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| {{Adams | | {{Adams |
| | number = ML20215E655 | | | number = ML20210A794 |
| | issue date = 12/09/1986 | | | issue date = 01/30/1987 |
| | title = Insp Repts 50-413/86-47 & 50-414/86-50 on 861026-1125. Violation Noted:Failure to Provide Adequate Procedure & to Perform Safety Evaluation Re Low Level Vibrational Testing | | | title = Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Repts 50-413/86-47 & 50-414/86-50 |
| | author name = Lesser M, Peebles T, Skinner P, Vandorn P | | | 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 = Tucker H |
| | addressee affiliation = | | | addressee affiliation = DUKE POWER CO. |
| | docket = 05000413, 05000414 | | | docket = 05000413, 05000414 |
| | license number = | | | license number = |
| | contact person = | | | contact person = |
| | document report number = 50-413-86-47, 50-414-86-50, NUDOCS 8612230053 | | | document report number = NUDOCS 8702090015 |
| | package number = ML20215E648 | | | title reference date = 01-09-1987 |
| | 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 = 15 | | | page count = 1 |
| }} | | }} |
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| J UNITE 3 STATES
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| / km RCg*o ' NUCLEAR REGULATORY COMMISSION
| | JAN 3 01987 Duke Power Company LATTN: Mr. H. B. Tucker, Vice President Nuclear Production Department 422 South Church Street Charlotte, NC 28242 Gentlemen: |
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| | SUBJECT: REPORT NOS. 50-413/86-47 AND 50-414/86-50 Thank you for your response of January 9,1987, to our Notice of Violation, issued on December 11, 1986, concerning activities conducted at your Catawba facilit We have evaluated your response and found that it meets the requirements of 10 CFR 2.201. We will examine the implementation of your corrective actions during future inspection We appreciate your cooperation in this matte |
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| REGION 18 g g 101 MARIETTA STREET. * * ATLANTA, GEORGI A 30323
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| Report Nos.: 50-413/86-47 and 50-414/86-50 Licensee: Duke Power. Company 422. South Church Street
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| Charlotte, NC 28242 Docket Nos.: 50-413 and 50-414 License Nos.: NPF-35 and NPF-52 Facility Name: Catawba 1 and 2 Inspection Conducted: October 26 - November 25, 1986 Inspectors: [, /2*/-5 P. K. VanDoorn Date Signed YY P. Hf Skinner
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| Date Signed Yk Y M. S. Le'sser
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| }A*f-ET Date Signed Approved by: Y N -[/d T. A. Pe bles, Sectio ~n Chief
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| /) -/-6 Date Signed
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| Reactor Projects Branch 2 Division of Reactor Projects SUMMARY Scope: This routine, unannounced inspection was conducted on site inspecting in the areas of review of plant operations; surveillance observation; maintenance observation; review of licensee nonroutine event reports and construction deficiency reports; followup of previously identified items; plant start-up from refueling; follow Jp of part 21 reports; review of calibration program; and maintenance program implementatio Results: Of the nine (9) areas inspected, one apparent violation was identified (Failure to provide an adequate procedure and failure to perform a safety evaluation, paragraph 4.b.).
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| 8612230053 861211 PDR ADOCK 05000413 '
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| O PDR
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| | Sincerely, ORIGINAL StGNED BY |
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| | \" >~ $ f. '. PTWNLSE j Luis A. Reyes, Acting Director Division of Reactor Projects cc: t)fW.Hampton,StationManager bec:GNRC Resident Inspector ( W N. Jabbour, NRR Document Control Desk State of South Carolina |
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| | $ t rig / , RI RII CBurger:dm TPeebles VBrownlee 01/A9/87 01/Af/87 Olp/87 |
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| REPORT DETAILS Persons Contacted Licensee Employees J. W. Hampton, Station Manager E. M. Couch, Construction Maintenance Central Manager H. B. Barron, Operations Superintendent W. H. Bradley, QA Surveillance A. S. Bhatnager, Performance Engineer T. B. Bright, Construction Engineer Manager S. Brown, Reactor Engineer B. F. Caldwell, Station Services Superintendent
| | 8702090015 870130 PDR G ADOCK 05000413 ( |
| *J. W. Cox, Superintendent, Technical Services T. E. Crawford, Superintendent of Integrated Scheduling C. S. Gregory, I&E Support Engineer C. L. Hartzell, Compliance Engineer
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| *D. E. Kinard, Mechanical Maintenance Engineer J. Knuti, Operating Engineer P. G. LeRoy, Licensing Engineer
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| *F. N. Mack, Jr., Project Services Engineer W. W. McCollough, Mechanical Maintenance Supervisor C. E. Muse, Operating Engineer
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| *F. P. Schiffley, II, Licensing Engineer
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| *G. T. Smith, Maintenance Superintendent
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| *J. Stackley, I&E Engineer D. Tower, Operating Engineer J. W. Willis, Senior QA Engineer, Operations Other licensee employees contacted included technicians, operators, mechanics, security force members, and office personne * Attended exit interview.
