IR 05000324/1982010

IE Insp Repts 50-324/82-10 & 50-325/82-10 on 820119-21. Noncompliance Noted:Failure to Follow Procedures,Failure to Meet Commitments of TMI Action Item I.C.6,failure to Take Corrective Action & Failure to Implement Procedures

ML20054H028
Person / Time
Site:	Brunswick
Issue date:	04/01/1982
From:	Burger C, Garner L, Julian C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20054G982	List: IR 05000324/1982010 ML20054G981 ML20054G994 ML20054H005 ML20054H013
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-1.C.6, TASK-TM 50-324-82-10, 50-325-82-10, NUDOCS 8206220508
Download: ML20054H028 (7)

Text

( l D# UNITED STATES 8" 1, NUCLEAR REGULATORY COMMISSION g s REGION 11 o # 101 MARIETTA ST., N.W., SUITE 3100 o ATLANTA, GEORGIA 30303

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Repo r.t. flos. 50-324/82-10 and 50-325/82-10 Licensee: Carolina Power and Light Company 411 Fayetteville Street Raleigh, flC 27602 Facility flame: Brunswick Docket flos. 50-324 and 50-325 License Nos. DPR-62 and DPR-71 Inspection at Brunswick site near Southport, North Carolina Inspectors: b' i hv L. Garnep, Resident inspector

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Date Signed C-W C.Julian,ProyctInspector y/>/8 u Date Signed

Approved by: #-/- P L

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gC. Burger, Section Chief, Division of Resident Date Signed and Reactor Project Inspection SUltitARY Inspection on January 19-21, 1982 Areas Inspected This special announced inspection involved 45 inspector-hours on site in the area of review of the scram event of January 16, 1982 in which all RHR service water pumps failed to star Results Of the area inspected, four violations were found (Failure to follow procedures-paragraph 5.b; Failure to meet commitments of Till Action Item I.C.6 as ordered by the flRC - paragraph 5.b; Failure to take corrective action - paragraph 7; and Failure to implement procedures - paragraph 7).

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DETAILS Persons Contacted Licensee Employees

• J. A. Jones, CP&L Vice Chairman

• E. E. Utley, CP&L Executive Vice President

• L. W. Eury, CP&L Senior Vice President

• B. J. Furr, Vice President, Nuclear Operations

• C. R. Dietz, Plant General Manager

• R.11 organ, Plant Operations fianager

• R. Knobel, Assistant Operations Manager

• R. Poulk, Regulatory Specialist Other licensee employees contacted included operators, maintenance personnel, security force members and office personne ~

• Attended exit interview of January 21 Exit Interview The inspection scope and findings were summarized on January 21, 1982 with those persons indicated in paragraph 1 above. NRC concerns about the events involving the open breakers and failed instruments without appropriate action were discussed by R. C. Lewis with Mr. Ben Furr, Vice President, Nuclear Operations, on January 19, 198 . Licensee Action on Previous Inspection Findings

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Not inspecte . Unresolved Items l

Unresolved items were not identified during this inspectio . Inoperable Residual Heat Removal Service Water Pumps , Review of the Circumstances of a Scram Event of January 16, 198 On January 16, 1982, the licensee infonned Region II of a reactor scram that occurred on Unit 2 at 4:38 The cause of the reactor scram was a turbine trip due to low condenser vacuum. Electrical maintenance technicians were replacing a pressure sensing instrument on the steam jet air ejector (SJAE) system when they inadvertently grounded a hot lead and caused the SJAE to trip at 4:25 p.m. Condenser vacuum began to decrease and the operators reduced reactor power by reducing recirculation flow and inserting control rods, started a mechanical vacuum pump, and attempted to restart the SJAE. At 4:38 p.m. a turbine trip and reactor scram occurred, followed one minute later by a group l

