IR 05000271/1986015

Insp Rept 50-271/86-15 on 860701-0804.Violation Noted: Failure to Determine Standby Liquid Control Sys Liquid Poison Tank Boron Concentration Following Water Additions to Tank

ML20203N240
Person / Time
Site:	Vermont Yankee
Issue date:	09/15/1986
From:	Elsasser T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20203N234	List: IR 05000271/1986015 ML20203N232 ML20203N237
References
50-271-86-15, NUDOCS 8609230230
Download: ML20203N240 (15)

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U. S. NUCLEAR REGULATORY COMISSION

REGION I

Report N ,

Docket N License No. DPR-28

. . Licensee

Vermont Yankee Nuclear Power Corporation RD 5, Box 169, Ferry Road Brattleboro, Vermont 05301

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Facility: Vermont Yankee Nuclear Power Station l_ Location: Vernon, Vermont i

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Dates: July 1 - August 4, 1986 Inspectors: . William J. Raymond, Senior Resident Inspector Cornelius F. Holden Jr., Senior Resident Inspector, Maine Yankee Thomas B.- Silko, Resident Inspector Donald _R. Hav amp,-Project Engineer

John A. Schu her, Rea r ngineer Approved by: M f/1!Js Thomas C. Elsasser y f, Reactor Projects Section 3C Date Inspection Summary: Inspection on July 1 - August 4,1986 (Report No. 50-271/86-15)

Areas Inspected: Routine, unannounced inspection on day time and backshifts.by

resident and region-based inspectors of

actions on previous inspection findings; plant startup operations, full power operations including traversing in-core probe (TIP) withdrawal in the reactor b'uilding, plant physical security, surveillance l testing, maintenance activities, and licensee response to selected safety issues (TI 2515/77). The inspection involved 198 hours0.00229 days <br />0.055 hours <br />3.27381e-4 weeks <br />7.5339e-5 months <br />.

{ Results: One violation was identified in 7 areas inspecte The violation con-cerned the failure to satisfy the Technical Specification surveillance requirements

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of the standby liquid control (SLC) system (Section 6.3). Further licensee evalu-ation-of the missed SLC surveillance appears warranted to determine reportability under 10 CFR 50.73. The licensee continues to evaluate the flow oscillations in

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the recirculation loop flows (Section 6.7). Long term corrective actions to pre-vent a recurrence of the TIP withdrawal incident are warranted (Section 6.4).

Further NRC review is required to evaluate the adequacy of the licensee's post i maintenance testing program (Section 5.1).

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DETAILS

Persons Contacted

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I Interviews and discussions were conducted with members of the licensee staff and management during the report period to obtain information pertinent to the areas inspected. Inspection findings were discussed periodically with the management and supervisory personnel listed belo Mr. P. Donnelly, Maintenance Superintendent Mr. G. Johnson, Operations Supervisor Mr. J. Pelletier, Plant Manager Mr. S. Skibnowski, Plant Chemist

~Mr. S. Wender, I&C Engineer Summary of Facility Activities

[ The plant completed a recirculation pipe replacement and; refueling outage.

Reactor startup activities were in progress at the beginning of the inspection perio During the startup on June 30, 1986, a scram occurred due to spurious hi-hi intermediate range monitor (IRM) signals. The scram is discussed in Section 5.1 belo Following replacement of the IRMs, the reactor was critical at 5

51 a.m. on July 2, 1986. The plant reached 100% power on July 13, 1986 and remained at

that power level for the remainder of the inspection period.

1 A meeting between the licensee and the NRC Region I Safeguards staff was held ,

on July'21, 1986, at the NRC Region I office to discuss projected security

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plan change On July 29, 1986, a meeting between the licensee and the Region.

i I staff was conducted to discuss the identification of material deficiencies

during procurement, installation and testing of system components.

i' On July 23-24, 1986 a' Region I inspector conducted a. followup review of the commitments made by the licensee during a management meeting on July 3,1985 i in the area of emergency preparednes The results of that inspection are l provided in NRC Region I Report 50-27/86-16. During the week of July 28, 1986, i

NRR personnel conducted a second audit of the detailed control room design review completed by Vermont Yankee.

