IR 05000354/1986040

Insp Rept 50-354/86-40 on 860812-0908.Violation Noted: Operator Aids Identified in 120-volt,IE Electrical Distribution Panels 1AJ481,1BJA81,1CJ481 & on Relay Cabinets in Lower Control Equipment Room Not Formally Approved

ML20206U612
Person / Time
Site:	Hope Creek
Issue date:	09/25/1986
From:	Norrholm L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20206U593	List: IR 05000354/1986040 ML20206U590 ML20206U602
References
50-354-86-40, NUDOCS 8610080049
Download: ML20206U612 (10)

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br U. S. NUCLEAR REGULATORY COMMISSION

REGION I

050354-860709 050354-860712 Report N /86-40 050354-860712 050354-860713 Docket 50-354 050354-860714 050354-860715 License NPF-57 050354-860717 050354-860717 Licensee: Public Service Electric and Gas Company 050354-860719 Facility: Hope Creek Generating Station Conducted: August 12, 1986 - September 8, 1986 Inspectors: R. W. Borchardt, Senior Resident Inspector D. K. A sopp, Resident Inspector Approved: . _/1 - 3 L. jtorrh'olm, Chief, Reactor Projects / Datd 54ction 2B Inspection Summary:

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Inspection on August 12, 1986 - September 8, 1986 (Inspection Report Number 50-354/86-40)

Areas InspecteJ: Routine onsite resident inspection of the following areas: Operational safety verification, surveillance testing, maintenance activities, engineered safety feature system walkdown, and licensee event report followup. This inspection involved 92 hours0.00106 days <br />0.0256 hours <br />1.521164e-4 weeks <br />3.5006e-5 months <br /> by the inspector Results: This inspection identified a violation of the Station Aids Program (SA-AP.ZZ-44), in that numerous unapproved aids were identified throughout the plant. It should be noted that this condition has been brought to the attention of plant personnel on previous occasions but has not been corrected. Because of the uncontrolled manner in which these aids are used there is a high risk of mislabeling errors. Further information can be found in paragraph 2.3. It is also noted that the differences between the as-built feedwater control system and the simulator may have contributed to the reactor scram experienced on August 31, 1986. Had the control room operators been able to more quickly restore a reactor feed pump to service, the low water level condition may have been prevented. The operators were unable to reset the RFP turbine in a timely manner because the simulator did not accurately reflect the as-built configuration. This item is discussed in paragraph 3 of this repor e610090049 e60926 4 PDR ADOCK 0500 G

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Details 1. Persons Contacted l

Within this report period, interviews and discussions were conducted with members of the licensee management and staff and various contractor personnel as necessary to support inspection activit . Operational Safety Verification 2.1 Documents Reviewed

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Selected Operator's Logs

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Senior Shift Supervisor's Log

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Jumper Log

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Radioactive Waste Release Permits (liquid & gaseous)

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Selected Radiation Work Permits (RWP)

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Selected Chemistry Logs

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Selected Tagouts

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Health Physics Watch Log 2.2 The inspectors periodically toured the plant during regular and -

backshift periods. These tours included the control room, Reactor, Auxiliary, Turbine and Service Water buildings, and the drywell (when access is possible). During the inspection, discussions were held with operators, technicians (HP & I&C),

mechanics, supervisors, and plant management. The purpose of the inspection was to affirm the licensee's commitments and compliance with 10 CFR, Technical Specifications, and Station Procedure (1) On a daily basis, particular attention was directed to the following areas:

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Instrumentation and recorder traces for abnormalities;

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Adherence to LCO's directly observable from the control room;

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Proper control room shift manning and access control;

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Verification of the status of control room annunciators that are in alarm;

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Proper use of procedures;

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Review of Logs to obtain plant conditions; and,

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Verification of surveillance testing for timely completio (2) On a weekly basis, the inspectors confirmed the operability of selected ESF-trains by:

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Verifying that accessible valves in the flow path were in the correct positions;

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Verifying that power supplies and breakers were in the correct positions;

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Visually inspecting major components for leakage, lubrication, vibration,. cooling water supply, and general operating conditions; and,

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Visually inspecting instrumentation, where possible, for proper operabilit (3) On a biweekly basis, the inspectors:

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Verified the correct application of a tagout to a safety-related system;

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Observed a shift turnover;

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Reviewed the sampling program including the liquid and gaseous effluents;

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Verified that radiation protection and controls were

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Verified that the physical security plan was being implemented;

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Reviewed licensee-identified problem areas; and,

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Verified selected portions of containment isolation lineu .3 Inspector Comments / Findings:

2. The unit entered this report period in operational condition 2 with the reactor operating at approximately 10% power, making preparations to synchronize the main turbine generator with the grid. At 3:35 a.m. on August 13, the reactor was placed in operational condition 1 and the main generator was synchronized to the gri __ _ .- _ . . . _ _ . _ - _ . - ~ _ _ . . _ _ _ _ _

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2. At 2:40 p.m. on August 14, the licensee received a spurious channel "B" reactor vessel level oscillation which actuated numerous engineered safety feature (ESF) systems. This is the first and only unexplained actuation of this type to occur since implementation of corrective actions detailed in Licensee Event Report 86-039-00. The licensee is continuing to investigate this perturbation and is monitoring all loss of coolant accident logic channels to determine if other factors should be addressed to further minimize the probability of such events. The resident inspectors will continue to follow the licensee's actions on this event and any future level perturbation .3.3 At 3:12 p.m. on August 19, a backup protective relay caused a trip of the main generator. At the time of the event, the unit was at 16% power (below the power level at which a turbine / generator trip would cause a scram). Following the mild transient, the reactor was stabilized at approximately 18% power with five bypass valves open. All systems performed as expected. The licensee replaced the relay and continued test condition one testin .3.4 At 1:50 p.m. on August 22, a spurious high pressure coolant injection (HPCI) actuation occurred, but did not inject-water into the reactor vessel. All systems functioned as designed for plant conditions (operational condition 1,19.5 % power). The root cause has been determined to be operation of the post accident sampling system (PASS) which is isolated from vessel level common reference lines by a single Rosemount transmitter. The licensee has determined that a pressure spike can be transmitted through a Rosemount transmitter and has recreated this event (when shutdown) by drawing a PASS sampl For corrective action, the licensee has included in the PASS procedure the requirement to isolate the Rosemount transmitter prior to drawing a PASS sampl .3.5 The shutdown from outside the control room test was conducted on August 22, at about 3:15 p.m. The resident inspector and region based inspectors were on site to evaluate this tes Near the completion of the test, at 4:50 a.m. on August 23, the licensee decreased pressure to establish conditions for initiation of shutdown cooling. The "H" safety relief valve (SRV) was manually opened from the remote shutdown panel at approximately 82 psig. The valve was observed by control room indications and

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alarms to fail open. The licensee completed action for a stuck open relief valve but was unable to shut the SRV. The plant stayed shutdown until the "H" SRV was replaced. For additional information on the shutdown from outside the control room test, refer to Inspection Report 86-38. Inspection of the faulty SRV showed that the cause of failure was damage to the solenoid actuating valve. The SRV itself was undamaged and was successfully retested on a test stan .

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2.3.6 While the inspector was conducting a routine inspection of the reactor building, he determined that the " scram dump valve" (F 182 A&B) was incorrectly labeled "back-up scram pilot valve".

This labeling error was brought to the attention of the system engineer. The correct labeling of the scram dump valve (F 182 A&B) will be tracked as an open item. (86-40-01)

2.3.7 During the licensee's investigation into the inoperative torus vacuum breaker butterfly isolation valve, (see paragraph 4) a potential problem was identified with ASCO air operated solenoid valves. ASCO recommends no restrictions in the air supply lines to ensure 10 psid is achieved between the pressure port and exhaust port during valve operation. The licensee currently utilizes air regulators (restrictions) on the air supply piping on 33 separate valves which may prevent the valves from meeting the 10 psid requirement The licensee believes these valves to be operable for the following reasons: 1) calculations have been performed which indicate the 10 psid requirement is met; 2)

all these valves have been stroked during preoperational testing and meet test requirements; and 3) there are no documented incidents of valves failing to operate. After an engineering evaluation, the licensee replaced the air supply piping on both torus vacuum breaker butterfly isolation valves with larger diameter piping to ensure the 10 psid requirement was met. The licensee is currently writing procedures to verify the 10 psid requirement is met in all 33 valves. During this engineering evaluation, the licensee identified a separate operability issue and, at 12:00 p.m. on August 28, declared 12 air operated ASCO solenoid valves used in the Safety Auxiliary Cooling system and Containment Atmosphere Control system inoperable. These ASCO valves were supplied with air through a non-Q regulator attached to a 120 psig air header. When it was discovered the ASCO valves were only rated for 100 psig supply pressure, they were declared inoperable based on the possibility the non-Q regulators could fail, exposing the ASCO valves to 120 psig air pressure. The reactor was in Mode 4 when the ASCO valves were declared inoperable. The licensee upgraded the inoperable valves to valves rated for 175 psig supply pressure. These upgraded valves do not require the supply air regulators which were removed to minimize piping restrictions. After removing the 12 supply air regulators and upgrading the 12 ASCO air operated solenoid valves, the licensee took the reactor critical on August 31, 2.3.8 At 9:43 p.m. on August 31, the licensee received a low vessel level reactor scram. A loss of level control occurred during plant startup when a secondary condensate pump was started. The startup control system was controlling reactor level since the feedvater control system does not have single element control capability. To improve reactor water level control, the 6A

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feedwater heater bypass valve was opened approximately 30%.