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| ! Exit Interview The inspection scope and findings were summarized on November 25, 1986, with those persons indicated in paragraph 1 above. The inspector described the areas inspected and discussed in detail the inspection findings. The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspectio The following new items were discussed with licensee personnel: C Violation 413/86-47-01: Failure to provide an adequate procedure to conduct a special test on a diesel generator load sequencing panel - see paragraph 4.b; and, Failure to perform a safety evaluation of the special test conducted on a diesel generator load sequencing panel as required by 10 CFR 50.59 (b) - see paragraph _ __ _ __ _ _ _ _ . _ _ ,
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| Inspector Followup Item 413/86-47-03: Review of licensee long term corrective action regarding valve positioning and locking -
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| see paragraph Inspector Followup Item 413/86-47-04: Review of_ licensee action relative to improper torque value - see paragraph ;
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| Inspector Followup Item 413/86-47-05: Review of licensee action for diesel generator visicorder calibration methods - see paragraph 1 The licensee has stated that a portion of violation 413/86-47-01 will be denied. Licensee Action on Previous Enforcement Matters (Units 1 & 2) (92701)
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| (92702)
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| (CLOSED) Unresolved Item 414/86-18-02: Verification of removal of or evaluation of equipment remaining from Unit 2 construction. The inspector verified that the licensee had removed or appropriately evaluated flush piping which had been left in turned over areas. In addition, tours of various areas have been conducted to verify that this is not a significant proble No violations or deviations were identified. Plant Operations Review (Units 1 & 2) (71707 and 71710) The inspectors reviewed plant operations throughout the reporting period to verify conformance with regulatory requirements, Technical Specifications. (TS), and administrative controls. Control room logs, danger tag logs, Technical Specification Action Item Logs, and the removal and restoration logs were routinely reviewed. Shift turnovers were observed to verify that they were conducted in accordance with approved procedure The inspectors verified by observation and interviews, the measures taken to assure physical protection of the facility met current requirements. Areas inspected included the security organization, the establishment and maintenances of gates, doors, and isolation zones in the proper condition, that access control and badging were proper and procedures followe In addition to the areas discussed above, the areas toured were observed for fire prevention and protection activities. These included such things as combustible material control, fire protection systems and materials, and fire protection associated with maintenance activitie _
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| In addition the inspectors reviewed licensee actions relative to license condition 2.C. (9) for Unit 2 which required the licensee to conduct a turbine trip test prior to exceeding 70*4 power to verify tnat PORVs would not be challenged when the anticipatory trip bypass is in effect. The licensee successfully conducted the test which was witnessed by NRC (see Report No. 50-414/86-30). Therefore, the inspector considers that this license condition has been satisfie On November 13, 1986, NRC:RII, NRC:NRR and licensee management held discussions relative to operability of the Auxiliary Feedwater System turbine-driven pumps. Licensee testing while in Mode 3, had resulted in measurement of flows slightly less than requirements. The licensee had previously requested a change to TS 4.7.1.2.1.a.2 to raise the RPM ,
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| of the pumps which had been approved by the vendor. The licensee would have exceeded the 72-hour action statement of TS 3.7.1.2.a. for both Units 1 and 2 at 1845, November 13, 1986, and 0306, November 14, 1986, respectively. The licensee submitted a letter dated November 13, 1986, ,
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| requesting an emergency TS change to extend the action statement time l limit an additional 72-hours and to raise the pump RPM to up to 380 l The NRC gave verbal approval to this request and further granted the i RPM change by letter (Novak to Tucker) dated November 13, 1986. The '
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| NRC issued the licensee amendment for this request by letter (Jabbour to Tucker) dated November 20, 198 Plant events which occurred during the reporting period were reviewed -
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| and discussed with plant personnel. These were as follows:
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| Main Feedwater System (CF) isolation due to hi-hi level on steam generator (S/G) D, Unit 2 on 11/1/86 Containment Air Release and Addition System (VQ) isolation due to loss of flow to containment gaseous (EMF-39) monitor, Unit 2 on 11/4/86