' Status of Previous Inspection Findings

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3.1 (Closed) Unresolved Item 85-10-01: Fuel Oil Storage Tank Drain Lin The licensee completed actions per MR 86-498 by May 31, 1986 to address all corrective actions required to stop further corrosion and to justify acceptance of the remaining pipe wall thickness. The licensee also com-

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pleted calculations to demonstrate that the remaining pipe wall thickness was adequate to meet anticipated normal and transient stresses on the

lines. These calculations were previously reviewed by the inspector and l

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.- 3 found acceptable. The inspector toured the area on July 30, 1986 to review the corrective actions and identified no inadequacies. This item is close .2 (Closed) Violation 85-36-01: Refuel Floor ARM Calibration. A violation was issued for the failure to properly calibrate the east and west re-fuel floor area radiation monitors (ARM) according to procedure OP 4511 during two consecutive quarterly calibrations. The licensee's corrective actions, documented in letter FVY 86-10 dated February 5,1986, were re-viewed and found acceptabl The inspector reviewed procedure, OP 4511, Revision 14, and found the revised procedural instructions to be accept-able. Calibration data sheet, VYOPF 4511.08, was revised to include a provision for recording the alarm setpoints. The revised data sheet will aid supervisory personnel during their review of completed test result A new Co-60 ARM calibrator was received on sit The source strength of the new calibrator was sufficient to produce radiation fields of ac-ceptable magnitud Completed ARM calibration data sheets were reviewed by the inspector and found acceptabl Licensee review of other calibration procedures identified the need to make similar changes to OP 4520. The inspector reviewed the changes to OP 4520 and no inadequacies were identified. This item is close .3 (Closed) Unresolved Item 86-10-08: MDR 86-173 Disposition. The inspector met with the maintenance engineer to review the status of MDR 86-173 concerning slag deposits on 7/8-inch carbon steel stud nuts obtained under Purchase Order 2816 Four additional nuts were buffed at the in-spector's request to verify the slag deposits were readily removed. The inspector reviewed the 4 nuts and noted that the deposits were removed and that no crack indications were present. This item is close .4 (0 pen) Unresolved Item 85-40-09: RHR Pump Wear Ring Monitoring. The in-spector witnessed RHR pump vibration measurements completed on July 2, 1986 by the licensee and a contracto The measurements were taken at two locations on each pump and at various flow rates - 50%, 75% and 100%

of rate The data was sent offsite for reduction and analysis by a consultant. The vibration data (frequency spectra) and preliminary re-sults were reviewed by the inspector with the licensee's Mechanical En-gineer on July 17, 1986. The data indicated no pump operability problems, and showed no adverse trends when compared to the baseline dat The licensee committed to NRC:NRR by letter FVY 86-62 dated July 19, 1986

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to inspect the RHR pump wear rings during the present operating cycle and to replace the rings as necessary. The inspector and the NRR operat-ing reactor project manager met with licensee representatives on July 31, 1986 to review pump design, the proposed inspection methods and re-placement options, and the time required to disassemble and inspect a pum . . , , . . __ - . . - . - - -

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, The inspector noted that'.the present wear rings material have Brinell ( hardness numbers in the range of 218 to 345, which is above the desired range of.215 to 265 to lessen the susceptibility of the material to IGSCC

, attack. The Vermont. Yankee pumps are slightly smaller-(16X18X26 v i 18X24X28 series) than the pumps at the Peach Bottom and Browns Ferry

. ' plants which have.had cracked wear rings. .The inspector noted that the-

preliminary inspection results for..similar pumps at the Pilgrim plant have revealed no cracks in the wear rings for 2 of~the 4 pumps. -Also,

the core spray pumps'at VY contain. integral wear rings, which would not

be subject to the postulated failure mechanis ' The inspector had no further comments on this item at the present tim This item remains open pending NRC review of (i) the VY. replacement pro-posal, (ii) the subsequent vibration monitoring results on the 4 RHR pumps, and (iii) completion of licensee actions to inspect the pump wear rings per FVY 86-62. This matter is redesignated as an unresolved item pending completion of the commitment to NRC:NR .5' (0 pen) Unresolved Item 86-10-02: Condensate Storage Tank (CST) Leakage

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-Monitoring. The inspector reviewed the results of data collected per OP 4196 to monitor the CST for leakage following reactor startup._ The i

inspector also toured the CST tank enclosure area and witnessed the col-1ection and analysis of leakage data.

