IR 05000334/1987007

Insp Rept 50-334/87-07 on 870418-0604.Violations Noted: Unplanned Gaseous Release & Failure to Make Required 10CFR50.72 Rept in Timely Fashion.Three Unresolved Items Noted.Two Previously Open Unresolved Items Closed

ML20216G536
Person / Time
Site:	Beaver Valley
Issue date:	06/17/1987
From:	Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20216G462	List: IR 05000334/1987007 ML20216G458 ML20216G486
References
50-334-87-07, 50-334-87-7, NUDOCS 8707010112
Download: ML20216G536 (17)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /87-07 Docket N Licensee: Duquesne Light Company One Oxford Center 301 Grant Street j Pittsburgh, PA 15279 "

Facility Name: Beaver Valley Power Station, Unit 1 l l

Location: Shippingport, Pennsylvania j

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Dates: April 18 - June 4, 1987 Inspectors: J. E. Beall, Senior Resident Inspector, BV-2 )

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F. I. Young, Senior Resident Inspector, BV-1

$1 M. Pi' dale, Resident Inspector, BV-1 Approved by: / .

$0 h 7/87 (..E.Tripp, Chief,ReactorProjectsSection3A 'ddte '

Inspection Summary: Inspection No. 50-334/87-08 on April 18 - June 4, 1987 Areas Inspected: Routine inspections by the resident inspectors (188 hours0.00218 days <br />0.0522 hours <br />3.108466e-4 weeks <br />7.1534e-5 months <br />) of licensee actions on previous inspection findings, plant operations, physical !

security, radiological controls, housekeeping and fire protection, maintenance, surveillance testing, in-office review of LERs and licensee events, i

Results: Two violations were identified for evolutions resulting in an unplanned gaseous release (Detail 4.2.4) and failure to make a required 10 CFR 50.72 report '

in a timely fashion (Detail 7.3). Three unresolved items were identified concern-ing potential auxiliary feedwater pump performance degradation (Detail 6), lack of adequate control over sampling evolutions (Detail 7.2) and non-conservative I Overpressure Protection System setpoints (Detail 8). Two previously open NRC un-resolved items were closed this inspectio ;

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i TABLE OF CONTENTS j PAGE Persons Contacted.................................................... 1 Summary of Facility Activities................... ................... 1 Followup on Outstanding Items........................................ 1 Plant 0perations..................................................... 3 General......................................................... 3 Operations...................................................... 3 1 4.3 Plant Security / Physical Protection.............................. 6 l 4. 4 Radiation Controls..............................................

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4.5 Plant Housekeeping and Fire Protection.......................... 7 Maintenance Activities............................................... 8 Surveillance Testing................................................. 9 ESF Actuations....................................................... 10 Non-Conservative OPPS Setpoints...................................... 12 In-Office Review of LERs............................................. 13 10. Review of Periodic and Special Reports. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 11. Unresolved Items........... ......................................... 14

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12. Exit Interview....................................................... 15

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A DETAILS 1. Persons Contacted During the report period, interviews and discussions were conducted with mein-bers of licensee management and staff as necessary to support inspection activitie . Summary of Facility Activities At the beginning of the inspection on April 18, 1987, the plant was operating at 100% power. On April 24, a planned shutdown was initiated primarily to remove the wall separating Unit 1 and Unit 2 control rooms. NRC Chairman Zech toured Beaver Valley Power Station (BVPS) and met with the senior Duquesne Light Company (DLC) management on May 11, 1987. The plant completed shutdown activities and returned to full power on May 31, 1987. On June 1, a turbine /

reactor trip occurred due to turbine control problems (Detail 4.2.3). The unit returned to full power on June 3 and continued through the end of this inspectio . Followup on Outstanding Items The NRC Outstanding Items (0I) List was reviewed with cognizant licensee per-sonnel. Items selected by the inspector were subsequently reviewed through discussions with licensee personnel, documentation reviews and field inspec-tion to determine whether licensee actions specified in the OIs had been satisfactorily completed. The overall status of previously identified in-spection findings were reviewed, and planned / completed licensee actions were discussed for those items reported below:

3.1 (Closed) Unresolved Item (85-18-01): The licensee was to resolve an in-consistency between separate technical specifications (TS) requirements regarding the ESF instrumentation actuation operability versus the main steam isolation valve (MSIV) hardware operability. Specifically, TS 3.7.1.5 required each MSIV to be operable when the plant is in modes 1 through 3, while the high steam pressure rate actuation logic for the ESF actuation system instrumentation (TS Table 3.3.3) was required to be operable when the plant is in Modes 3 and 4. A high steam pressure rate signal initiates a steam line isolation signal which_in turn, closes the MSIV By letter dated July 25, 1986, (TS Change Request No. 125), the licensee requested to modify plant TS to resolve the inconsistency. On April 21, 1987, NRC Licensing issued TS Amendment No. 108 which deleted the mode 4 requirement for the high steam pressure rate initiatio This change is consistent with current guidance provided in NUREG-0452, Standard Technical Specifications for Westinghouse Pressurized Water Reactors, Rev. 4. No hardware changes were associated with the TS ievision. The inspector verified that the appropriate plant procedures were revised to reflect the TS change. This item is close ]

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3. 2 (0 pen) Unresolved Item (86-18-01): The licensee was to identify the cause of the failure of pressurizer safety valve RV-RC-551C. On August 7, 1986, the safety valve was replaced due to indication that a bellows failure was evident during reactor coolant system heatup. The valve was replaced with a qualified spare manufactured by Target Rock. During the subse-quent heatup and pressurization to 2235 psig, a flange leakage developed on the installed spare valve. The licensee's investigation of the flange leak found that the seating gasket had been crimped. The valve was sub-sequently repaired satisfactoril The valve, RV-RC-5510, was sent offsite for repair and testing, which determined that the leakage was from the valve's main stage in the main disc and that the valve seat was full of dirt and rust. Additionally, the offsite laboratory discovered that a plug and 0-ring were missing from the valve. The laboratory report stated that the parts would be replaced by plant personnel. Discussions with Maintenance personnel noted that the parts were replaced when the valve was received. During normal operation, pressure sensing instrumentation is installed at that connection. The inspector noted that no other actions were taken by the licensee regarding determination of the source of the. excessive dirt and rust deposits in the disc and seat are During plant cooldown for the planned mini-outage to remove the BV-1 and BV-2 common cuntrol room wall, it was identified by plant operators that RV-RC-551A was leaking at its flange connection. The licensee planned to replace the gasket and retorque the valve, however, the mechanics noticed three cracked tack welds on the safety valve. The cracked tack ,

welds were on an internal sleeve which is provided to guide the valve i stem. The licensee subsequent _1y decided to remove the valve and replace it with the spare RCS safety v'alve (previously installed as RV-RC-551C).

Inspection of the spare valve identified that one cracked tack weld ex- 1 isted on its internal guide sleeve. An Engineering Memorandum (EM) was generated and dispositioned and stated that the valve was acceptable for use. Following installation of the spare valve, it was not able to main-tain pressure and leaked through to the pressurizer relief tank. The licensee then generated another EM (No. 62380) for the safety valve (RV- l RC-551A) to determine whether the valve could be used with its- 3 cracked i tack welds on the sleeve. Following vendor and engineering evaluation, it was determined that it could be used and that the operation of the i valve would be unaffecte The original safety valve was then re- l installe '

The inspector _ identified several concern First, corrective action or further evaluation was not initiated regarding excessive. dirt and rust deposits found on RV-RC-551 Second, the cause and future corrective j actions for the cracked tack welds and their potential significance has-not yet been addressed. Third, problems remain with the spare safety valves, since the valves could not hold pressure when installed. 'Pending resolution of the above issues and the addressal of potential generic implications, item 86-18-01 remains unresolved, i

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3.3 (Closed) Unresolved Item (86-18-04): Implementation of new procedures for Design Change Package (DCP) handling and development of a tracking system for all DCP open items. The licensee developed a Site Adminis-trative Procedure (SAP-45), Design Change Control. SAP-45 defines, at a departmental level, the responsibility, requirements and guidelines for the initiation, control and documentation of plant design change The procedure also specifies responsibilities for tracking and resolution of DCP open item SAP-45 was approved for use on April 13, 1987. The proper implementation of the new DCP procedure will be reviewed during subsequent routine inspections. Action on this item is complete and UNR 86-18-04 is close . Plant Operations 4.1 General Inspection tours of the follcwing accessible plant areas.were conducted during both day and night shifts with respect to Technical Specification (TS) compliance, housekeeping and cleanliness, fire protection, radiation control, physical security / plant protection and operational / maintenance administrative control Control Room -- Safeguards Area

-- Auxiliary Building -- Service Building

-- Switchgear Area -- Diesel Generator Buildings

-- Access Control Points -- Containment

-- Fence Line (Protected Area) -- Yards Area

-- Turbine Building -- Intake Structure  !

