT was not affected by the error. No valves changed positions r
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| as a result of the error. The N011A instrument could not by itself
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| cause an inadvertent actuation and was only required to be operable j
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| in conditions 1, 2 and 3.
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| The unit was in operational condition 5
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| (refueling).during the ILRT.
| | 8/,3 /87 8/g/87 8/3/87 8/ @7 8708110523 070004 DR ADDCK 0500 |
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| -i The steam dome pressure instrument, B21-PT-N021A, had been inadver-tently de-energized, was required to. be operable, per technical specification-(TS)'3.3.3, in operational condition 5.
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|
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| The applicable i
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| action statement requires the licensee to declare the associated ECCS systems inoperable. Both Core Spray (CS) and RHR Division I should have been declared inoperable. The CS ACTION statement in TS 3.5.3.1 for operational condition 5 requires that at least one Low Pressure Coolant Injection (LPCI) system is operable within four hours.
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| The licensee had complied with the TS based on the inspector's review of the' operator's log and the ILRT procedure.
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| The fuses were not labeled in accordance with the applicable drawing-for the XU-63 cabinet, F-39031, sheet 6, Revision 8.
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| Fuses in block RF, which were labeled B21-F1-A and B21-F2-A on the drawing, were labeled E11-F1-A and E11-F2-A in the. cabinet.
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| Fuses in block RT, which were' labeled E11-F1-A and E11-F2-A on the drawing, were labeled 821-F1-A and B21-F2-A in the cabinet.
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| The Integrated Primary Containment Leak Rate Test (IPCLRT) procedure, PT-20.5, Rev.13, section VI, Initial Conditions, step DD., requires
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| "high drywell pressure instruments defeated by completing Section ' A of Appendix I.
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| Appendix I, step A.1.e, states, " Pull the fuses in -
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| i Figure.1-1 and place under clearance."
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| Figure I-1 requires that fuses E11-F1-A and E11-F2-A be pulled for instrument E11-PT-N011A.
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| i Contrary to. the. above, on May 18, 1987, fuses E11-F1-A and E11-F2-A
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| for N011A were not pulled prior to the start of the -IPCLRT.
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|
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| This failure to. follow procedure is a violation of TS 6.8.1.c',
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| which requires that written procedures be established and implemented covering. surveillance and test activities of safety-related equipment, i.e.,
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| the IPCLRT.
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|
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| This is a Violation:
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| Failure to Properly: Implement Surveillance Procedure (325/87-13-02).
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| b.
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| On May 8, the inspector observed that the anchor plate associated with an adjustable rigid strut (support No. E41-3PG62), on the Unit 1 HPCI injection line had one side of the anchor plate not in contact l
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| with the concrete ceiling.
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|
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| It appeared that both anchor bolts had been partially pulled out of.the concrete. Work request 87-AQBL1 has been issued to repair the condition prior to the unit refueling startup. The support was installed in 1985 as part of Plant Modifi-cation PM-84-381. This modification moved the inboard HPCI injection valve'into the main steam line valve pit. The licensee has concluded i
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| i that the as found condition did not render the support inoperable.
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| J An ' evaluation is being conducted to determine the probable cause of i
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| the as found condition.
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| The inspector will review this evaluation I
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| when completed, c.
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| On May 21, the inspector observed the following items on Unit 1.
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| The drywell-suppression chamber vacuum breaker X18F had its backup close j
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| position indication switch, LS-4, inoperable. The suppression spray
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| header support brackets had E five nuts with less than' full thread l
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| engagement and eight additional nuts were not in contact with the bracket top. ' These items were corrected under work requests 87-ARDF1 and 87-ARDC1, respectively, bne violation and no deviations were identified.
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| 7.
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| ESF System Walkdown (71710)
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| f During the report period, the inspector performed an inspection of the i
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| accessible components of the Unit 2 Automatic Depressurization System (ADS).
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| This verification included the following items:
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| Initiation and permissive instrumentation are valved into service.
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| Backup reactor building air compressors (air supply to ADS valves)
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| are energized.and operable.
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| Backup nitrogen supply system is pressurized and operable.
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| Both temperature and acoustical monitors of the ADS valve discharge
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| . piping are operable.
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| .The ADS actuation logic.3 energized.
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| ADS override switches are in the NORM position.
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| j No. problem was 'found which would render the system inoperable. Two items were.noted.
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| Sporadically during the month, point 10, associated with Safety Relief I
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| Valve (SRV) B21-F013K, of the tail pipe temperature chart recorder,
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| '!
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| 2-B21-TR-R614, has been fluctuating around the alarm setpoint. The normal value for this point during this cycle 'has been right below-the alarm point.
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|
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| However, with the onset of warmer weather, the corresponding increase in the average drywell temperature has resulted in this point's
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| !
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| normal operating band to sometimes overlap with the alarm setpoint. This
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| results' in an erroneous annunciation on the main control panel of a leaking SRV. Sometimes this condition exists for several continuous hours before it clears. Because this could potentially mask a real problem with another SRV, the licensee is in the process of evaluating the feasibility of defeating this input when it is in the alarm state.
