IR 05000440/1986028

Insp Rept 50-440/86-28 on 861017-1202.Violation Noted: Failure to Provide Adequate Procedures for Activities Affecting Quality

ML20207D226
Person / Time
Site:	Perry
Issue date:	12/23/1986
From:	Knop R, Oestmann M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20207D149	List: IR 05000440/1986028 ML20207D144 ML20207D166
References
50-440-86-28, NUDOCS 8612300354
Download: ML20207D226 (11)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION III

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Report No. 50-440/86028(DRP)

Docket No. 50-440 License No. NPF-58 Licensee: Cleveland Electric Illuminating Company Post Office Box 5000 Cleveland, OH 44101 Facility Name: Perry Nuclear Power Plant, Unit 1 Inspection At: Perry Site, Perry, OH Inspection Conducted: October 17 through December 2, 1986 Inspectors: K. A. Connaughton G. F. O'Dwyer M. J. Destmann 7//.h 84y""- /MWNS RFil) J &

Approved By: R. C. Knop, Chief /2 /2 7/74

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Reactor Projects Section 18 Date Inspection Summary Inspection on October 17 through December 2, 1986 (Report No. 50-440/80628(DRP))

Areas Inspected: Routine unannounced inspection by resident and region based inspectors of previous inspection items, I.E. Bulletins, I.E. Information Notices, licensee actions regarding General Electric Service Information Letter No. 445, inspection and testing of seismic monitoring instrumentation, operational safety, Licensee Event Reports, nonroutine events, and onsite review committee activities. Management meetings were held on October 24, and November 21, 1986 to discuss overall plant status and recent Reportable Event Results: Of the 9 areas inspected, one violation was identified in one area (failure to provide adequate procedures for activities affecting quality -

Paragraph 7). During this inspection period, the licensee completed nuclear-heatup phase startup test activitie pp P G

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DETAILS l

1. Persons Contacted A. Kaplan, Vice President, Nuclear Operations Division

• 1. C. M. Shuster, Manager, Nuclear Engineering Department (NED)

1. M. D. Lyster, Manager, Perry Plant Operations Department (PP00)

• 1. R. A. Stratman, General Supervising Engineer, Operations Section, (PPOD)

• 1. R. P. Jadgchew, General Supervising Engineer, Instrumentation and Controls Section (PP00)

• A. F. Silakoski, Operations Section (PP00)

• G. R. Anderson, Instrumentation and Contr ols, (PP00)

• L. R. Teichman, Maintenance Planning, (PPOD)

• 1. F. R. Stead, Manager, Perry Plant Technical Department (PPTD)

• 1. S. F. Kensicki, Technical Superintendent (PPTD)

• P. A. Russ, Licensing and Compliance Sectien (PPTD)

• D. C. Jones, Licensing and Compliance Section (PPTD)

• G. S. Cashell, Licensing and Compliance Section (pPTD)

• 1. R. A. Newkirk, Technical Section (PPTD)

• B. D. Walrath, General Supervising Engineer, Operational Quality Section (NQAD)

• 1. V. K. Higaki, Maintenance and Modification Quality Section, (NQAD)

• Denotes those attending the exit meeting held on December 2, 198 # Denotes those attending the Status Meetings held on October 24, 1986, and November 21, 198 . Licensee Action on Previous Inspection Findings (92701, 92702)

(Closed) Open Item (440/85070-01(DRS)): The licensee agreed to issue a corporate policy regarding water chemistry control within six months after operating license issuance. The licensee issued Plant Operating Procedure (P0P)-0804, " Plant Chemistry Control Policy," Revision 0, on August 28, 1986, and administrative procedure OMIA, Plant Administrative Procedure (PAP)-1102, " Plant Chemistry Control Program," Revision 0, on September 9, 1986. These procedures were issued within six months after operating license issuanc . Inspection and Enforcement Bulletin (IEB) Followup (92703)

(Closed) IE Bulletin 86-03, " Potential Failure of Multiple ECCS Pumps Due to Single Failure of Air Operated Valve in Minimum Flow Recirculation Line." The inspector reviewed the licensee's response letter dated November 10, 1986 which documented the results of the licensee's review of the subject IEB. The licensee determined that the Perry design was not vulnerable to multiple ECCS pump failures as a result of a single failure in any of the minimum flow recirculation lines provided these pumps. The inspector confirmed the licensee's findings by review of the Perry FSAR as well as current revisions of controlled piping and instrument diagrams for all Perry ECCS subsystems. The inspector further noted that the licensee's response was submitted within the 30 day time frame specified in the IE .