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| CF System isolation caused by reactor trip signal with lo T-ave during testing due to deficient procedure, Unit 2 on 11/7/86 CF isolation and Auxiliary Feedwater System start caused by reactor trip signal during rod drop testing apparently due to personnel error, Unit 1 on 11/11/86 Safety injection on lo steam line pressure due to a spurious signal on one channel with another channel in test, Unit 2 on 11/12/86 VQ termination due to spurious EMF-39 signal, Unit 2 on 11/18/86
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| Reactor trip from 98% power caused by turbine trip caused by loss of CF pumps after loss of condensate booster pumps. Condensate pumps were lost due to low suction and discharge pressure which was caused by a failed open CF recirculation valve. The logic for this valve was apparently wired backwards due to a design error and this was coupled with a failed pressure transmitter. Unit 2 on 11/20/86 CF pumps tripped caused by low lube oil pressure windmill protection, apparently due to transmitter problem. Unit 2 on 11/22/86 Reactor trip from 2.5% power caused by lo-lo level in S/G CF pump B went to low speed when switched to automatic because of a faulty card in the Main Steam System input circuitry. Operators attempted to regain control manually; however, hi-level (P-14) was reached in S/G C causing a turbine trip, the turbine trip caused a shrink in the S/G's resulting in the 10-10 level in S/G B. Unit 2 on 11/22/86
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| , While conducting a routine plant tour, the resident inspector observed that testing was being performed on the IB diesel generator load sequencer panel. The inspector noted that a procedure was not being used nor had a work request been issued authorizing the work. He questioned the persons conducting the test and was informed that low level vibrational testing was being performed to obtain seismic response data after modifications had been made to the panel. The workers stated that the IB diesel generator was inoperabl The inspector advised the Shift Supervisor (SS) of his finding The SS had no knowledge of the testing and stated that the IB diesel generator was operable. The testing was stopped until adequate control of the test could be gained by the licensee. The licensee later produced a
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| vendor procedure "Stevenson and Associates In-Situ Modal Testing ,
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| Procedure" and stated that it was the reference procedure for the tes It had not, however, been reviewed by station management. This violates Technical Specification 6.8.2 which requires station management approved procedures. The licensee stated that the test was being performed to gather seismic data which would be used for computer modeling which would be used at a later date to verify appropriate seismic response for modified electrical cabinet The licensee also stated that vibration amplitude was well below that analyzed for maximum earthquake condition The inspector did observe the cabinet shaking more during initial setup of equipment than later during data gathering, so one concern is the potential to affect the equipmen An evaluation should have been made prior to testing and when testing was completed the equipment should have been assured to have been left in its normal configuration. This is violation 413/86-47-01: Failure to use written and approved procedures while testing safety-related equipmen _ - _ _ _ - _
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| Additionally, a safety evaluation to verify that the testing did not involve an unreviewed safety question had not been performed. This violates 10CFR 50.59.(b) which requires a written safety evaluation be performed to provide the bases for the determination that a change, test or experiment does not involve an unreviewed safety questio Further investigation revealed other weaknesses in the licensee's program for special tests. The licensee's Administrative Policy Manual for Nuclear Stations (APM) provides guidance for conducting special tests in section 3.2.3, for implementing Temporary Procedures in section 4.3, and for the use of QA approved vendor procedures in section 2.7.1.3.2. There exists no Station Directives to implement this guidance. Consequently, there were no procedures to ensure that special tests on safety related equipment were reviewed by station management and that safety evaluations were performed. In addition, there was no mechanism to require authorization from the SS to conduct the tes Therefore, the test on the load sequencer panel was commenced without SS knowledge and without a determination being made to determine if the equipment should be placed in an inoperable status for conduct of this test. Appropriate personnel were not given the opportunity to review the test for possible quality control (QC)