l In addition, the inspector attended an Offsite Review Committee (0RC) {

meeting on May 5, 1987. TS 6.5.2 requirements for member attendance were met. The meeting agenda items observed by the inspector included review of previous ORC meeting minutes, station operating status, Maintenance and Construction Subcommittee report review and ORC open items. The

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meeting was generally characterized by frank discussions and questionin Dissenting opinions were recognized and documented for resolution. The inspector had no concerns at this tim .2 Operations During the course of the inspection, discussions were conducted with operators concerning knowledge of recent changes to procedures, facility configuration and plant conditions. During plant tours, logs and records were reviewed to determine if entries were properly made, and that equip-ment status / deficiencies were identified and communicated. These records included operating logs, turnover sheets, tagout and jumper logs, process computer printouts, unit off normal reports and draft incident report The inspector verified adherence to approved procedures for ongoing

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activities observe Shift turnovers were witnessed and staffing re-quirements confirme Inspector comments or questions resulting from these reviews were generally resolved by licensee personne . During a routine control room tour with the unit in mode 5, the inspector noted that several figures were lying on the benchboard section "B" of the main control board. Four of the eight drawings were labeled " Control Room Controlled Documents" while the remaining figures were labeled " Uncontrolled". The inspector questioned the plant operators concerning the use l of the uncontrolled documents. Some of the operators were not l aware of the purpose of the figures being on the benchboard, !

the associated source document, or in some cases, what the ;

figures were plotting. The inspector questioned the licensee '

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on the use of these uncontrolled documents. Licensee manage-ment stated that operators have been instructed that only con-trolled documents are to be used, and the uncontrolled docu-ments were immediately removed from the control roo i Operating Manual (0M) Chapter 48, Conduct of Operations, specifies that procedures, tables and/or prints may be posted in plant areas for local usage. An index for these documents is included in OM Chapter 48 (0perator Aids). The controlled drawings on the benchboard were listed in the Operator Aids Index. The licensee stated that a review of in plant operator aids is currently being performed and will be compared to the index. This review will include the upgrade and controlling of useful documents not currently in the operator aids index (if applicable). This event appears to be an isolated occur-rence regarding misuse of uncontrolled documentation for which immediate and appropriate corrective actions were implemente The inspector will continue to monitor the'use of local docu-mentation during future inspections. There are no further concerns at this tim . Plant Shutdown Activities The inspector conducted backshift inspections to observe operational activities to bring the unit to cold shutdown be-ginning April 24, 1987. Technical Specification (TS) require-ments related to minimum shutdown margin, power distribution, and pressure / temperature limits for the reactor coolant system and pressurizer were independently verified. The initiation of the Overpressure Protection System, as required by TS 3.4.9.3 was also observed. The inspector noted operator at-tention to plant procedures and logs throughout shutdown evolutions. It was also noted that independent reviews for various TS requirements were performed by the Shift Technical Advisor. No concerns were identifie I

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I 4.2.3 Turbine / Reactor Trip Due to Malfunction in the EHC System On May 31, 1987, the licensee noticed that high pressure oil systems associated with the main turbine generator electrical hydraulic control (EHC) system was running below normal pres-sure. The licensee's troubleshooting of the major components of this system determined that the cause of the low hydraulic oil pressure was that a partially opened intercept valve al-lowed part of the oil to be dumped back to the main lube oil sum The position of the intercept valve is controlled by auto-stop control oil system (control oil). This led the licensee to look at the auto-stop oil system to see why the intercept valve was partially ope The licensee's' investigation determined that the control oil system was running below normal pressure at approximately 80 psig. Normal control oil pressure is between 100 and 110 psi The licensee suspected that a cup valve in the control oil system was allowing oil to be bled by thus causing the inter-cept valve to be in a partial positio l On June 1, the licensee decided to perform Operations Surveil-lance Test (OST) 1.26.4, Turbine Pedestal Check, to investigate the proble During manipulation of the trip test lever latch, the control oil pressure dropped below 45 psig, the EHC control oil system trip setpoint. Movement of this trip test latch should have not affected control oil pressure. This drop of control oil pressure below the trip setpoints caused a turbine trip which resulted in a reactor trip from 96% power. Subse-quent investigation of the EHC system indicated that the trip valve cup was not fully seated. On June 2, 1987, the licensee completed troubleshooting and repair of the trip valve cu Licensee investigation of the EHC system also determined that the trip latch test lever, when it was in the test position, also allowed oil to bleed off from the system. The~ combination of the cup valve being not fully seated and the trip latch allowing oil to be bled by allowed the pressure in the system to drop below the setpoint, thus giving the turbine / reactor tri Because there are no spare parts on site, the licensee ha elected to defer the repair on the trip lever mechanism until the next refueling outage. Because of the leakage past the trip lever, the licensee cannot perform OST 1.26.4 until-the unit is off the lin During the reactor trip, all safety systems functioned as re-quire On June 2, the licensee commenced a plant startup and returned the unit to powe The inspector reviewed the licen-