. Followup of IE Information Notices and IE Bulletins Sent for Information (92701) Programmatic Review The inspector reviewed, on an audit basis, the licensee's program for the handling of NRC Information Notices and NRC Bulletins sent for information. The review was conducted to determine whether or not selected Information Notices had been received by licensee management, that a review for applicability was performed, and that for'each applicable Information Notice appropriate corrective actions were taken or scheduled to be taken. The inspector reviewed Project Administration (PA) procedure 1601, " Evaluation of IE/INP0 Documents." This procedure, which remained in effect until October 3, 1986, required a review for applicability and if applicable, a designation of responsibility for disposition of all IE Bulletins, Circulars, and Information Notices as well as INPO SERs and SOERs. On October 3, 1986, these requirements were incorporated into Reliability and Design Assurance Section Instruction 1602, " External Operating Experience Reports (0ER)

Review Processing Instruction," and Plant Administrative Procedure 0604, "NRC IE Documents and INPO Reports."

The inspector randomly selected the following Information Notices issued between August 1985 and October 1986 for detailed review:

85-88,85-100, 86-10, 86-22, 86-34, 86-41, 86-51, 86-64, 86-65, and 86-80. 'The sample size was approximately 10% of the total number of Information Notices issued over that time frame. The review of the selected Information Notices disclosed that a documented review for applicability was made in each case. The inspector agreed with the results of each review. Where the Information Notice was determined to be applicable to Perry, appropriate personnel received "

the Notice for further action. The appropriate actions were documented and taken or were scheduled to be taken. Before an Information Notice was considered closed, a final review of actions planned or taken was made by a review group within the Reliability and Design Assurance Section of the licensee's Nuclear Engineering Department. The inspector agreed that all of the actions taken with regard to the Information Notices reviewed were proper and .

technically soun Based on this review of licensee actions with regard to Information a Notices, the inspector concluded that the licensee's program resulted in thorough and well documented reviews as well as appropriate actions with regard to identified issues. Detailed discussions of licensee actions regarding Information Notice 86-53 and Information Notice 86-72 are provided in Paragraphs b and c below. Inspector followup of these Information Notices was conducted in response to specific requests received from the NRC Region III Offic . - - - . . -_ - -

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I IE Information Notice 86-53 IE Information Notice 86-53, " Improper Installation of Heat

. Shrinkable Tubing."~ The inspector reviewed the licensee's file response-to NRC Information Notice.86-53 and interviewed licensee personnel to determine the extent of identified problems involv'ing improper installation-of heat shrinkable tubing at Perry. .The licensee had not identified any problems in this area aside from a problem reported pursuant.to 10 CFR 50.55(e) in 1984. The problem concerned the use of heat shrinkable tubing in containment which was not previcusly specified for use in containment. Subsequent review of vendor specifications determined that-the' tubing was acceptable for use in containment and the 10 CFR 50.55(e) report was withdraw The licensee was satisfied that the installation. instructions and training provided for craft and QC inspection personnel were adequate to preclude the existence of problems identified in Information Notice 86-5 The inspector informed the licensee that further review of the licensee's procedures and practices with regard to the installation of heat shrinkable tubing would be conducted during a future NRC inspectio IE Information Notice 86-72 IE Information Notice 86-72, " Failure 17-7 PH Stainless Steel Springs in Valcor Valves Due to Hydrogen Embrittlement." The inspector verified, by review of licensee file information, that the licensee had received the subject IEN and distributed it to appropriate personnel within the licensee's organization for-a determination of applicability and an evaluation of the need for additional actions. The licensee's evaluation identified a total of 11 valves supplied by Valcor with 17-7 PH springs. In consultation with General Electric, the licensee concluded that spring failures described in the subject IEN would not be experienced at Perr This conclusion was based upon a review of valve application The failures identified in the subject IEN involved valves subject to various combinations of high temperature, hydrogen rich process

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fluids, and process fluid flow conditions. The valves supplied to Perry were utilized in scram air headers and scram discharge volume vent and drain valve pilot air header In these applications, the valves were not subject to the conditions to which valve spring failures had been attribute The inspector verified the valve applications were as stated in the licensee's evaluation by comparison of Master Parts List (MPL) valve identification numbers and controlled, as-built piping and instrumentation diagrams. The inspector was satisfied that the licensee had adequately addressed the issue identified in the subject IE .

