IR 05000272/1986024
| ML18092B321 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 10/08/1986 |
| From: | Norrholm L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18092B320 | List: |
| References | |
| 50-272-86-24, 50-311-86-24, NUDOCS 8610210413 | |
| Download: ML18092B321 (13) | |
Text
I
Report No Docket Nos.-
License No Licensee:
U. S. NUCLEAR REGULATORY COMMISSION REGION I*
50-272/86-24 50-311/86-24 50-272 50-311 DPR-70 DPR-75 050272-860627 050272-860731 050272-860805 050311-860716 050311-860826 Public Service Electric and Gas Company 80 Park Plaza Newark, New Jersey 07101 Facility Name:
Salem Nuclear Generating Station - Units 1 and 2 Inspection At:
Hancocks Bridge, New Jersey Inspection Conducted:
August 19, 1986 - September 30, 1986 Inspectors:
Approved by:
1rir7/sro
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No. 2, DRP Inspection Summary:
Inspections on August 19, 1986 - September 30, 1986 (Combined Report Numbers 50-272/86-24 and 50-311/86-24)
Areas Inspected:
Routine inspections of plant operations including: followup on outstanding inspection items, operational safety verification, maintenance, surveillance, review of special reports, and licensee event followu The inspection involved 121 inspector hours by the resident NRC inspector Results:
During this inspection period two reactor trips with accompanying electrical perturbations were experienced by Unit These events are de-scribed in Inspection Reports 50-311/86-26 and 86-2 An identified violation relating to fire protection is described in Section 7 of this repor..
DETAILS Persons Contacted Within this report period, interviews and discussions were conducted with members of licensee management and staff as necessary to support inspec-tion activit.
Followup on Outstanding Inspection Items (Closed) Inspector Follow Items (272/86-03-01 and 311/86-03-01).
On com-pletion of the analyses of water samples by the licensee and Brookhaven National Laboratory, a statistical evaluation was to be mad The analys-es were completed and an evaluation was performe During combined inspection 272/86-03 and 311/86-03, the steam generator, refueling water storage tank, and a well were sampled for analysi Duplicate samples were sent to the Brookhaven National Laboratory (BNL) for independent verification of analysis, with the follow results; Salem - Split Samples Ammonia (ppb)
Hydrazine (ppb)
Boron.(ppm)
Iron (ppm)
Silica (ppm)
NRC (BNL)
145 200 2120
.025 1 Salem 145 270 2100.2 A statistical comparison of the results was not made because the uncer-tainties were not availabl (Closed) Unresolved Item (311/86-26-05).
See Section 4 of this repor.
Operational Safety Verification 3.1 Documents Reviewed Selected Operators' Logs Senior Shift Supervisor's (SSS) Log Jumper Log Radioactive Waste Release Permits (liquid & gaseous)
Selected Radiation Work Permits (RWP)
Selected Chemistry Logs Selected Tagouts Health Physics Watch Log 3.2 The inspector conducted routine entrie~ into the protected areas of the plants, including the control rooms, Auxiliary Building, fuel buildings, and containments (when access is possible).
During the
inspection activities, discussions were held with operators, techni-cians (HP & I&C), mechanics, supervisors, and plant managemen The purpose of the inspection was to affirm the licensee 1 s commitments and compliance with 10 CFR, Technical Specifications, and Administrative Procedure (1)
On a daily basis, particular attention was directed to the fol-lowing areas:
Instrumentation and recorder traces for abnormalities; Adherence to LC0 1 s directly observable from the control room; Proper control room shift manning and access control; Verification of the status of control room annunciators that are in alarm; Proper use of procedures; Review of logs to obtain plant conditions; and, Verification of surveillance testing for timely completio (2)
On a weekly basis, the inspector confirmed the operability of selected ESF trains by:
Verifying that accessible valves in the flow path were in the correct positions; Verifying that power supplies and breakers were i.n the cor-rect positions; Verifying that de-energized portions of these systems were de-energized as identified by Technical Specifications; Visually inspecting major components for leakage, lubrica-tion, vibration, cooling water supply, and general operat-ing conditions; and, Visually inspecting instrumentation, where possible, for proper operabilit (3)
On a biweekly basis, the inspector:
Verified the correct application of a tagout to a safety-related system; Observed a shift turnover;
Reviewed the sampling program including the liquid and gas-eous effluents; Verified that radiation protection and controls were prop-erly established; Verified that the physical security plan was being implemented; Reviewed licensee-identified problem areas; and, Verified selected portions of containment isolation lineu.3 Inspector Comments/Findings:
The inspector selected phases of the units* operation to determine compliance with the NRC 1s regulation The inspector determined that the areas inspected and the licensee's actions did not constitute a health and safety hazard to the public or plant personne The fol-lowing are noteworthy areas the inspector researched in depth: Unit 1 Unit 1 maintained 100% power from the start of the report period until September 2, 1986, when the power level was reduced to approximately 83% due to the imposition of limited loading requirements on the station power transformers (SPT) (removal of balance of plant loads, i.e. condensate, circulating water, and heater drain pumps) as a result of the Unit 2 reactor trip, safety injection, false loss of offsite power event of August 26, 198 (See Section 3.3. and NRC Special Report 50-311/86-26 for details.) On September 17, 1986 at 4:21 p.m., unit power was increased to 98%.