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see's post-trip review documentation, witnessed portions of the licensee's reactor startup and power escalations. The licensee's actions were found to be acceptabl j 4. Unplanned Gaseous Release

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On May 14, 1987, gaseous waste system decay tank GW-TK-1C was I being released using Radioactive Waste Discharge Authorization -

Gas (RWDA-G) No. 0856. During the release, an increase in pressure in the surge tank was observed. The control room operator therefore began filling an empty waste decay tank ,

(GW-TK-1B) to reduce the surge tank pressure. However, during l the filling process of GW-TK-18 the operator noticed a pressure 1 increase in GW-TK-1C (from 28 to 32 psig), and subsequently 1 isolated the surge tank flow path and terminated PWDA-G N l 0856. The evolution resulted in a portion of the surge tank (

contents to be released through GW-TK-1C. The cause of the I unsampled and unplanned gaseous waste discharge was determined- i to be improper system alignmen )

l The feed line to each of the three decay tanks from the surge tank share a common discharge line with the individual sample return lines. The three sample return lines'in turn, share a i common discharge header. During the process of filling GW-TK-1B and opening its associated sample return valve, the operator did not notice that the sample return valve to GW-TK-1C was previously left opened. Consequently, a surge tank flow path to both GW-TK-18 and GW-TK-1C was provided via the associated sample return line Immediate corrective actions taken by the licensee included 3 cautioning the operator involved on attention to detail during I plant evolutions and requesting a sample of the surge tank contents. The sample showed that a slight but insignificant increase in the activity was determined to have occurred for approximately four minutes. No technical specification limits j were exceede The licensee also initiated Operating Manual 1 Deficiency Report (0MDR) to provide a step in procedure 1.19.4.E, Decay Ta~nk Discharge, to ensure that the respective sample return valve on the tank to be discharged is shu Procedure 1.19.4.B. System Running Procedure, provides in-structions on filling and isolating the decay tank. Procedure 1.19.4.8 specifies that the operator close the associated j sample return valve when isolating the decay tank. Failure 1 to shut the 1C sample return valve is considered a violation i of procedure 1.19.4.8 (50-334/87-07-01). J 4.3 Plant Security / Physical Protection Impleme<;tation of the Physical Security Plan was observed in various plant areas regard to the following:

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Protected area barriers were not degraded;

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Isolation zones were clear;

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Persons and packages were checked prior to allowing entry into the Protected Area; I

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Vehicles were properly searched and vehicle access to the Protected 1 Area was in accordance with approved procedures;

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Security access controls to Vital Areas were being maintained and that persons in Vital Areas were properly authorize Security posts were adequately staffed and equipped, security per-sonnel were alert and knowledgeable regarding position requirements, j and that written procedures were available; and i

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Adequate lighting was maintaine Physical Security Event Reports No. 87-03 thru 87-07 were submitted to the NRC Region I Office during this inspection period. Both specialists and resident inspectors conducted followup inspections for the event Corrective actions were promptly taken by the licensee. All NRC concerns have been adequately addressed. The inspector had no further questions at this tim .4 Radiation Controls Posting and control of radiation and high radiation areas were inspecte Radiation Work Permit compliance and use of personnel munitoring devices were checke Conditions of step-off pads, disposal of protective clothing, cleanliness of work areas, radiation control job coverage, area monitor operability and calibration (portable and permanent) and person-nel frisking were cbserved on a sampling basi No discrepancies were identifie .5 Plant Housekeeping and Fire Protection Plant housekeeping conditions including general cleanliness conditions and control and storage of flammable material and other potential safety hazards were observed in various areas during plant tours. Maintenance of fire barriers, fire barrier penetrations, and verification of posted fire watches in these areas were also observe The BV-1 and BV-2 combined control room share separate emergency venti-lation equipment to meet specific technical specifications. The licensee determined that electrical cabling for BV-2 control room emergency ven-tilation system (CREVS) equipment are contained in areas which do not currently meet all fire protection requirement This is due to in-j