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5. Followup--General Electric Service Information Letter No. 445 (92701)

In response to a request from the NRC Region III office, the inspector conducted a review of licensee actions with regard to General Electric Service Information Letter (SIL) No. 445. General Electric SIL-445 described several deficiencies encountered in a multiple fuse failure event in the intermediate range monitoring channels of the neutron flux measuring system at the Monticello Nuclear Plant in mid-June 198 As a result of General Electric's evaluation of this event, several recommendations were made to BWR licensees. Two of the recommendations were applicable to Perr The first recommendation was that the licensee evaluate procedures for establishing operability of safety related instrument channels following their loss. The recommendation further stated that licensee procedures should specify that channel functional tests be performed on each associated channel upon completion of fault repair. The channel functional tests should be performed in accordance with Technical Specification surveillance procedures. The second recommendation made by General Electric was that the licensee consider adding voltage sensing relays which would monitor each SRM/IRM chassis to provide a reactor protection system trip in response to loss of negative power. The inspector reviewed licensee file information to determine what actions if any had been planned or taken in response to these two recommendation Regarding the first recommendation, the licensee determined that existing procedures for controlling corrective maintenance activities were adequate for the specification of retest requirements following instrument repair. Inspector review of the applicable procedure, PAP-0905, " Work Order Process," determined that while the procedure assigned responsibility for the specification of retest requirements, it did not specifically incorporate the recommendation for channel functional testing of safety related instrument channels following their loss and completion of fault repair. The inspector informed the licensee that this procedure did not specifically incorporate General Electric's recommendation regarding channel functional testing. This matter is considered an unresolved item (440/86028-01(DRP)).

Regarding the second recommendation, the licensee initiated Engineering Design Change Request 860977, November 6, 1986. This design change, which was to be implemented by first refuelling will provide nonitoring, alarm, and trip functions for the negative 20 volt DC IRM power supply, as recommende . Periodic Inspection and Testing of Seismic Monitoring Instrumentation (61726,62703)

During this inspection period, the inspector received a request from the NRC Region III office to perform an initial inspection of licensee activities pertaining to seismic monitoring instrumentation preventive maintenance, testing, and operability. The inspector reviewed applicable

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sections of the Perry FSAR, Technical Specifications, and Vendor manuals to become familiar with. seismic monitoring instrumentation, maintenance, surveillance test, and operability requirements. The inspector reviewed the following Surveillance Test Instructions (SVIs) which implemented these requirements:

SVI N Title

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D51-T0278 Triaxial time-history accelerographs channel functional for D51-N101 and D51-N111 051-T0279 Triaxial time-history accelerograph channel calibration for D51-N101 and D51-N111 D51-T0289-A Triaxial peak accelerographs channel calibration for D51-R120 (Reactor Recirculation Pump)

D51-T0289-B Triaxial peak accelerographs channel calibration for D51-R130 (HPCS Piping in Reactor Building)

D51-T0289-C Triaxial peak accelerographs channel calibration for D51-R140 (HPCS Pump Base Mat)

D51-T0294 -Triaxial seismic switch channel functional test for D51-N150 D51-T0295 Triaxial seismic switch channel calibration for D51-N150 051-T0302- Seismic instruments channel check 051-T0304-A Triaxial response spectrum recorder channel calibration for D51-R160 (Reactor Building Foundation)

D51-T0304-8 Triaxial response spectrum recorder channel calibration for D51-R170 (Reactor Recirculation Piping Support)

051-T0304-C Triaxial response spectrum recorder channel calibration for D51-R180 (HPCS Pump Base Mat)

051-T0304-D Triaxial response spectrum recorder channel calibration for D51-R190 (RCIC Pump Base Mat)

The inspector determined that the foregoing procedures were written, reviewed, and approved in accordance with licensee administrative controls and that each Technical Specification Surveillance Test requirement was addressed in these procedures. The inspector also determined that these procedures contained recommended preventive maintenance items contained in seismic monitoring instrumentation vendor manual Based upon discussions with licensee technical personnel, the inspector determined that seismic monitoring instrumentation has been maintained in