Since Unit 2 was in cold shutdown (Mode 5) and Unit 2 loads on the No. 21 SPT were minimal, the N condensate pump for Unit 1 could be reenergized from the previously installed temporary crosstie to No. 21 SPT, keeping within the SPT limited loading requirements referred to abov At 9:13 p.m. on September 17, 1986, Unit 1 was taken off line and a controlled shutdown was commenced to repair a steam leak on high pressure turbine cold reheat piping to the moisture separator reheater (MSR) and reheat steam sys-te The cause of the leak was steam erosion on the sec-tion of 20 11 bleed steam piping to the No. 15 Feedwater heaters that connects from the 44 11 MSR pipin The unit was maintained in hot shutdown (Mode 3) during the repair Replacement of the affected sections of piping on both the
- east and west bleed steam p1p1ng were completed on September 21, 198 The unit returned to power on September 24, 1986 at 10:17.
Unit 2 Unit 2 began the report period at 100% powe On August 26, 1986 at 8:28 a.m., an Unusual Event (UE) was declared due to a turbine/reactor trip and safety injection (SI) resulting from troubleshooting No. 22 Steam Generator level channe The event was compounded by a false loss of offsite power (11blackout 11 ) signal which caused the vital busses to strip their loads and the emergency diesel gener-ators to load on the safeguards sequenc The SI and UE were terminated at 8:35 During the August 26 event, a service -water p1p1ng leak was identified on the return line from No. 23 Containment Fan Coil Unit (CFCU) motor coole This piping along with sim-ilar piping on Nos. 21, 22, and 25 CFCUs was replace Following the August 26 event, as a result of commitments made at an NRC/licensee meeting on August 30, 1986, and by a licensee Justification for Operation for restart of Unit 2 and continued operation of Unit 1 dated August 31, 1986, NRC permitted restart of Unit 2 (and continued operation of Unit 1).
The commitments consisted of limiting loading on the station power transformers, powering the group busses from the SPTs, and performing additional analyses on the electrical syste Details of the above events are discussed in NRC Special Report 50-311/86-2 On September l, 1986 at 10:04 p.m., the unit was synchro-nized to the grid and subsequently brought to 75% powe At 6:58 p.m. on September 11, 1986, the unit tripped from 75% power due to deenergization of the 2F and 2G 4160 volt non-vital (group) busses and the resulting loss of Nos. 22 and 23 reactor coolant pump The deenergization of the 2F and 2G group busses resulted from an electrical short on a non-vital 4160/230 volt transformer associated with the 2F bus and a coincident internal failure of the No. 22 SP The unit was placed in cold shutdown (Mode 5) to perform repair A spare 4160/230 volt transformer and SPT were tested and installed.
- Details of this event are discussed in NRC Special Report 50-311/86-2 The unit returned to power on September 28, 1986 at 9:24 p..