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| inspection or retest requirements. The licensee has been requested to evaluate their methods for ensuring that strict control is maintained over special tests. The licensee does have adequate programs which do exert control over similar evolutions such as maintenance, surveill-ances and periodic test The additional concern relative to notification of the SS should also be addressed by the licensee in review for corrective actions for the violations. The licensee stated that this would be accomplished as part of the review for required improvements in their program to control special test During valve position and system walkdowns conducted by the inspectors two valves were found which were not properly locked in accordance with procedure requirements. These were containment isolation valve IWE56 and Auxiliary Feedwater System valve ICA39. A previously identified violation remains open in this area (see Report No. 414/86-18). The valves were found in the proper position and, therefore, this problem in minimally technically significant. However, an apparent continuing problem exists relative to locking of valves. Discussions with the licensee indicated that they had recognized the need for long term corrective action in this area and have begun a review of all locked valves in the plan In addition, the licensee recently identified a steam supply valve to the Auxiliary Feedwater System, ISA3 to be mispositioned. The licensee had previously found valve ISA6 misposi- ,
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| tioned. The licensee indicated that further review of valve posi-tioning problems was also being conducted and consideration was being given to providing better guidance relative to system walkdowns after _ -- .-. - _ _
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| major activities on a system such' as might occur during refuelings and modification work. Further review of licensee actions are necessary in this , are This is Inspector Followup Item 413/86-47-03: Review of licensee long term corrective actions regarding valve positioning and lockin Two violations were identified as described abov . SurveillanceObservation(Units 1 & 2) (61726)
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| ' During the inspection period, the inspector verified plant operations were in compliance with various TS requirements. Typical of these requirements were confirmation of compliance with the TS for reactor coolant chemistry, refueling water tank, emergency power systems, safety injection, emergency safeguards systems, control room ventila-tion, and direct current electrical power sources. The inspector verified that surveillance testing was performed in accordance with the approved written procedures, test instrumentation was calibrated, limiting conditions for operation were met, appropriate removal and restoration of the affected equipment was accomplished, test results met requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel, i The following surveillances were observed by the inspector:
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| Reactor Coolant System Leak Test (PT/2/A/4150/01A) | |
| Operational Hydrostatic Test (PT/2/A/4150/01)
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| No violations or deviations were identifie . Maintenance Observations (Units 1& 2) (62703) Station maintenance activities of selected systems and components were observed / reviewed to ascertain that they were conducted in accordance with requirements. The inspector verified licensee conformance to the requirements in the following areas of inspection: the activities were accomplished using approved procedures, and functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities performed were accomplished by qualified personnel; and materials used were properly certified. Work requests were reviewed to determine status of ,
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| outstanding jobs and to assure that priority is assigned to safety-related equipment maintenance which may effect system performance, On November 5, 1986 the inspector attended a management meeting with
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| Duke design engineering to discuss recent problems concerning Rotork valve actuators. The problem was discovered while testing an actuator following motor replacement. The actuator did not exhibit a linear
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| l relation., iip between the five torque switch settings. and the output
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| ' torque as previously assumed by the license Subsequent testing of l other actuators revealed a variety of torque characteristics with wide scatter, resulting in a concern by the licensee on the operability of l all Rotork actuators. Both units were in Mode 5 at the tim Typically each actuator had been certified by either Duke or Rotork at one to three different switch settings and a linear correlation among l all settings was assumed. Switch settings were routinely adjusted in the field without requiring a certification of the torque value at that settin Bench test . certifications recently conducted at Catawba Nuclear Station (CNS) and McGuire Nuclear Station (MNS) have revealed the linear correlation to be a poor assumption. Thus it was concluded that adequate certification did not exist to allow torque switch !