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|
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| i Also, on May 15, the inspector observed that the 2-B21-TR-R614 chart recorder was printing scattered points versus its usual slightly wavy lines; however, the wavy lines were still discernible.
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| The licensee declared the recorder inoperable and repaired it under work request 87-AQIF1. The chart had been initialed by two different operators while it was printing in'this erratic pattern before the inspector observed the j
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| condition.
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| One of these signatures was right after the problem had
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| 1
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| 4.
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| sta rted.'
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| Because there was an abrupt change in the recorder's perfor-mance, which was clearly visible on the chart, the inspector believes that this condition should have been detected by this. operator.
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| This concern was discussed with operations management.
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|
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| At the time, the primary o
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| indication of SRV positions, the acoustic monitoring system, 'was: fully j
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| operational and TS required minimum number of channels was met during the l
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| time the recorder was malfunctioning.
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|
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| No violations or deviations we e identified.
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|
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| 8.
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|
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| Unit 2 Forced Outages'(93702)
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| 6.
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|
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| Vacuum Leak in Main Condenser y
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| The ' licensee rapidly removed the generator off the grid due to condenser ' vacuum problems.
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|
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| On May 22, 1987, at 4:40 p.m.,
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| vacuum started decreasing and off gas flow increased.
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|
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| Power reduction was started at,4:45 p.m. and the main turbine was manually tripped at 5:16 p.m.
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|
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| The licensee found the vacuum. leak:
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| the low pressure turbine bearing dirty oil drain header that passes through the
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|
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| !
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| condenser was' cracked.
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| The pipe was plugged and capped.
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|
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| The generator _ was synchronized to the grid on May 23,.1987 at 4:16 p.m.
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|
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| Main generator reverse power.resulted in an' anticipatory start of all four emergency Diesel Generators (DG). The DGs were not required to tie onto'the' emergency buses because offsite power was available.
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|
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| b.
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|
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| Water in Off gas Line On May.27, 1987, Unit 2 experienced a forced outage due to decreasing vacuum. caused' by water in. the off gas header.
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| At 3:55 a.m.,
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| condenser vacuum took a 2 to'3 inches sudden decrease. The operators reduced power by running back the recirculation pumps to minimum
| |
| !
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| speed and inserting control rods'. ' An auxiliary ' operator found the.
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|
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| .l off gas loop ' seal reservoir fil1 valve, 2-0G-SV-4906, wide open,
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| a thereby allowing demineralized water to flow into the off gas header.
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|
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| He manually. isolated the fill line. At 4:20 a.m., the turbine was manually tripped. Reactor power was maintained at approximately 20%
| |
| until the off gas header was drained.
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|
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| The turbine generator was i
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| synchronized back to the grid at 12:02 p.m.
| |
|
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| The event was attributed to an inadequate surveillance procedure and equipment failure.
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|
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| The problem occurred during performance of PT-4.1.8, Off gas System
| |
| !
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| Automatic Isolation Operability Check,.on Unit 1, as required by TS
| |
| ,
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| 4.3.5.9.
| |
|
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| The procedure steps caused the Unit 2 off gas loop seal reservoir tank drain valve, 2-0G-SV-4907, to open.
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|
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| The procedure
| |
| !
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| should have resulted in the Unit i valve, 1-0G-SV-4907, to open. The opening of the Unit 2 drain and resulting level loss, caused the fill
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| ,
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| valve, 2-0G-SV-4906, to open.
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|
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| This valve then stuck in the open j
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| position allowing demineralized water to back up into the 30 minute j
| |
| l | | l |
| | | _ _ _ _ _ _ _ _ _ _ |
| ,
| | _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ |
| | | . _ - |
| L..
| | - _ _ - |
| . -..
| | . _ - |
| l t-
| | .__ |
| '
| | _ - -_ |
| .r
| | ' Iso t -. |
| ,.
| | .-- |
| off gas holdup _line. This partially restricted off gas flow, thereby causing the decreasing vacuum. On May 28, 1987, a temporary revision was Jissued to 'PT-4.l.8 to correct its errors and provide additional
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| | |
| clarification The temporary revision added steps to open the Unit 1 l
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| valve.
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| Specifically, the. steps require lifting of wire LY6 from terminal 58 and jumping terminal'56 to 57 in panel XQ9. Apparently, when the Unit 1 and 2 procedures, PT-4.1.8-1 and-2, respectively, were combined into one PT-4.18 (revision 24 dated February 2,1987),
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| these steps were deleted.
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| The inspectors need to review the licensee's procedure development i
| |
| process in; general and how the PT-4.1.8 revision was done in parti-cular.
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| Pending the inspectors'- ' review, this will remain an Unresolved Item:
| |
| PT-4.1.8, Off gas Automatic Isolation Operability j
| |
| Check Procedure Inadequate (325/87-13-03).
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| i No violations or deviations were identified.
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| l i
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| }} | | }} |