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The licensee noted additional water drainage from the area below the tank floor based on water collected from the. tank base mat drain pipe. The rate of water leakage from the base mat area varied from 0.0 liters /hr to 1.9 liters /hr since the start of data collection on July 3, 198 The increase in drainage appeared to correlate directly with recent

! rainfall, which suggests that rain water is getting under the tank. The drainage water contains tritium at a concentration that is about the same

as that in CST (3.5 X 10 - 3 uci/ml). The tritium in the drainage water

! is most likely from the residual tritiated water trapped in the sand F beneath the tan The CST tritium values are trending upward as expected i following reactor startup. The tritium concentration in the drainage water is less than the CST levels and apparently trending less than the

[ CST values. The licensee plans to continue weekly analysis of the

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drainage water to verify the expected trend The inspector met with licensee personnel and the plant manager to dis-

! cuss the leakage monitoring program. The following items were discussed:

(i) actions to confirm that the observed drainage is the result of rain  !

[ water, and not additional tank leakage; (ii) actions to identify and

eliminate as practicable the migration path of rain water beneath the  !

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tank - actions were completed at the request of the inspector to re-seal the insulation cover on the manway at the top of the tank; and, (iii) revision of the leakage criteria in OP 4196 as necessary to better reflect expected drainage conditions from the drain pipe. This item will i

remain open pending further NRC review of the above licensee actions, along with those documented in Inspection Report 86-10.

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4.0 Observations of Physical Security

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Selected aspects of plant physical security were reviewed during regular and backshift hours to verify that controls were in accordance with the security plan and approved procedures. This review included.the following security i measures: guard. staffing; verification of physical barrier integrity in the protected and vital areas;' verification that isolation zones were maintained; and implementation of access controls, including identification, authorization, 3 badging, escorting, personnel and vehicle searche The inspector reviewed the compensatory measures taken on July 28, 1986 during the hydro-fracturing activities completed on the new water well. These acti-

-vities involved the placement of vehicles within the isolation zone of the

!- protected area fence in the southwest corner of the sit No inadequacies i

were identified.

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3.0 Review of Startup Activities Plant tours were conducted routinely to review startup activities in progress

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and to verify compliance with regulatory and administrative requirements.

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Inspection emphasis was placed on control room activities and the review of

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startup testing activitie Events that received further review are discussed belo .1 Reactor Scram During Startup-

A plant startup was conducted on June 30, 1986 following a 9-month re-fueling and maintenance outage (reference IR 86-10). The reactor was l taken critical at 9:40 p.m. The inspector witnessed the startup from

the control room and noted that the startup was completed in a cautious, orderly manner.

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As neutron levels entered the intermediate range, plant operators noted that intermediate range monitors (IRMs) B, E and F did not respon Technical specification 3.1 requires two operable IRM's per RPS channel.

. Operators placed RPS B in a manual (1/2) scram. condition. While ranging l IRM D from range 6 to 7, IRM A failed offscale low and IRM D went off-l scale high. Since RPS B was already in trip because of the inoperability of detectors B and F, the plant subsequently tripped at 10:05 p.m. on

, June 30,.1986 when IRM A went offscale high, completing the RPS logic to trip RPS A on hi-hi IRM A. The plant responded as expected following the reactor shutdown and operator actions were appropriate to stabilize the plan In particular, no anomalies were noted in control rod motion during the scram.

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Following the reactor scram, source range monitor (SRM) C would not re-tract fully. An investigation revealed that the cable to SRM C was situ-

ated on the wrong side of a transversing in-core probe (TIP) tube below

! the vessel. When the SRM was withdrawn it became tangled with the TIP l

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tube and the SRM was ben The licensee replaced the SRM prior to plant restar All SRM's performed properly, inc!uding retracting, during the startup on July 2, 198 Troubleshooting of the three IRM channels that failed to respond did not identify any problems within the IRM channels. All six IRM detectors had been replaced during the recent outage. The licensee must wait until the flux of the reactor is sufficiently high (as during a startup) to determine if the detectors are operable. The licensee replaced the de-tectors for IRM channels B, E and F, and conducted a prestartup calibra-tion check. The licensee's investigation also revealed that control wiring from IRM D cabinet was cross connected to IRM A preamplifier cabinet. When IRM D was ranged from scale 6 to 7, the preamplifier as-sociated with IRM D should shift to a higher band preamplifier. Since the control cable was cross-connected to IRM A, the preamplifier associ-ated with IRM A switched. Similarly, when IRM A was ranged up, the preamplifier on IRM D shifted. When IRM A responded offscale high to the shift in preamplifier, it tripped the RPS logic for Channel A. The inspector verified that the miswiring was corrected and that all remain-ing SRMs and IRMs were properly wire A reactor startup-was conducted on July 2,1986 and the reactor was taken critical on 5:52 a.m. The inspector witnessed the startup and verified proper operation of all IRM channel The inspector reviewed the licen-see's evaluation of the scram documented in the Post Trip Report for June 30, 1986 The report was consistent with the requirements of AP 0154, Post Trip Review, Revision 2 dated March 12, 1986, and was accurate in the description of the even The inspector also reviewed LER 86-13 which was found to accurately describe the even Licemee and inspector review of the event determined the following:

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IRMs A anJ D operated correctly in ranges 1 through 6 and, in con-junction with IRM C, were capable of providing the intended safety function in response to operational transient The IRMs were not required to be operable prior to the startup on l June 30, 1986, except for several hours on June 4, 1986 when in-i sequence critical testing was conducted. Protective trips from the l APRM reduced setpoint at 15% of rated full power (FP) were available

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l at that time, in addition to the trip from IRMs A, D and IRM cables were replaced during the outage under MR 86-0267 in con-l junction with EDCR 84-427. Post maintenance continuity and meggar l testing was performed. Functional testing per OP 4301 was alsc completed on May 15, 1986.

I Plant operators noted that IRM detectors B,E and F did not appear i to be responding during the Insequence Critical Checks on June 4, 1986. However, reactor flux was not taken above IRM Range 1 and

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-~ detectors would come on scal Maintenance requests (MRs) 86-1134, 1135 and 1136 were issued to investigate the detector ;

Functional testing of the IRMs per OP 4301 was completed on June 17, 1986. The test only checks the channel circuits from the con-trol room electronics and did not . identify the faulty IRM detector The testing did, however, detect the-interaction between IRM A and f

D, and MR 86-1348 was issued to further investigate that proble The MR was open as of the startup to allow observation of the chan -

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nels as the detectors came on scal Based on the above, it appears that one of the two IRM anomalies, the problem of the cross connection between IRMs A and D could have a been corrected prior to June 30, 1986 had additional followup ac-

, tions been take The licensee determined that plant procedures should be changed to

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enhance detector operability checks. IRM functional, calibration-and surveillance procedures will be changed to require testing of

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the. electronics by inputting signals at the preamplifier This j method would better detect problems such as crossed cables.

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The breakdown voltage on the failed B, E and F detectors were within-the " operable range" specified by the manufacturer, but changes in their values were not noted by the licensee and were indicative of a bad detector. Plant procedures will also be revised to require comparison of detectors breakdown voltages for data taken at the

ti;ne of receipt and during post installation checks. The licensee will further request detector breakdown voltage readings on future

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detector purchase order The licensee noted that all SRM and IRM detectors were installed pre-assembled in the drive tubes from the top of the vessel during

the outage. Information from the NSSS vendor indicated that other facilities have experienced detector problems after_using the same

installation method. The licensee decided not to use this instal-

! lation method until further information on the failure mechanism

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The licensee stated that the failed IRM B, E and F detectors will 4 be returned to the manufacturer for analysis and further evaluation of the failure mechanism.

. This item is unresolved pending completion of the procedural changes noted above, and pending further NRC review of the licensee's program for post maintenance functional testing. The inspector noted that this

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was the second recent example (reference Inspection Report 86-10, para-graph 6.0) of inadequate post-maintenance functional testing (UNR 86-15-