Maintenance Observations The resident inspector reviewed the following completed work orders and related documents to verify corrective actions relative to plant equipment problems identified during the Unit 2 reactor trip/safety injection/false loss of offsite power event of August 26, 198 Refer to Appendix C of NRC Special Inspection Report 50-311/86-2 Work Order Number 86-08-26-047-6 (21SGFP)
86-08-26-053-1 (22SGFP)
86-08-26-045-0 (Annunciator)
Procedures/
Related Documents IC-14.1. 001 General Troubleshooting Procedure for Plant Equipment IC-14.1.001 General Troubleshooting Procedure for Plant Equipment M3Z Electrical Equipment Troubleshooting and Repair Description Verified No.~21 SGFP tripped upon 2/3 SG High High Level signals and timed trip response via se-quence of events comput-er using a simultaneous Low Level bistable tri Verified acceptable No. 22 SGFP trip time on 2/3 SG High High Levels using Low Level bistabl Used sequence of events computer to calculate trip time from High High Leve Verified First Out Annunciator Window F-46 fl ashes and resets properl Confirmed thru engineer-ing testing that circuit is satisfactor,.
86-08-26-055-7 (Reactor trip breaker)
(Manual reactor trip)
86-08-26-056-5 (Turbi*ne stop valves/
Remote emergency trip)
86-08-26-057-3 ( 23 AFP)
86-08-26-046-8 (BIT Flow Indication)
IC-14.1. 001 General Trouble-shooting Procedure for Installed Plant Equipment IC-14.1.001 General Trouble-shooting Procedure for Installed Plant Equipment IC-14.1. 001 General Troubleshooting Procedure for Installed Plant Equipment 2PD-2.9.177 Channel Sensor Calibration 2IC-2.10.177 Channel Calibration Procedure Containment Fan Coil Units 86-08-27-024-2 (No. 21)- A28 Department Control (Code Job Package of Code Work S-86-158)
- MllD Pressure Test 86-08-27-031-5 (No. 22)
(S-86-159)
- MllE Mechanical 86-08-26-032-8 (No. 23)
(S-86-156)
Equipment Trouble-shooting and Repair
- NDWP-13 Welding Procedure Specification Verified time response from No. 23 Loop Low Fl ow 2BS434, 2BS435 to reactor trip breaker ope Verified manual reactor trip sequence of events indicatio Verified time response from No. 23 Loop Low Fl ow 2BS434, 2BS435 to turbine stop valves and remote emergency tri Provided signal inputs to verify time response from 2 SG Low Low Leve 1 signals to No. 23 AFP star *
Recalibrated SI charging pump discharge (BIT) flow transmitter 2FT91 Verified calibration to flow indicator 2FI-91 Rep 1 aced 2 11 p1 ping on discharge of motor coo 1 er at 10
stainless steel spoo NOE and hydro-teste.. '
86-08-27032-3 (No. 25)
(S-86-160)
22 SG Level Channel III 86-08-26-002-6
- NDWP-9 Welding Procedure Specification
- Deficiency Reports: SMD-M-86- 0295 (21CFCU)
- 0293 (22CFCU)
- 0290 ( 23C FCU)
- 0294 (25CFCU)
- Inspection Point Checklists
- Weld History Records
- Storeroom Material Issue & Returns
- NOE Reports
- NR-1 Forms
- PD-14.1.110 General Troubleshooting Procedure for Plant Equipment
- 2PD-2.2.044 2LT-528 No. 22 SG Protection III Channel Calibra-tion Check
- 2IC-18.1. 009 SSPS Train B Functional
- 2IC-18.1. 010 SSPS Train A Functional -
Reactor Trip Breaker UV Coil and Auto Shunt Trip
- 2PD-2.6.044 2LT-528 No. 22 SG Level Protection III Channel Functional
- Calibration Data Sheets
- Equipment History Cards Changed comparator 2LC-528 Performed channel calibration and functional test on 2LT-52 Replaced K325 input relays in Trains A and Completed SSPS Trains A and B Functional tests-SA,.
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2B Diesel Generator The inspector reviewed selected maintenance and surveillance work orders for the 2B Diesel Generator pertaining to the event dated August 26, 198 FINDINGS:
During review of the above documents and discussions with licensee person-nel, the inspector noted the following: SGFP tripping problems could not be duplicated during testin Both feed pumps (Nos. 21 and 22) tripped as designed when teste Each pump was tested three time.
The licensee determined that, due to filtering and a time delay to reduce annunciator response from spurious signals (noise spikes), the SI signal that resulted from the momentary (0.15 sec.) ground was too fast for the annunciator circuit to pick u.