one isolation which closed the main steam isolation valves. Isolation J

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occurred when the mode switch was switched from "run" to the " shutdown" position by normal procedure. Steam flow switches apparently still indicated greater than 40% flow, which initates a group one isolation with the mode switch not in ru Reactor Core Isolation Cooling (RCIC) was manually started to increase the reactor vessel water level. RCIC exhausts the steam from its turbine to the torus, so the residual heat removal (RHR) system was aligned in the torus cooling mode to control the slowly increasing torus temperature. At approximately 4:40 p.m. it was found that none of the four RHR service water pumps would start, apparently due to low pump suction pressure, as indicated by control room annunciatio The group one isolation was reset at 4:50 p.m. restoring the condenser as a heat sink. At 4:55 p.m. RCIC was secured and a reactor feed pump was started for vessel level control, thus halting the torus heat input. Technical Specification maximum of 120 F on torus temperature was not approache The 2B and 2D RHR service water pumps i.e., the 2B RHR service water subsystem, were restored to operability at 8:58 p.m. The 2A RHR service water subsystem i.e. the 2A and 2C pumps, were restored at 11:54 Conclusions Relating to the Inoperable RHR Service Water System Although the RHR service water (RHRSW) system was not required during this plant transient, never the less, the two redundant RHRSW sub-systems, composed of 4 pumps, would not start when +he operator attempted to place the system in servic The low pump suction pressure is sensed by Barkesdale pressure switches (PS) 2-SW-PS-1175 and 2-SW-PS-1176 for subsystems 2A and 2B respectively. PS 1176 was found to be inoperable due to circuit breaker 19, panel 2B in the cable spreading room being open, thus interrupting power to the low suction protection circuit resulting in an electrical start inhibit of pumps 2B and 2 The circuit breaker was apparently incorrectly left open following a well water flush of the RHRSW piping conducted earlier that day. An entry in a facility log shows that the flush operation was completed at 4:50 a.m. on January 16. The procedure OP-43 Service Water System, Revision 20 approved September 30, 1981, specifies flushing the RHRSW piping with fresh water to remove salt water to prevent corrosion of the pip Step G.3.2 of the procedure specifies that breaker 19 be opened during flushing operations to allow the motor cooler supply solenoid valve to open to permit the motor coolers to be flushed along with the rest of the piping. Circuit 19 supplies power to operate the motor cooler supply valves for pumps 2B and 2D and the valves fail open on a loss of power; therefore, opening the appropriate circuit breaker is a convenient way to open the valves for the flus Step G.3.10 of the procedure specifies that the breakers be reclosed after the flus j

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This step was apparently not done on the morning of January 16 and failure to follow procedure OP-43 resulted in disabling of the RHRSW system and is a violation.(50-324/82-10-01)

Procedure OP-43 does not require that individuals performing step

/ G.3.10 sign them off or otherwise indicate that the step was complete However, in their letter of December 31, 1980, in response to NUREG 0737 Clarification of TMI Action Plan Requirements, CP&L committed to the following in regard to item I.C.6 (Verify correct performance of operating activities).

"When returning equipment to service which has not been under clearance, for example, instruments or hydraulic snubbers removed for surveillance testing, a second person will verify proper system alignment unless functional testing can be performed without compromising plant safety, and can prove that all equipment, valves, and switches involved in the activity are currently aligned. The person performing the verification will have the qualifications necessary for returning the equipment to service or will be a QA inspector".

On July 10, 1981 the NRC issued an order to CP&L which required implementation by January 1,1981 of procedures to verify correct performance of operating activities as specified by NUREG 0737 item I. The inspector observed that, in general, valve lineups have require-ments for double verification, but certain other procedures, such as Periodic Test procedures for surveillance testing of Technical Spec-ification required equipment, do not require double verificatio The licensee committed to review all procedures to determine if the double verification requirement is consistently applie Procedure 0P 43 is an example in which this commitment to double verification was not implemented. Failure to comply with the order issued by the NRC is a violation (50-324, 325/82-10-02).

6. History of Previous Similar Events There were several others events that served as precursors to this event involving circuit breaker No.19 being incorrectly positioned. Three were reported to the NRC in Licensee Event Reports (LERs) as described belo LER 1-80-8: On January 15, 1980 circuit 19 was found deenergized on Unit The licensee concluded that the breaker was left open as a result of a system flush 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> earlie To prevent recurrence, a precaution was placed in procedure OP-43.G to advise the operators that the faRSW pumps are inoperable with the breakers ope LER 2-80-66: On September 5,1980 during normal operation, the corre-sponding breaker for the Unit 2 A loop tripped, rendering the 2A and 2C J

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pumps inoperabl No cause could be found so the breaker was reset and operation continue LER 1-81-95: On December 6,1981, the Unit 1 B loop of RHRSW was being put in service for torus cooling when it was identified that the pumps would not start. The cause was identified as breaker 19 being open. An investigation by the licensee failed to identify any individual who admitted being responsible for opening the breaker. As corrective action, the breaker panels were locked close The inspectors verified that the licensee took the corrective actions s ta ted. However, these three LERs should have served to alert the licensee to potential problems associated with these circuit breaker . Problems with RHR Service Water Pump Suction Switches As detailed below, the switches have experienced numerous repeat failure This documented poor performance record should have been cause for corrective action which would have mitigated this even On January 16,1982, PS 1175 was found to be inoperable due to air accumulation in the oil-filled chemical seal attached to the- pressure swi tch. To prevent chemical corrosion of the Barkesdale pressure switch, the pressure switch is isolated from the brackish water by a diaphragm and a short section of pipe filled with glycerol, which contacts the pressure swi tch. Technicians found that the oil had apparently leaked from the chemical seal, allowing an air bubble to fom which renders the pressure switch inoperable. The chamber was refilled with oil and the switch was recalibrated to restore operability. The inspector reviewed calibration records for these two pressure switches and found that numerous problems had occurred with the switches. Records show that switch 1175 was found to be inoperable on August 4, and October 31, 1981, due to an oil leak from the chemical seal. On November 2,1981 a new switch was installe Switch 1176 was also found inoperable on flarch 20 and September 3,1980, with no logged entry of the exact cause. On July 7,1981 the switch was replaced. On July 11, 1981 the switch was found to again be inoperable due to an oil lea Calibrr ion records demonstrate that these switches as installed and/or maintained are unreliable. The pressure tap comes off piping in the overhead and runs down to the instrument. flaintenance personnel state that this causes a recurent problem of plugging of the instrument lines with debris; however, specific data was not developed to support this contention of failur In the exit interview, in response to the inspector, the General llanager stated that timely action will be taken to improve the reliability of the syste This is an inspector followup item. (50-324, 325/82-10-03)