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an operable status since issuance of the Perry, Unit 1 operating license on March 28, 1986, except for brief periods when surveillance testing was being performed. The inspector will review current seismic monitoring instrumentation surveillance test results during a future inspectio No violations or deviations were identifie . Operational Safety Verification (71707)

The inspectors observed control room operations, reviewed applicable logs, and conducted discussions with control room operators during this inspection period. The inspectors verified the operability of selected emergency systems, reviewed tag-out records and verified tracking of Limiting Conditions for Operation associated with affected component Tours of the intermediate, auxiliary, reactor, and turbine buildings were conducted to observe plant equipment conditions including potential fire hazards, fluid leaks, and excessive vibrations, and to verify that maintenance requests had been initiated for certain pieces of equipment in need of maintenance. The inspectors by observation and direct interview verified that the physical security plan was being implemented in accordance with the station security pla '

The inspector observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection control These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under technical specifications,10 CFR, and administrative procedure During initial surveillance testing of the Reactor Core Isolation Cooling System (RCIC), leak detection system RCIC steam flow transmitter 1E31-N083A was observed to be indicating improperl Subsequent investigation by the licensee determined, among other things, that the incorrect instrument readings were due to the collection of condensed steam in portions of the instrument sensing lines. As an interim corrective action, the licensee installed heat tracing on the affected portions of the sensing lines to maintain sensing line temperature above saturation temperature and thereby eliminate the condensation. The installation of the heat tracing was accomplished between September 26, and September 30, 1986. Subsequently, on November 2, 1986 during testing of the RCIC system, licensee personnel observed a portion of the sensing line glowing red and the presence of scaling. Following this observation, the heat tracing was deenergized. Licensee investigation and evaluation of this occurrence concluded that temperature sensors utilized for temperature control had been placed such that they did not provide temperature readings which were representative of actual sensing line temperature (i.e. indicated temperature was lower than actual).

Based upon initial observations of the sensinj lines and subsequent visual examination of scale which had accumulated on portions of the sensing lines, the licensee estimated that the sensing lines had reached a maximum temperature of approximately 1400 F. Licensee metallurgical specialists reviewed the event for impact upon sensing line integrity and

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determined that while the overheating event resulted in a decrease in sensing line corrosion resistance, the physical integrity of the sensing line was unimpaired. Since the sensing lines were not exposed to a corrosive process medium, the licensee determined that the sensing lines were acceptable "as is" on an interim basis and that the sensing lines would be *eplaced following the nuclear heatup phase of the startup test progre ,. The inspector, in consultation with NRC Region III metallurgical specialist inspectors, determined that the licensee's disposition of the instrument sensing lines was acceptabl The inspector reviewed Work Order Nos. 860013435 and 860013436 which were utilized as temporary work instructions to accomplish installation of the heat tracing. The inspector also reviewed Temporary Instruction (TXI)

0005, " Temporary Heat Trace for RCIC Leak Detection," Revision 0, dated September 28, 1986. Based upon these reviews, the inspector determined that the instructions contained in the work orders were inadequate in that no guidance was provided as to the number and/or placement of temperature sensors to assure that temperature readings were representative and valid for use in controlling sensing line temperatur Further, as of the time of this inspection, the licensee had not developed generic instructions for the proper installation of temporary heat tracing in safety related application Failure to provide adequate instructions, procedures, or drawings for the installation of temporary heat tracing on the safety-related leak detection system RCIC steam flow transmitter sensing lines is contrary to 10 CFR 50, Appendix B, Criterion V and is considered a Violation (440/86028-02(DRP)).

8. Licensee Event Reports Followup (92700)

Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with technical specification LER 86001-1 " Electrical Noise Caused Spikes on IRMs Resulting in RPS Actuations" LER 86025-0 "CVDPS Design Deficiency Causes Containment Vacuum Relief Valves to Open" LER 86027-0 " Personnel Errors During Design Change Installation Cause RWCU and RHR Isolations" LER 86033-0 " Instruction Deficiency Causes Inadvertent HPCS Diesel Generator Autostart "

LER 86037-0 " Faulty Leak Detection Switch Causes Residual Heat Removal System Isolation" LER 86038-0 " Instrument Inaccuracy Causes AEGTS Autostarts"