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operable fire protection system components and/or open penetration Since BV-2 CREVS equipment is required to be operable per the BV-1 plant Technical Specifications, the licensee's approach to support BV-1 plant startup was to place fire watches in the affected BV-2 areas in accord-ance with BV-1 TS requirement The licensee provided the fire watch members with fire watch logs, which contained information regarding the type of fire watch required for the specific areas, locations / drawings and a Duties and Instructions summary. Specific instructions are pro-vided to the fire watch members regarding the length of a tour of the fire zone (if applicable). Each affected fire zone'is assigned a sepa-rate fire watch. The inspector verified that the members of the fire watch crew were trained to the licensee's Fire Fighting Training Progra The fire watch members are Security Force personnel. The licensee plans to routinely verify the proper performance of the fire watches. The inspector will also monitor these activitie There are no further con- )

cerns at this tim . Maintenance Activities s j

s 3 i The inspector observed / reviewed various maintenance and problem identification l activities for compliance with requirements and applicable codes and standards, QA/QC involvement, safety tags, equipment alignment /use of jumpers, personnel I qualifications, radiological controls, fire protection, retest and report-abilit The following activities were reviewed:

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MWR 875129, Arc Strike on Recirculation Spray Line

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MWR 875984, Thermal Overload Replacement for M0V-SI-8608

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DCP 611, Unit 1/ Unit 2 Control Room Emergency Air System

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DCP 766, Removal of Wall Separating BV-1 and BV-2 Control Rooms Design Change Package (DCP) 766 maintenance related activites were observed by the inspector. DCP 766 defined the removal of the temporary wall separat-ing the BV-1 and BV-2 control rooms. The design change was to remove the temporary wall and install a new partition that would be largely transparen The design of the new partition incorporates such features and considerations as minimization of noise distractions and activity levels. The inspector routinely observed various phases of the wall removal and partition installa-tion. Implementation of security requirements related to the control room were verified. It was noted that during the wall removal, plastic sheets with taped seams were used in both control rooms to prevent dust from settling'

on/inside control room equipment. The inspector questioned whether fire load analyses for the control room were affected/ updated. The licensee stated that the fire hazard analysis summary sheets for both BV-1 and BV-2 will be updated to reflect the additional fire load in the control roo These changes will-be updated.in a future revision to the Updated Fire Protection Appendix R;; ^

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I DCP 611 rodified the BV-1 control room emergency bottled air pressurization  !

(CREBAP) system to accommodate the addition of the BV-2 area to the control l roo The compressed air storage capacity of the CREBAP system was increased

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by the addition of five new compressed air storage tanks to double the system's

previous capacity. A larger air compressor was also installed, VS-C-2, to increase the charging capability of-the existing system. Control circuitry for existing BV-1 CREBAP system components were modified to' incorporate sig-nals for BV-2, and the BV-2 components will receive signals from. 3V-1. iThes inspector reviewed several portions of system installation and testing. DCP, 611 required changes to the plant Technical

Specifications-(TEs). The licen-see submitted Proposed Operating' License Change' Request No. 126 on October 9, 1986, to NRC Licensing. The change request reflected the increased capacity of both the .CREBAP and the Control Room' Emergency Ventilation Systems. Fol- 3 lowing discussions among licensee and NRC Licensing personnel, and several  !

supplements to the initial submittal', NRC Licensing issued Amendment No. 109 1

to the AV-1 TSs on May 20, 1987. No deficiexies were noted. The inspectors {

s will verify implementation of the new TS requirements during a future inspec- 1 tio . Surveillance Testing

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Theinspectorwitnessed/reviewedselecteds$rveillancetest[t'odetermine '

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whether properly approved procedures were in use, details were adequate, test 1 instrumentation was properly calibrated arid used, technical specifications  ;

were satisfied, testing was performed by qualified personnel and test results '

satisfied acceptance criteria or were properly dispositioned. The following ';

nurveillance testing activities were reviewed: t 3 DST 1.1.13;ChannelCheckofGroupDemandCountersWithinaBank/and'

Overlay Verification '

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OST 1.11.16, Leakage Testing RCS Pressure, Isolation Valves '

OST 1.12.4, Containment Pressure Check for, Air Inleakage

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OST 1.13.6, 2B Recirculation Pump (1RS-P-2B) Dry. Test