10 CFR 50, Appendix B, criterion XVI, Corrective Action, requires that when conditions adverse to quality occur, the licensee shall detemine the cause and take corrective action to preclude repetition. The CP&L Corporate J

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Quality Assurance Program Section 15 requires for significant conditions adverse to quality, the cause and corrective actions taken to preclude recurrence be determined, recorded and reported to appropriate levels of management. Failure to take adequate corrective action is a violation of 10 CFR 50, Appendix B, Criterion XVI (50-324, 325/82-10-04).

Procedure it.I. 3-3A34, S.W. - P.S.1175 and 1176 Service Water Pressure Switch, D2T-M8055-L6, revision 0 December 26, 1979 is the specific procedure intended to be used to calibrate pressure switches 1175 and 117 The frequency of calibration is listed as semiannual. Calibration records from September 1979 to date were reviewed for both units and the following discrepancies found. PS-1176 for Unit 2 has no record of calibration between September 3,1980, and July 11, 1981. This interval exceeds the 6 month frequency specified. Although approved on December 26, 1979, procedure MI3-3A34 was apparently not fully implemented. The majority of the completed data sheets are not the one from the approved procedure; rather they come from procedure HI3-3A, Procedure for General Calibration of Pressure Switches. The frequency for f113-3A states "As Required" and this procedure is stated to be used for non Q-list pressure switches. Records show that of the 24 calibrations performed since December 26, 1979, on these pressure switches on both units, only 5 used procedure f113-3A34 and the rest used procedure f113-3 Some used procedure f113-3A but the data was recorded on a data sheet from procedure f113-3 dated February 12, 1976. Calibra tion data recorded was adequate, however, with the exception that data sheet 3-3A34 requires a signature reflecting that pennission was granted by the Shift Foreman to remove the instrument from servic These discrepancies constitute a violation of Technical Specification 6.8.1.c which requires that procedures shall be established, implemented and maintained for surveillance and test activities of safety-related equipmen (50-324,325/82-10-05) Q-List Discrepancies With respect to PS 1175 and 1176, Procedure 1113-3A34 specifies, that this is not a 0-list (safety-related) item and that this procedure is not Technical Specification related. The procedure is in conflict with Volume XI, Book 2, Table I of the Brunswick Plant Operating flanual, Q-list, which correctly identified these two switches as Q list item Volume XI, Book 2 of the Plant Operating fianual contains the plant Q-list in several parts. Table I lists Q-list items on a plant system basis and the service water system section lists SW-PS-1175 and 1176 as Q list item Table I.a of that procedure is a computer output listing of Q list instru-ments titled Brunswick Instrument Calibration Cross Reference, Revision 11 flay 25,1979. Table I.a. does not appear to list the subject switches and thus there is an apparent discrepancy between the two lists. Ta bl e I .a . i s often used by plant personnel for quick reference to determine if a particular instrument is on the Q-list and therefore it should be current complete and accurate. The inspector stated in the exit interview that Table I.a. should be upgraded promptly. The Plant General flanager agreed to J

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review the matter and take the appropriate action. This is an inspector followup item. (50-324,325/82-10-06)

9. Reactor Scram and Group One Isolation i

At the exit interview the inspector expressed concern that a group one isolation closing the main steam isolation valves (t1SIV) very often occurs following a reactor scram for unrelated reasons. The apparent cause is unreliable reset action of the main steam flow switches B21-dPIS-N006A, 7B, 8C and/or 90. Following a reactor scr?m, steam flow and steam pressure gradually lowers as the turbine coas+ down. Normal procedure requires the operator to move the mode switch from the "run" position to '!startup" or

" shutdown" (low pressure isolation is oypassed if the mode switch is not in run). However, with the mode switch in startup or shutdown a group one isolation will occur at steam flows greater than 40% of rated flow. Even though the indicated steam flow is less than 40% when the operator moves the mode switch, the 40'. trip function of these switches apparently resets in a delayed and inconsistent manner. The failure of these switches to reset precisely often causes a group one isolation, when isolation is not required. Operations personnel at Brunswick often experience problems and long delays in getting the group one isolation reset to open the llSIV' The resulting closure of the i1SIV's removes the main condenser as a heat sink and forces the operators to control the reactor vessel water level with High Pressure Coolant Injection (HPCI) or Reactor Core Isolation Cooling (RCIC) and to control pressure by manual use of the Safety Relief Valves (SRV). The SRV's have a history of sticking open, resulting in an uncontrolled pressure decrease. Although the reactor has already scrammed, this represents an unnecessary challenge to other safety systems. The General flanager stated at the exit interview that they are aware of the situation and are taking corrective actio This is an inspector followup ite (50-324,325/82-10-07)

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