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. 8 5.2 Control Rod 18-31 Plant operators noted on July 5,1986, that control rod 18-31 could not be fully withdrawn from position 46 to 48. No problems were experienced when moving the rod from 0 to 46, but the rod would consistently settle back to position 46 when attempting to pull it.to 48. The licensee de-cided to leave the rod at position 46 pending further investigatio The inspector discussed this item with the Reactor Engineering and Com-puter Supervisor on July 7, 1986. The licensee considers the rod to be fully operable for its safety function because of its ability to insert (scram) was not affected, as demonstrated by scram time testing completed at 25% FP per the technical specifications. The rod scram times met the technical specification limit Additional friction tests were completed on the rod on July 6, 1986 and the data was sent to General Electric for evaluation to diagnose the cause of the proble The inspector noted that the mechanical problems recently experienced with rod 18-31 have no relation to the previousl identified problem with the drive down seals, which cause a slow with-drawal speed from notch 00 to 03. No conditions adverse to plant safety were identified. The results of future licensee / vendor evaluations will be reviewed on a subsequent inspection to determine what impact, if any, the rod problem will have of future Cycle XII operation .3 Recirculation Pump Trips Testing The inspector observed the recirculation pump trip tests conducted on July 10, 1986, as part of the licensee's program to obtain baseline statiblity data on the new recirculation system. The test was conducted satisfactorily and all systems responded properly. The inspector re-viewed the APRM, steam flow, reactor pressure, feedwater flow and reactor level traces for the pump trip tests conducted on July 10 and July 13, 198 All parameters responded as expecte No inadequacies were iden-tified. The licensee's stability test results will be reviewed further on subsequent inspection .0 Inspection Reviews of Operations Plant tours were conducted routinely to observe operating activities in pro-gress and verify compliance with regulatory and administrative requirement Tours of accessible plant areas included the control room, reactor building, cable spreading and switchgear rooms, diesel rooms, turbine building, intake structure and grounds within the protected area. Radiation controls were reviewed to verify access control barriers, postings, and posted radiation levels were appropriate. Plant Housekeeping conditions were verified to be in accordance with the requirements of AP 0042. Shift logs and records were reviewed to determine the status of plant conditions and changes in opera-tional status. Items that received further review are discussed belo _, _ _ _

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6.1 Service Water Pump

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Service water (SW) pump A was repl&ced during the outage as the first of the four SW pumps scheduled to be replaced over three consecutive outages. The replacement was necessary due to normal wear experienced by the pumps. The increased efficiency of the pump caused the.SW motor A to operate at an increased load, which raised the motor winding tem-peratures above the process computer alarm point (250 degrees F). The

licensee removed the remaining 3 SW pumps from service to adjust the

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pumps in an attempt to increase their efficiency. This effort was suc-cessful in lowering the, load and operating current of the SW pump A, but motor winding temperatures continued to peak at 275 degrees F and fluctu-ated with ambient air temperature The SW pumps are driven by 250 HP Westinghouse induction motors with Class B windings. The licensee reviewed the appropriate NEMA Standard (MG1-20.40-Temperature Rise) and' determined the maximum operating tem-perature limit of the windings to be 298 degrees F. The operations staff was instructed to secure the SW pump A if the temperature limit was ex-ceede The inspector verified that the operators understood the temperature limit and reviewed the winding temperatures routinely. The inspector found no inadequacies. Further review of the SW pump performance will be completed on a subsequent inspection.

i 6.2 Potential Reportable Occurrences

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The inspector reviewed the licensee's Potential Report Forms (PR0s) for

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events that occurred during the inspection period to determine if the events should be reported to the NRC. The review was conducted for PR0s 86-44 thru 86-54. Events that received additional review are identified below.

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PRO 86-45 and PRO 86-47 were written for missed standby liquid con-

! trol (SLC) system sodium pentaborate concentration surveillances i following the addition of water to the SLC tan This subject is i

discussed further in Section 6.3 belo PRO 86-51 was written to evaluate an instance in which an auxiliary operator (AO) manipulated control rods. The licensee reviewed the event to determine whether the requirements of 10 CFR 55.9(b) were i satisfied. The inspector discussed the event with the operations I

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supervisor and noted that the A0 in question had been selected for training in the next operator licensing class, and that the rod pulls were conducted to meet the requirements of license trainin The inspector discussed the event with the A0 and verified that the rod manipulations were carried out "under the direction and in the presence of a licensed operator". The inspector concurred with the

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licensee. determination that no violation of 10 CFR 55.9(b) require-

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ments occurred, and therefore, the event was-not reportable. Th inspector has no further questions. No inadequacies were identifie .3 : Failure to CompleteSLC Tank Boron Analysis

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The inspector noted on July 9, 1986, that the' standby liquid-contro (SLC) system liquid poison tank level increased by 3% from the previous time the level was observed. The inspector determined that operators

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't added water to the tank at 5:45 a.m. on July 8, 1986 and requested themistry personnel.to sample the tank per the technical specification requirements. The on-shift chemist did not take the sample at the. time of notification due to a procedural requirement to.first mix the tank

for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. However, the chemist. failed to mention the~ sample request-j during the turnover reviews with the oncoming shift and there was no F formal mechanism in place to record the need to take the sample or com-

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} pletion of the analysis. When the inspector contacted chemistry on July

9, 1986 at 3:00 p.m. to review the analysis results, it was noted that-

the sample had not been taken.