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OST 1.24.8, Motor-Driven AFW Pumh Check Valve's and' Flow' Test

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OST 1.47.1, Containment dir Lock Test i

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OST 1. 9.2, Shutdown Margin Calculation (Plant Shutdown)'-

Between >May 24 and May 27, OST 1.24.8, Motor Driven . Auxiliary Feed Pump Check Valves and Flow Test, was performed several time The results.of.the test indicated.that the differential pressure for auxiliary feedwater (AFW) pump 3A was outside the band of acceptable values. The acceptance criteria for differential pressure is based upqn ASME Section XI requirements. An engi-neering memorandum (EM) No. 78251 was generated to evaluate pump performanc The EM stated th'at theJpump still meets the UFSARy design requirements. How-

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ever, per A',ME Section XI, the upper limit values are exceeded. Discussions were held with Inservice Testing Group (IST) personnel, who indicated that upper limit differential pressure criteria will be investigated for revision in accordance with ASME Section XI requirements. The inspector independently verified that UFSAR design requirements have been satisfied. IST personnel are currently 'nvestigating previous OST results to determine the cause of the change in pump performance and to define the required corrective action Pending licensee resolution of current AFW Pump 3A performance characteristics, this is an Unresolved Item (50-334/87-07-02).

7. ESF Actuations

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7.1 On March 25, 1987, during performance of monthly OST 1.24.4, Turbine !

Driven Auxiliary Feedwater (AFW) Pump Test, the "B" motor driven AFW pump (FW-P-38) was automatically started while the plant was at 100% powe The surveillance procedure directed Operations personnel to lift the leads for both motor driven AFW pump FW-P-3A and FW-P-3B timer output terminals to prevent an automatic start of either pump. The motor driven AFW pumps received an automatic start signal by the failure of the tur-bine driven AFW pump to develop the discharge pressure of 500 psia in ,

10 seconds after the turbine driven AFW pump receives a start signa !

Since the OST brings the turbine driven AFW pump up to speed in a slow, i controlled manner, sufficient discharge pressure is not developed in the i time allowed for emergency starts. Therefore, to prevent automatic start ng of the motor driven AFW pumps, the automatic start leads from )

the timer output terminals are lifte As the operator began to recon-nect the terminal lead to restore the system to its normal configuration following the test run, the screw for the terminal point was droppe The operator quickly reacted and attempted to retrieve it. This action caused the terminal lead wire he was holding to brush an adjacent ter-minal, thereby creating a short circuit, which in turn, closed the outpu contact for the start timer for FW-P-38. The pump was run for approxi-mately one-half minute before being shutdow This event was reported to NRC via ENS phon ,

To prevent recurrence of this event, the licensee required all licensed shift personnel to review the asuciated Licensee Event Report (No. 87-06) as a reminder of the importance of exercising caution whenever per-forming work in a confined spst.e with hot or energized leads, and the procedure was revised to delete the procedure requirement to disconnect the timer leads to eliminate the possibility of similar error The inspector had no further questicns at this tim . 2 On April 28, 1987, with the plant in Mode 5', the control room received a high-high alarm on the Auxiliary Building "B" Ventilation Exhaust Radiation Monitor (RM-VS-102B). The high-high setpoint initiated an automatic flow diversion of Auxiliary Building exhaust air through the main filter banks and then through the elevated release on top of the containment building. The automatic ESF actuation was reported to NRC via enc %ne as required by 10 CFR 50.72 (four hour non-emergency re-

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port). The high-high alarm was initiated due to a pressurizer sampling evolution. The purging of the high activity sample through the sample sink plus relatively high background activity levels caused the radiation monitor to increase above the high-high setpoint. The Radiological Con-trol department subsequently initiated an abnormal release record to determine that the resulting release was well within TS allowable limit Corrective action for this event is currently being developed by the licensee, and will be included in the Licensee Event Repor On May 1, 1987, with the plant in the same operational configuration, another high-high alarm on RM-VS-102B was actuated and resulted in an-other ESF occurrence. The alarm cleared within two minutes. A similar sampling evolution was being performed, however, this time leakage past a sample system valve to the sample sink caused the automatic flow diversio The licensee initiated short term corrective action which included: in-formal instructions to plant chemists to help recognize the impact on plant operations while performing sampling evolutions and to ensure that the control room is notified prior to performing sampling on related systems. Additional concerns have been identified as needing resolution:

(1) administrative controls which could prevent similar occurrences during sampling evolutions are not currently formalized in plant proce-dures; (2) sampling purge flow paths do not normally include recircula-tion back to the source as opposed to routing the sample through the sample sink where it is more likely to initiate ESF actuation Pending resolution of the above concerns, this is Unresolved Item (50-334/87-07-03).