No indication of BIT flow during the event was due to the control circuit being powered from B bus which was deenergized as a result of the false 11b-:lackout 11 signal and 118 11 diesel being tagged out for maintenance. The I&C technician was working to approved maintenance procedures (2IC-18.l.009 and PD-14.1.001) at the time of the tri.
Work orders on 2B diesel generator that were to be worked on the day of the event have since been complete The inspector has no further questions at this tim Unresolved Item 311/86-26-05 is considered close No violations were identifie.
Surveillance Observations During this inspection period, the inspector reviewed in-progress surveil-lance testing as well as completed surveillance package The inspector verified that the surveillance tests were performed in accordance with licensee approved procedures and NRC regulation The inspector also verified that the instruments used were within calibration tolerances and that qualified technicians performed the surveillance test The following surveillance tests were reviewed:
Unit 1 lPD-2.2.028 Channel Calibration Check, No. 11 SG Steam Flow Protection Channel I (lFT-512)
- lPD-2.2.038
1,.;.:.
Channel Calibration Check, No. 12 SG Steam Flow Protection Channel I (lFT-522)
The resident inspector reviewed the completed data for No. 11 Flow Channel and witnessed the No. 12 channel calibration check in progress. The SG steam flow portion of the SPDS wiring was being hooked up per OCR lEC-1365 and W.O. 86-05-22-106-Following hookup, a three point check was per-formed to ensure channel calibration was not affected as a result of the w1r1ng proces A full calibration was performed on the No. 12 Flow chan-nel since two of the three check points were found to be out of the accep-tance *criteria by a small amount attributable to normal drif lIC-18.1. 006 (Train A)
lIC-18.1.007 (Train B)
lIC-18.1. 010 (Train A)
lIC-18.1.011 (Train B)
Solid State Protection System (SSPS) Reactor Trip Breaker and Permissive P-4 Test Prior to Startup Solid State Protection System (SSPS) Reactor Trip Breaker and Permissive P-4 Test Prior to Startup SSPS Reactor Trtp Breaker UV Coil and Auto Shunt Trip SSPS Reactor Trip Breaker UV Coil and Auto Shunt Trip The resident inspector observed these tests being performed in preparation for Unit 1 startu The inspector discussed a potential procedure wording concern with the Senior I&C Supervisor who explained the history of and basis for the wordin The inspector had no further question Unit 2 SP(0)4.5.2H Throttling Valve Flow Balance Verification The resident inspector reviewed the completed test data for the high head (No. 22 charging pump) throttling valve test for the 21-24SJ16 valve This surveillance test was conducted to verify that proper flow balance and flow rates had not been altered by the performance of an internal in-spection of the No. 22 charging pum The flow characteristics acceptance criteria as delineated in Technical Specifications were met without posi-tion adjustment of the SJ16 valve No violations were identifie.
Review of Periodic and Special Reports Upon receipt, the inspector reviewed periodic and special report The review included the following:
inclusion of information required by the NRC; test results and/or supporting information consistent with design predictions and performance specifications; planned corrective action for resolution of problems, and reportability and validity of report informa-tio The following periodic reports were reviewed:
Unit 1 Monthly Operating Report - August 1986 Unit 2 Monthly Operating Report - August 1986 In addition, the inspector reviewed the following Special Reports; Special report, Unit 2, 86-6 (August 22, 1986) and supplement to Special report 86-6 (September 19, 1986).
In accordance with Technical Specifications, any fire barrier that has been inoperable for greater than seven days requires a special repor Special report 86-6 deals with the inoperability of certain fire barriers and the supplement addresses the licensee 1 s intentions in dealing with certain fire barriers that are becoming inoperable due to the installation of the Safety Parameter Display System (SPDS).
The resident inspector reviewed the licensee 1s intentions and actions with regard to the inoperability of fire barriers that have been made inoperable because of the pulling of cable, installation of scaffolding, and other restriction The inspector concluded that the licensee is tracking the progress of the impair-ments to fire barriers in a timely manner and that proper compensato-ry measures are being implemented by the licensee (except in the instance described in LER 86-11 n.oted in Section 7 of this report).