. The chemist promptly sampled the tank and. determined that the concentra-tion of sodium pentaborate (boron) was within acceptable limits as specified by Technical Specification Figure 3.4.1. -The inspector noted i that if it were not for NRC review of the activity, the SLC sample would not have been taken. After discussion of this item with operations and chemistry personnel, the licensee initiated actions to provide a method to document the completion of SLC analysis following. additions to the-

tank.

l During subsequent followup on this area on July 16, 1986 at 2:45 p.m.,

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the inspector noted that the SLC tank level had increased by 2% and that

.a boron concentration analysis sample was not completed. The water addi-tion occurred during the SLC monthly pump operability test which was attempted but not completed on July 15, 1986 due to problems with the i SLC cleanup pump. When the monthly test was stopped, the procedure step l to notify chemistry to sample the tank following the addition of the i water during the test was not reached, and plant operators failed to note

the need to check the tank boron concentration. The sample was taken at about 3:00 p.m. on July 16, 1986, which showed that the boron concen-

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tration was acceptabl ~

The failure to determine the boron concentration in the SLC tank follow-ing the addition of water on July 8 and July 16, 1986 violated the sur-l- veillance requirements of Technical Specification Section 4.4.c (VIO 86-

15-01).

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. PRO 86-45 and 86-47 were written to determine the reportability of the event The licensee concluded that since the SLC samples were eventu- '

ally taken and the concentration was within acceptable limits, the tech-nical specification surveillance was satisfied and this item, therefore, i.

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. 11 was not reportabl The licensee further concluded that since there is no time limit specified in the technical specifications as to when the samples need to be taken, then time limits should be established in the appropriate procedures, OP 4114 or OP 416 The inspector discussed these conclusions with the Engineer Support Supervisor and stated that successful completion of the surveillance item cannot take credit for actions taken as a result of NRC inspection of the activity.- The in-spector stated further that since the missed surveillance was a violation of technical specification requirements, the item was reportable under 10 CFR 50.73. This item is unresolved pending submittal of a licensee event report by the licensee and subsequent review by the NRC (UNR 86-15-03).

6.4 Traversing In-Core Probe (TIP) Withdrawal Into Reactor Building A simulated group 2 isolation signal generated per surveillance test OP 4331 at about 10:00 a.m. on July 22, 1986 resulted in the inadvertent retraction of the No. 2 TIP from its shielded position back into the TIP drive housing. Dose rates on the TIP detector were relatively low since it had been more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> since the probe was last in the reactor core. However, the event was significant since the inadvertent retrac-tion of a highly radioactive probe would create a significant exposure hazard to plant workers in the Reactor Buildin During the surveillance, the isolation signal caused a " withdrawn TIP" command as designated from the TIP control units (TCU). This action is an engineered safeguards feature designed to withdraw the TIPS out of the core, to allow the TIP ball valves to close, ensuring primary con-tainment integrity. Prior to the start of the surveillance, the TIPS were fully retracted into the PIG as required by OP 4331. The position of the TIP detector is monitored by an encoder in the TCU which reads

"0000" when the TIP is in the shielded position. The " withdrawn TIP" command from the TCU gave an immediate signal to the TIP drive unit to drive the detector in the high-reverse direction. Before to the TCU control circuits sensed that the detector was already in the shielded position (which would have generated a "stop drive" command), the drive unit apparently withdrew the detector slightly. The slight withdrawal caused the encoder to turnover from the full in "0000" position to the

"1999" position which indicated the TIP was full-out. The TCU responded to the erroneous position indication to continue the high reverse signal by withdrawing the detector until it was in the drive housing. Although all 3 of the TIP unit logics were being tested simultaneously, only the No. 2 TIP unit experienced the malfunctio The TIP remained unshielded until approximately 11:15 a.m. when I&C per-sonnel noted that TCU No. 2 was in the " withdraw TIP" mode. The techni-cians also noted that the local area radiation monitor read 20-30 mR/h Health physics personnel performed a survey of the area, which showed that No. 2 drive was reading 1 R/hr on contact at the top of the housing; 200 mR/hr on contact with the sides; and 20-30 mR/hr at the rope boundar __ .