7.3 The April 28th ESF actuation was reported by the licensee in accordance with 10 CFR 50.72 requirements. However, the May 1st event was no Operations personnel did not report the second ESF actuation because they felt that the actuation was a part of the pre planned sequence during reactor operation and therefore not roportabl The inspector reviewed various licensing documents and determined that during normal operation of the plant, the exhaust of the ventilation fan does not go through the filters and elevated release path. The ESF actuation causes realignment and is initiated upon either a containment isolation phase A signal or a high-high radiation signal from monitors in the ventilation system exhaust. Therefore, this event should have ,

been reported even if the potential for the ESF actuation was recognized while taking the high activity sample. This concern was brought to appropriate levels of licensee management on several occasions, but not reported until May 12, 1987. The licensee stated that an investigation is currently ongoing to determine the reportability requirements of this and similar events. The licensee also plans to provide guidance for Operations shift and Chemistry personnel to prevent additional automatic actuations. For the interim, a night order was placed in the Control l

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Room log which specifies that any further similar ESF actuations be re-ported per the requirements of 10 CFR 50.72. Failure to report.this event as required by 10 CFR 50.72 is an apparent Violation of NRC regu-lations (50-334/87-07-04).

7.4 On May 25, Procedure 1.44A.4.L, I-solating the Control Room Ventilation System, was entered to place the control room ventilation system in the recirculation mode of operation in accordance with Technical Specifica-tion 3.3.3.7 Action Statement requirements due to the inoperability of the chlorine detection system. To automatically close the two parallel inlet and two parallel outlet dampers for the ventilation system, the control room emergency air safety actuation pushbutton was actuated (wit the emergency pressurization air bottles isolated). This action auto-matically closed the inlet / outlet dampers and initiated the 60 minute control room pressurization timers, which start the control room emer-gency suppply fans. . After initiating the realignment, and subsequent de-energization of the motor operated dampers, both the train "A" and train "B" reset pushbuttons-were depressed to restore the system and actuation logic to normal. However, train "B" would not reset. Plant

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personnel were not able to reset train "B" until about 15 minutes after the. timer had timed out, which caused an automatic start of Unit 2' con-trol room emergency supply fan, 2HVC-FN-241B. This was considered an unplanned ESF actuation and reportd to the NRC via ENS.-

The licensee speculated that loose internal wiring of thei"B" reset-pushbutton was the cause of the event. Maintenance Work Request (MWR)

No. 870797 was initiated to replace the switch. Maintenance has not been performed on the switch, but is currently in the scheduling phase of the MWR proces The MWR specifies that procedure-1.44A.4.L be performed to verify the reset pushbutton' repai The inspector will review com-pletion of the maintenance and the Licensee Event Report when submitted to the NR The inspector had no further questions at this tim . Non-Conservative OPPS Setpoints On May 27, 1987, the licensee determined that the setpoint for the Overpres-sure Protection System (OPPS) valves were non-conservative. Two of the.three spower operated relief valves (PORVs) serve as the OPPS valves; protection is accomplished through the addition of a low pressure setpoint to their control logic, On May 26, the trip actuation setpoint was increased to the high side of the voltage tolerance band (1.467 +/- 0.020 vdc). The Operations Depart-ment requested that this action be taken since low flow conditions on the -i reactor coolant pump No.1 seal leak-off lines were being experienced when trying to start and run the pumps.- An increase in. reactor coolant system pressure was expected to minimize the problem. The setpoint-adjustments were performed per maintenance surveillance procedures.(MSPs) 6.68 and 6.69,:Reac-tor Overpressurization PORV Setpoint Functional Test. 'On May 27, the' licensee

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performed a calculation of the setpoint to provide Operations personnel a'more-precise setpoint value, which yielded.a value.of 364 psig. Technical.Speci-

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fication (TS) 3.4.9.3, OPPS, requires that two.PORVs be operable with'a nomi .