This allowed the 3" startup control valve to control level while the 12" control valve was shut. When the second secondary condensate pump was started, the 12" control valve cycled opened and reactor vessel level increased to level 8 which tripped both reactor feed punp turbines (RFPT). The operators were unable to restart the RFPT to regain level control before receiving a low level reactor scram. Contributing significantly to this scram was the operator's inability to quickly reset a RFPT. This operator training deficiency has been attributed to differences between the simulator RFPT reset logic and the actual plan The licensee required all operators to review the RFPT reset procedure and mounted an approved operator aid on the RFPT control panel detailing the RFPT reset procedure. The licensee reviewed the event, took the reactor critical, and entered operational condition 1 at 2:21 a.m. on September . At 6:17 a.m. on September 6, the unit scrammed from 38% power due to low water level (level 3) in the reactor vessel. The low water level condition occurred during reactor feed pump (RFP)

minimum flow valve (MFV) response tuning. In preparation for tuning the "C" RFP MFV, the "C" RFP was paralleled with the running RFP. While paralleling RFPs, the "C" MFV began oscillating which resulted in reactor level oscillations and a level 3 reactor scra The MFV controls about 30-40% of the RFP output. The licensee instituted the requirement that all subsequent RFP shifts will be performed with the RFP MFV in manual until all MFVs have been ,

completely tune I 2.3.10 During the course of this inspection, an increase in the use of unauthorized operator aids was observed. These aids included hand written ider.tification labels on instrumentation panels and unauthorized schedule cards inside of electrical distribution panels. Of specific concern was the identification of the schedule cards inside of 120 Volt, class 1E distribution panels (IAJ481, IBJ481, ICJ481,1DJ481) that were both unauthorized and incorrect. These schedule cards showed fuse ratings that were not in accordance with electrical drawing E0012-1 or the as-built cor. figuration. It also appears that the Instrument and Controls (I&C) technicians have developed the practice of hand '

writing labels on the instrument and relay panels throughout the plant to more easily identify the function of individual components. The inspector notes that labeling would be beneficial, but the method currently in use is uncontrolled and not in accordance with plant procedures. The inspector informed plant management and Quality Assurance that the use of unapproved operators aids was not in accordance with Station Administrative Procedure SA-AP.ZZ-44(Q) " Station Aids Program" and was considered a violation. (86-40-02)

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The inspector reviewed Quality Action Request (QAR) HS-86-Q045-0 and surveillance report 86-296 which were initiated as a result of this violation. The response to the QAR and the corrective actions taken will be reviewed in a future repor . Surveillance Testing During this inspection period the inspector performed detailed technical procedure reviews, reviewed in progress surveillance testing as well as completed surveillance packages. The inspector also verified that the surveillances were performed in accordance with licensee approved procedures and NRC regulations. The inspectoa also verified that the instruments :ised were within calibration tolerances and that qualified technicians performed the surveillance The following surveillances were reviewed, with portions witnessed by the inspector:

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IC-CC.SE-019 Channel Calibration Nuclear Instrumentation System

- Rod Block Monitor

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IC-FT.SK-011 Functional Test Main Steam Line Tunnel High Temperature (B21-N605C)

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IC-FT. AB-033 Functional Test Main Steam - Safety Relief Valve Position Indication (Acoustic Monitoring) System

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IC-FT.SP-004 Functional Test Process Radiation Monitor Division 4, Channel D Main Steam Line Radiation Monitor No violations were identifie . Maintenance Activities During this inspection period the inspector observed selected maintenance activities on safety related equipment to ascertain that these activities were conducted in accordance with approved procedures, Technical Specifications, and appropriate industrial codes and standard Portions of the following activity was observed by the inspector:

Work Order Description 8608230291 Bailey 745 Alarm Module Staple Jumper Verification