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| setting adjustments. This is discussed further in Information Notice 86-93. CNS identified fif ty-three (53) safety-related valve actuators (27 on Unit 1 and 26 on Unit 2) in which the field settings could not be certified. All other safety-related active actuators were at the 100% torque switch setting. Of the fifty-three (53) valve actuators, thirty (30) were field adjusted to the 100% setting, six (6) were
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| adjusted to limit closed (not torqued) and seventeen (17) were bench tested to determine the required settin Further concerns rose when it was discovered that setting 5 (100%) would not produce rated. torque in all cases. The licensee justified operability of all its actuators set at 100% on the basis that no actuators tested at MNS were found to put out less than required torqu The following maintenance activities were observed:
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| Bench testing of Rotork actuators and field verification of actuator switch settin Unit 1 Component Cooling (KC) heat exchanger cleanin No violations or deviations were identified. Review of Licensee Nonroutine Event Reports (Units 1 & 2) (92700)
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| The below listed Licensee Event Reports (LER) were reviewed to determine if the information provided met NRC requirements. The determination included:
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| adequacy of description, verification of compliance with Technical Specifi-cations and regulatory requirements, corrective action taken, existence of potential generic problems, reporting requirements satisfied, and the relative safety significance of each event. Additional inplant reviews and discussion with plant personnel, as appropriate, were conducted for those reports indicated by an (*). The following LERs are closed: ;
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| LER 413/85-42, R Nuclear Instrumentation System Power Range Reactor Trip
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| *LER 413/86-08, R Auxiliary Feedwater Start Due to Malfunction of Main Feedwater Control Valve
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| *LER 413/86-15, R Diesel Inoperable Due to Incorrect I
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| Specification on Support / Restraint Sketch
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| *LER 413/86-24 Automatic Realignment of Nuclear Service ,
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| Water Due to Personnel Error
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| *LER 413/86-27 Swap of Nuclear Service Water to Standby Pond Due to Defective Procedure
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| *LER 413/86-33 Swapover of Nuclear Service Water to Assured Source Due to Defective Procedure
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| *LER 413/86-43 Auxiliary Feedwater Actuation Signal Due to A Condenser Outlet Control Valve Failure LER 413/86-45 Purge System Trip Due to Conservative Radiation Monitor Setpoint LER 413/86-46 Auxiliary Feedwater Pump Auto-Start While Draining Steam Generators LER 413/86-47 Safety Injection Due to A Personnel Error LER 413/86-48 Three Containment Purge System Isolations Due to Conservative Trip Setpoints
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| *LER 413/86-49 Swapover of Nuclear Service Water to Assured Source Due to Perscnnel Error
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| *LER 413/86-50 Fuel Pool Ventilation System Inoperable Due to Blocked Open Doors LER 413/86-51 Emergency Hatch Not Inspected Due to Personnel Error LER 413/86-52 Missed Fire Watches Due to Personnel Error
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| *LER 413/86-53 Swapover of Nuclear Service Water to Assured Source Due to Personnel Error LER 413/86-54 Two Inadvertent Containment Purge Trips Due to Unknown Cause
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| *LER 414/86-20 Feedwater Isolations Due to Malfunction of Feedwater Control Valves
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| *LER 414/86-32, Rv. 1 Feedwater Isolation Due to Leaking Feedwater Control Valves
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| *LER 414/86-37, Rv. 1 Containment Air Release Termination Due to A Defective Procedure
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| *LER 414/86-39 Three Containment Isolations Due to Conservative Trip Setpoints
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| *LER 414/86-40, Rv. 1 Isotopic Analysis Missed Due to Personnel Error
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| *LER 414/86-41, Rv. 2 Safety Injection Due to Malfunction of Control Valve
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| *LER 414/86-42 Annulus Sprinkle Header Isolation Valve Position not Verified Due to Defective Procedure
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| *LER 414/86-43 Inoperable Fire Barrier Due to Personnel Error