The inspector also identified that the licensee has implemented continuous fire watches and will issue an additional supplemental report when the installation of the SPDS has been complete Special Report, Unit 2, 86-8, September 9, 1986 This report was submitted in accordance with special instructions of Bulletin 80-24 and describes a service water leak in containmen This leak was attributed to a dissimilar metal weld in the 2 11 carbon steel motor cooler return to the 10 11 stainless steel fan coil return lin This event is discussed in Inspection Report 311/86-2 Special Report, Unit 2, 86-9, September 23, 1986 This report was submitted in accordance with Technical Specifications 4.8.1.1.4, Valid Test Failure of a Diesel Generato The cause of the failure has been attributed to 118 11 diesel cooling service water throttle valve (22SW42) sticking which resulted in insufficient cool-ing to the diesel and the subsequent high jacket water temperature alar The diesel was tripped to prevent damage to the engin The valve has been stroked several times and the diesel has been operated successfully three times since the valve sticking problem with no further problem The resident inspector could not identify another incidence of this type on any of the diesel generators and considers this an isolated inciden No violations were identifie ~..
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12 Licensee Event Report Followup The inspector reviewed the following LERs to determine that reportability requirements were fulfilled, immediate corrective action was taken, and corrective action to prevent recurrence had been accomplished i~ accor-dance with Technical Specification Unit 1 86-011 Fire Watch Not Continuously Maintained I.A.W. Technical Specifications At 5:00 p.m. on June 27, 1986, shift personnel identified that an inopera-ble fire door (C15-1) had been without a fire watch for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (since 11:00 a.m.).
Licensee investigation results indicated that one fire watch was assigned for both the door and for an impaired fire barrier penetra-tion at the same location. At the conclusion of the work on the penetra-tirin, the fire watch was released, leaving no fire watch for the inoperable doo The fire alarm in the area was subsequently proved oper-able and the fire brigade would have been available to respond to the fire alarm if require Reactor Trips from 70% Power and 36% Power These events are described in Inspection Report 272/86-2 The inspector has reviewed this LER and has determined that the LER accurately describes the even The inspector has no further questions at this tim Failure to Enter Limiting Condition for Operation Action Statement At 9:08 p.m. on July 29, 1986, the licensee failed to provide a fire watch on impaired door C-8-1 (which is a Technical Specification fire door)
because the door had louvers installed in it and it was not identified as a fire doo Subsequent followup by the Site Protection department. iden-tified the door as being a fire door with fusible link louvers designed to close in the event of high temperature At 8:30 a.m. on July 31, 1986 a fire watch was provide The fire alarm in the area was subsequently proved operable and the fire brigade would have been available to respond to the fire alarm if require The events described in 86-011 and 86-017 occurred because of defective procedures that did not clearly delineate the responsibilities between departments or clearly define fire door The licensee took immediate corrective action upon the discoverie These events are a violation (50-272/86-24-01).
As stated in 10 CFR 2, Appendix C.V.A. the NRC will not generally issue a notice of violation for a violation that meets all of the following tests: (1) it was identified by the licensee; (2) it fits in Severity Level IV or V; (3) it was reported, if required; (4) it was or will be corrected, including measures to prevent recurrence, within a reasonable time; and (5) it was not a violation that could reasonably be expected to have been prevented by the licensee's corrective action for a previous violatio The licensee has:
Identified the two incidents and made the proper report Issued Quality Action Requests (internal quality assurance documents that identify degraded conditions).
These requests have been replied to and have delineated corrective action Re-written AP-25 11 Fire Protection Program 11 to clearly define respon-sibilities and fire protection equipment to preclude further confusio The resident inspector has reviewed the licensee 1 s corrective actions with regard to the above incidents and considers the matter close One Violation was identifie Unit 2 86-006 Reactor Trip from 52% Power This event is discussed in Inspection Report 50-311/86-1 The inspector has reviewed this LER and has determined that the LER accurately describes the even The inspector has no further questions on this even Reactor Trip/Safety Injection With Loss of Offsite Power Indication This event is discussed in Special Inspection Report 50-311/86-2 After review of the LER the inspector has determined that the LER accurately describes the even No violations were identifie.
Exit Interview At periodic intervals during the course of the inspection, meetings were held with senior facility management to discuss the inspection scope and finding An exit interview was held with licensee management at the end of the reporting perio The licensee did not identify 2.790 material.