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nal trip setpoint of less than or equal to 350 psig. Upon receipt of the calculation results, the licensee immediately declared both trains of OPPS inoperable, and initiated action to restore the setpoints to less than or equal to 350 psig. The TS Action Statement requires that with both OPPS l trains inoperable, the reactor coolant system shall be depressurized and '

vented within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The setpoints were restored to less than 350 psig !

within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and the valves declared operable. Further licensee corrective action included the generation of procedures change requests to change the trip setpoints in MSPs 6.68 and 6.69 to 1.466 +0.000, -0.010 vde, and to note that 1.466 is the TS maximum limit. The licensee is performing a safety significance determination of the event through the Unit Off-Normal Report proces The licensee stated that the cause of the setpoints being adjusted in the non-conservative direction was due to a misinterpretation of TSs. Specifically, the licensee interpreted " nominal" to mean the setpoint value excluding module j tolerance, i.e. actual setpoint would be greater than 350 psig if within the i instrumentation tolerance band. However, following research of the NRC safety .

evaluation report (SER) for the OPPS, the licensee determined that the set- )

point, including instrumentation tolerance, must be 350 psig or less. The inspector questioned the licensee if there were similar interpretation dis-crepancies between SER assumptions and TS implementation regarding other ESF or Reactor Protection System function The licensee has not yet pursued the resolution of the inspector's concern. Pending licensee resolution of this concern and investigation of the safety significance of the non-conservative OPPS valve setpoints, this is Unresolved Item (50-334/87-07-05). Inoffice Review of Licensee Event Reports (LERs)

The inspector reviewed LERs submitted to the NRC Region I Office to verify that the details of the event were clearly reported, including the accuracy of the description of cause and adequacy of corrective action. The inspector determined whether further information was required from the licensee, whether generic implications were indicated, and whether the event warranted onsite followu The following LERs were reviewed:

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LER 87-06, Inadvertent Motor Driven Auxiliary Feedwater Pump Start (Detail 7.1)

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LER 87-07, Inadvertent ESF Actuation (Detailed in NRC Inspection Report 50-334/87-06)

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LER 87-08, Operation Prohibited by Technical Specifications (detailed below)

LER 87-08 On April 22, 1987, the plant was operating at 100% power when one of the Con-trol Room Emergency f,ottled Air Pressurization System (CREBAPS) bottles was out of service for pressure instrumentation calibration. This placed the

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unit in the Action Statement requirements of Technical Specification (TS) 3.7.7.1, Control Room Emergency Habitability Systems. Additionally, the tem-perature control air compressor (TCAC), VS-C-1A was out of service since March 1 11, 1987, to repair a burnt-out motor. Two TCACs are used to provide sealing air for the control room inlet and outlet air dampers for conditions requiring control room isolation. On April 22, the redundant TCAC, VS-C-1B,' tripped due to an overheated motor. Subsequent automatic and manual restart attempts were unsuccessful. During licensee investigations to return the TCAC back to service, a portable fan was provided to cool the tripped compressor moto The failure of both TCACs rendered the CREBAPS inoperable. Consequently, both the CREBAPS and CREVS were out of service, and the plant was in a condition prohibited by the limiting condition for operations for TS 3.7.7.1. An in-itiation of a plant shutdown within one hour was required by TS 3.0.3. How-ever, prior to the initiation of a plant shutdown, the overheated motor had cooled sufficiently and VS-C-1B was started. The unit was therefore not re-quired to meet the provisions of TS 3.0.3. The licensee determined that the

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VS-C-1B motor overheated due to excessive cycling of the compresso The licensee is currently developing a surveillance procedure to verify proper operation of the TCACs. Additionally, an alternate method to pressurize the CREVS damper seals, using portable nitrogen bottles is being developed by the licensee. Implementation of the proposed corrective actions will be followed during. routine resident inspection. The inspector had no further questions ;

at this tim ,

10.' Review of Periodic and Special Reports Upon receipt, periodic and special reports submitted pursuant to Technical 4 Specification 6.9 (Reporting Requirements) were reviewed. The review' assessed whether the reported information was valid,-included the NRC required data and whether results and supporting information were consistent with design predictions and performance specifications. The inspector also ascertained whether any reported information should be classified as an abnormal occur-rence. The following periodic and special reports were reviewed:

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April Monthly Operating Report for Plant Operations from April 1-30, 1987

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1986 Annual Radiological Environmental Report The reports reviewed were found to be acceptabl . Unresolved Items Unresolved items are matters which more information.is-required in order to determine whether they are acceptable items or violations. Unresolved item ,

identified during this inspection are discussed in paragraphs 6, 7.2 and .!

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1 Exit Interview Meetings were held with senior facility management periodically during the course of this inspection to discuss the inspection scope and findings. A summary of inspection findings was further discussed with the licensee at the conclusion of the report perio i

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