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| No violations or deviations were identifie . Previously Identified Inspector Findings (92701) (CLOSED) Inspector Followup Item 413/84-63-01: Upgrading of Incore Thermocouples and Subcooling Monitor System. The licensee was required by License Condition (15) to upgrade the existing subcooling margin monitor and existing backup display prior to startup following the first refueling outage. This upgrade was detailed in Nuclear Station Modification (NSM) #CN-10210. NSM CN10210 was completed November 4, 198 (CLOSED) Inspector Followup Item 413/86-15-01, 414/86-16-01: Verify that licensee is obtaining updated information relative to Woodward Governors. The inspector verified that appropriate information is being received and manuals are being update . Maintenance Program Implementation (Units 1 & 2) (62700) The inspector reviewed the implementation of the maintenance program to determine if it is meeting regulatory requirements, to determine the effectiveness of the program on important plant equipment, and review the maintenance staff's activities in this are This review consisted, in part, of a review of equipment operating history for specific component failures leading to plant shutdowns and recurring safety related equipment failure. Attributes covered iW'\his review i
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| included evaluation of the cause of the failure and corrective action taken to reduce the probability of recurrence, that procedures were adequate to perform the maintenance and that these procedures met manufacturer's recommendations, and that vendor maintenance recommen-dations were adequately incorporated into maintenance procedures. In addition, various records associated with the maintenance activities performed were reviewed to assure approved procedures were used when l
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| required, limiting conditions for operations were met while the work was being performed, procedures were adequate for the work being performed, functional _ testing was performed as required, measuring and test equipment used was calibrated and controlled, personnel were trained to perform the maintenance as necessary, and miscellaneous other required controls were me The maintenance master schedule was also reviewed and the station lubrication manual as they pertain to the work request and procedures identified below. Procedures were reviewed to assure vendor technical manual information was incorporated, the activity was in sufficient detail to perform work, inspection and hold points identified, radiological and cleanliness controls addressed where required, control of lifted leads and jumpers, and cor. trol of materials and replacement -
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| materials were maintaine ~
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| The following documents were reviewed as part of this process:
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| WORK REQUESTS
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| 14988 OPS 759 PRF '
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| 476 IFI 3678 IAE . . -
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| 4518 PRF 8381 OPS et; 16474 OPS 10279 OPS 18902 OPS 2593 PRF 18933 OPS 4381 PRF 8621 OPS 31970 OPS 1699 PRF _ 2803 SWR 4760 SWR 5253 PRF 3463 MNT 21790 OPS Procedures associated with work request listed above:
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| PM/IG-106 Centrifugal Charging Pump Motor Preventive Maintenance MP/0/A/7650/02 Lubrication of Safety Related Equipment MP/0/A/7600/73 Rockwell 3" Gate Valve Corrective '
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| Maintenance MP/0/A/7600/65 Kerotest 1 1/2, 2 Inch Bellville "Y"-Type Globe Valve Corrective Maintenance MP/1/A/7650/50 Mechanical Maintenance Work on Unstamped Systems IP/0/A/3820/01 Limitorque Valve Operator Corrective Maintenance
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| - 11 IP/0/A/3820/20 Namco IE Limit Switch Installation Requirements IP/0/A/3890/01 Controlling Procedure for Troubleshooting and Corrective Maintenance IP/0/A/3890/03 Installation of Instrument Line Fittings and Tubing IP/2/A/3173/05 Containment Air Return and Hydrogen Skimmer Systems (VX) | |
| MP/0/A/7650/01 Flange Gasket Removal and Replacement IP/0/A/3820/08 Maintenance Instructions for Normally Closed (V52600-555-2), Normally Open (V52600-554-1), and Normally Closed (V52600-5551-1-1) Valcor Solenoid Valves IP/0/A/3820/08D Series 526 Valcor Valve Solenoid Removal /
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| Installation MP/0/A/7400/55 Diesel Flow Control Valve (Valcor)
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| Removal-Repair-Installation IP/0/A/3820/08B Series 526 Valcor Valve Cover Removal / Installation IP/0/A/3820/08A Procedure for Testing, Adjusting and Replacement of Valve Position Indication Magnetic Reed Switches on Valcor 2 Way Solenoid Type Valves IP/0/A/3820/08C Procedure for Testing, Adjusting and Replacement of Valve Position Indication Magnetic Reed Switches on Valcor 2 Way Solenoid Type Valves Without Remote Indication MP/0/A/7600/47 Fisher-Diaphragm and Manual Actuated Control Valves, Types DBQ, DBQA and DBQNS and Type 1008 Manual Operator Corrective Maintenanc ,
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| The inspector verified calibration equipment used during portions of work requests listed above. This equipment is as follows:
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| Fluke 8600 S/N 18965
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| 18967
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| 19045 Digisnap DSA 1000 S/N 18929
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| 18931 Torque Wrench CMMNT S/N 18054 -
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| 18115