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m On August 8, 1986, the licensee ~ determined that the Reactor Building to Suppression Chamber Pressure Relief (RBSCPR) system was inoperable and conducted a reactor shutdown. An investigation was conducted to determine the cause for the inoperable condition. A description of the event, the results of the investigation, and the corrective actions taken, are discussed in special Inspection Report 50-354/86-41. As a result of this investigation, an apparent problem was identified regarding the staple jumper configuration on the Bailey 745 Alarm modules used in the RBSCPR system as well as other plant systems. Of particular concern was the apparent lack of explicit information on some system design drawings which forced the startup engineer to determine the proper staple jumper configuration during the preoperational test program. A contributing factor to the

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confusion in this area was the fact that General Electric (GE) and Bailey use slightly different conventions on the system design logic drawings, and not all of the system engineers were completely familiar with the Bailey method. Therefore, even though the necessary information was available on the Bailey drawings, the system engineers did not always correctly translate the information into the proper staple jumper configuratio The investigation into the RBSCPR system logic determined that although the logic cards functioned as intended, the staple jumper configuration was not in accordance with the design drawing Additionally, a loss of logic card power would not result in a fail safe condition. In all cases however, a loss of motive power to the system component would result in a fail safe condition as required by the General Design Criteri The licensee conducted a review and in-field verification of all Bailey 745 logic cards installed in the plant. Twenty-six discrepancies were identified out of a population of 142. All of the discrepant cards were similar to the RBSCPR cards described above, in that the system functioned properly, but would not have failed safe on a loss of logic card power. Corrections to the cards were made in the field when require The inspector witnessed portions of this in-field verification process performed under work order 8608230291, and has no further questions at this time.

l Although this issue appears to be a violation of 10 CFR 50 Appendix B

Criterion III, " Design Control" the inspector notes that 1) it was identified by the licensee, 2) it fits in Severity Level IV, and 3)

corrective actions have been completed in a timely manne For the above reasons, no violations were cite .

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.- Engineered Safety Feature (ESF) System Walkdown The inspectors verified the operability of the selected ESF systems by performing a walkdown of accessible portions of the system to confirm that system lineup procedures match plant drawings and the as-built configuration. This ESF system walkdown was also conducted to identify equipment conditions that might degrade performance, to determine that instrumentation is calibrated and functioning, and to verify that valves are properly positioned and locked as appropriat The "B" Loop of Low Pressure Coolant Injection (Residual Heat Removal) system was inspecte No violations were identifie . Licensee Event Report Followup The licensee submitted the following event reports during the inspection period. All of the reports were reviewed for accuracy and timely submission. The asterisked reports received additional followup by the inspector to verify corrective action implementatio * LER 86-34 Main Steam Isolation Valve Closure and Subsequent Manual Scram LER 86-37 Failure to Comply With Technical Specifications Action Statement LER 86-38 Missed Channel Checks on Reactor Protection and Isolation Actuation LER 86-39 "A" Channel LOCA Logic Actuation

• LER 86-40 Inoperable RCIC Actuation Instrumentation

• LER 86-41 Inadvertent HPCI System Initiation

• LER 86-42 Inadvertent HPCI System Initiation LER 86-43 Inadvertent HPCI Sy: tem Initiation Due to An I&C Error LER 86-45 Reactor Scram Due to IRM Ranging Error

• LER 86-46 Inadvertent HPCI System Initiation LER 86-34 Describes a spurious high steam flow signal that shut all main steam isolation valves (MSIV) and resulted in the reactor being manually scrammed. The root cause of this event was two faulty flow transmitters which erroneously sensed high steam flow and shut all MSIV With the reactor isolated, power and pressure increased rapidly and a manual scram was inserted to prevent exceeding five

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percent power (low power license limit). The licensee's immediate corrective action was to test and replace the faulty flow transmitters, and to send the old transmitters to the manufacturer for additional testing to determine the cause of the failur LER 86-40 Details the sequence of events that resulted in the reactor core isolation. cooling system condensate storage tank auto-swap feature being disabled when required to be operable. Details of this event are discussed in paragraph 3.3 of NRC Inspection Report 86-3 LER 86-41, 86-42, and 86-46 Describe spurious high pressure coolant injection (HPCI) actuations. These events are all related in that the cause appears to be a hydraulic transient within the common reference sensing lines initiated by personnel in the drywell stepping on the sensing lines. Details of these events are described in paragraph 3.3 of NRC Inspection Report 86-36. The licensee's corrective action included extensive labeling of sensing lines in the drywell and the installation of two permanent ladders in the drywel . Exit Interview The inspectors met with licensee and contractor personnel -

periodically and at the end of the inspection report to summarize the scope and findings of their inspection activities. Written material was not provided to the licensee during the exi Based on Region I review and discussions with the licensee, it was determined that this report does not contain information subject to 10 CFR 2 restrictions.