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LER 86042-0 " Control Tachometer Problems Cause Diesel Generator Building Fan Autostarts" LER 86049-0 " Personnel Errors Result in Missed Gaseous Effluent Vent Stack Flow Estimates" LER 86050-0 " Personnel Error Results in Loss of Reactor Protection System Bus" No violations or deviations were identifie . Onsite Followup of Non-Routine Events at Operating Reactors (93702)

At approximately 7:24 a.m. on November 26, 1986, while in operational condition 2 with reactor power at approximately 2%, a reactor scram occurred due to upscale trips of intermediate range monitoring (IRM)

instruments A, D, and E. The power transient was caused by an increase in feedwater flow to the reactor vessel and resultant void collaps The feedwater increase occurred when operating personnel were shifting feedwater supply from motor driven feed pump "C" to turbine driven feed pump "A".

The inspector arrived in the Control Room approximately 30 minutes after the scram. The inspector observed that the reactor had been placed in hot shutdown and that all control rods were fully inserted. The inspector was briefed immediately by licensee personnel. The inspector observed discussions between Operating, and Instrumentation and Control System engineering personnel regarding operator actions and feedwater control system behavior at the time of the even On November 27, 1986, the licensee issued Scram Evaluation Report No. 86-4 which documented the licensee's evaluation of the event including a determination of root cause. The licensee's evaluation concluded that the excessive feedwater addition was primarily due to operator erro In preparation for placing the "A" reactor feed pump in service on the startup level controller, the operator began increasing reactor feed pump "A" output to match the startup level controller output. The operator failed to recognize soon enough that feedwater flow to the vessel had exceeded demand because his attention for a period of time was focused exclusively on feedwater controller respons Corrective actions taken by the licensee included remedial training of the operator involved and procedural enhancements to System Operating Instruction (S0I) C34, "Feedwater Control System," to caution operators to monitor feedwater pump flows and to allow time for pump flows and controller outputs to stabilize. The inspector determined these corrective actions to be adequat The licensee's evaluation of plant equipment response following the scram determined that all equipment functioned per design with one exception; scram discharge volume drain valve IC11-F011 could not be remote-manually reopened after the scram was reset. Subsequent investigation by the licensee disclosed that the valve actuator stem had become uncoupled from

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the valve disk stem. Corrective actions by the licensee included replacement of the valve ~ actuator coupling and verification of proper bolt torque on all scram discharge volume vent and drain valve actuator couplings.. During inspector review of this matter, the inspector noted that scram discharge volume vent and drain valve coupling failures had been experienced at several boiling water reactors and were discussed in NRC Information Notice 86-82, " Failures of Scram Discharge Volume Vent and Drain Valves," Revision 0, dated September 16, 1986 and Revision 1, dated November 4, 1986. Failures reported in the Information Notice were attributed to automatic valve actuations which occurred while the handwheels provided for manual valve operation were partially engaged with the valve stem. Further inspector review of licensee actions to determine the cause of the valve failure in light of the failures reported in NRC Information Notice 86-82 will be tracked as an open item (440/86028-03(DRP)).

10. Onsite Review Committee (40700)

The inspectors reviewed the minutes of the Plant Operations Review Committee (PORC) meetings No. 207 through 242 and 244, conducted prior to and during the inspection period to verify conformance with PNPP procedures and regulatory requirements. These observations and examinations included PORC membership, quorum at PORC meetings, and PORC activitie No violations of regulatory requirements or deviations from commitments were identified in this are . Plant Status Meetings (30702)

On October 24, and November 21, 1986, NRC management met with CEI management at the Perry site to discuss the current status of the plant and recent events. These meetings are being held with CEI management on a periodic (initially monthly) basi Key personnel attending the meetings are identified by a #, in Paragraph 1 of this repor The meetings each included discussions of: the status of the plant; recent Licensee Event Reports (LER); corrective actions taken or planned to be taken to preclude repetition; and the schedule for future evaluation . Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether it is an acceptable item, a violation or a deviation. An unresolved item is identified in Paragraph . Open Inspection Items Open inspection items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or both. Open

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inspection items disclosed during the inspection are discussed in Paragraph . Exit Interviews (30703)

The inspectors met with the licensee representatives denoted in Paragraph 1 throughout the inspection period and on December 2, 1986. The inspector summarized the scope and results of the inspection and discussed the likely content of the inspection report. The licensee did not indicate that any of the information disclosed du'ing r the inspection could be considered proprietary in natur i l'

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