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| " "- 18003 No violations or deviations were identified, During review of procedure MP/0/A/7600/65 it was noted that the yoke to body torque value (2600 ft-lbs) was in accordance with the valve manual requirements for stainless steel valves only. The procedure applies to carbon steel valves also which requires, by the valve manual, a smaller torque (1200 ft-lbs). The licensee indicated that this problem had been evaluated via a Nonconforming Item Report and that procedures would be changed to reflect the proper torque for carbon steel valve Further review is necessary of the licensee evaluation and procedure change This is Inspector Followup Item 413/86-47-04: Review of licensee actions relative to improper torque valu . Followup of Construction Deficiency Reports 50.55(e) (92700)
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| (CLOSED) CDR 413/84-01: Modification of Cross-Over Isolation Valves. This deficiency was initially' discussed in a letter to NRC Region II dated February 13, 1984. Subsequent correspondence was provided in letters' dated April 20, July 26, October 30, and December 11, 1984, and February 28, 198 The action detailed in these documents have been implemented and reviewed in detail by the resident inspector and found to be acceptable. This item is close . Calibration (Unit 1) (56700) During this inspection period, the inspector reviewed certain portions of the licensee's calibration program to ascertain whether plant instrumentation calibration is in conformance with licensee require-ments. The inspector reviewed documentation and records to verify licensee conformance to calibration requirements for instrumentation in the Reactor Protection, Erigineered Safety Features and selected safety-related system The following areas were verified: instru-mentation calibration and channel functional tests were accomplished in ;
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| accordance with the specified frequency in Technical Specifications I (TC); test documentation was complete; acceptance criteria was met; and approved test procedures were used, l
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| 13 A blackout: test of IB diesel generator was conducted on October 15, 1986. Review of the -test data by the licensee revealed the voltage visicorder to have read 475 volts high. Other voltmeters indicated normal voltage during the test. The inconsistency -was traced to an -
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| improperly calibrated visicorder. The entire loop is calibrated by two independent groups. Transmission department calibrates the voltage transducer (input to visicorder) every 18 months while IAE department calibrates the visicorder every 90 days. In this particular case, the transducer was calibrated a few days - af ter IAE had calibrated the visicorder. Since changes were made during the transducer calibration, the visicorder calibration was nullified. The licensee verified the 475 volt error to be consistent with the calibration adjustment of the transducer. The inspector requested the licensee to verify that the same problem did not exist on the three other diesel generators'
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| instrumentation and to review other plant instrumentation for similar problems. Most plant equipment loop calibration is performed by one group hence this problem appears to be isolated. The licensee has discussed possible solutions in either the scheduling of the two calibrations or calibration method changes to solve this proble (Documented under Problem Investigation Report 1-C86-0092.) This is Inspector Followup Item 413/86-47-05: Improvement in diesel generator visicorder calibration method The following calibration procedures were reviewed for technical content:
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| IP/0/8/3225/05 ITE System Auxiliary Speed Sensor Unit Calibration IP/1/A/3222/15C Reactor Coolant Flow Channel 3, loop calibration IP/1/B/3600/01B Calibration of Instrumentation in the Diesel Generator 1B Fuel Oil System IP/1/A/3200/03 Reactor Protection / Engineering Safeguards Features Response Time Testing No violations or deviations were identified.
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| 12. Plant Startup From Refueling (Unit 1) (71711)
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| The inspectors observed testing, conducted system walkdowns, observed operations, reviewed licensee documentation and attended licensee meetings to verify that systems disturbed or tested during refueling were returned to an operable status before plant startup and to determine whether plant startup, heatup, approach to criticality, and core physics tests were conducted in accordance with approved procedures. The inspectors verified that appropriate administrative requirements were in place and followed, Technical Specifications were complied with, and plant startup was orderly and controlled with appropriate rod withdrawal criteria followe .-. .- --- . . --- - - - -
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| The inspectors conducted walkdowns of the Auxiliary Feedwater System and the Component Cooling Water System. 'The inspectors witnessed the following startup tests:
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| PT/1/A/4150/19) I/M Approach to Criticality PT/1/A/4150/12A Isothermal Temperature Coefficient of Reactivity Measurement PT/1/A/4150/11A Control Rod Worth Measurement by Boration/ Dilution PT/1/A/4150/11B Control Rod Measurement by Red Swap No violations or deviations were identified.
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| 13. Followup of Part 21'(Units 1 & 2) (92700)
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| (CLOSED) P21 85-06: Defective acoustic valve flow monitor. The inspector verified that the licensee had received and reviewed this report and had taken appropriate corrective actio No violations or deviations were identified.
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