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>- 12 By 1:30 p.m., the detector was placed in the shielded position which re-duced the area radiation levels to normal. Inspector review of the RWP used for this activity noted that health physics coverage and practices were acceptable. The inspector noted that there were no active RWPs for work in the contaminated area around the TIP drive units at the time of the incident. Therefore,-the inspector concluded that there were no people in the area at the time of the even Short term corrective actions included tagging open the field drive breaker on distribution panel PP-6A, during periods when the TIPS are not in use (shift supervisor's tag 86-1251). Additionally, caution tags (86-135-1,2,3) were attached to the TCUs which instructed workers that the TCU must be manned at times when the circuit breaker is closed. The inspector reviewed the above tags and interviewed licensee personnel regarding their understanding of the established controls. No inade-quacies were identifie The licensee contacted GE to aid in the evaluation of the root cause and long term corrective actions (i.e., software and/or hardware changes)

that are required to establish a permanent fix. This item is unresolved, pending implementation of long term corrective actions or permanent ad-ministrative.(procedure) changes to prevent recurrence of the event, and subsequent review by the NRC (UNR 86-15-04).

6.5 Safety System Reviews The residual heat removal service water, high pressure coolant injection, core spray, standby gas treatment, diesel generator B lube oil, cooling water and starting air, 120/240 VAC instruments bus, 120/240 vital MG bus, and reactor core isolation cooling (RCIC) systems were reviewed to verify the systems were properly aligned and fully operational in the standby mode. When applicable, the review included: (1) verification that accessible, major flow path valves were correctly positioned; (2) verification that power supplies were properly aligned; and, (3) visual inspection of major components for leakage, proper lubrication, cooling water supply, and general condition. No inadequacies were iden-tifie .6 Review of Inoperable Equipment The actions taken by plant personnel during periods when equipment was inoperable was reviewed to verify technical specification limits were met; alternate surveillance testing was completed satisfactorily; and, equipment return to service upon completion of repairs was proper. The above reviews were completed for the following items: removal of the containment air monitor for maintenance on July 14, 1986; RCIC taken out of service for maintenance on July 15, 1986; service water pump D removed from service for maintenance; loss of closed indication on torus /drywell vacuum breaker I on July 15, 1986; loss of control power to diesel gene-

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rator B exhaust fan-on July 27, 1986; and loss of control power to al-ternate shutdown cooling tower fan on August 1, 198 No inadequacies were identifie .7 Oscillations The licensee informed the inspector on July 31, 1986 that plant operators had noted an apparent small oscillation in the_ recirculation loop A drive flow. The licensee initiated plans to gather plant system data to evaluate the nature and cause of the oscillations. The flow oscillations involve an apparent flow increase of about 0.35 to 0.5 Kgpm over a loop baseline flow rate of about 31 Kgpm. One objective of the evaluation will be to determine whether changes in loop A flow correlates with changes in any other reactor system parameter The inspector noted that the flow oscillations did not present an opera-tional problem, and that all plant parameters remained within established operating and technical specification limits. Although the oscillations do not appear to constitute a condition adverse to safety, further review and followup is warranted to assure the cause and nature of the oscil-lations are understood, and that any applicable remedial measures are taken as necessary. This item is unresolved pending completion of the licensee's evaluation of the flow oscillations and subsequent review by the NRC (UNR 86-15-05).

7.0 Surveillance Testing The inspector reviewed portions of the surveillance tests listed below to verify that testing was performed in accordance with administrative require-ments. The review included consideration of the following: procedures tech-nically adequate; testing performed by qualified personnel; test data demon-strated conformance with technical specification requirements; test data anomalies appropriately resolved; surveillance schedules met; test results reviewed and approved by supervisory personnel; and, proper restoration of systems to servic OPF 4020.16 Annual fire protection system operational performance and capacity check, completed on July 23, 198 OPF 4114.01 SLC pump capacity test (various test results reviewed).

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OPF 4302.01 APRM Functional Test, conducted on July 18, 198 OPF 4511.08 Calibration of Reactor Building Ventilation and Refuel Floor Area Zone Monitors (various test results reviewed).

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OPF 4520.01 Source Calibration of ARM System (various test results re-viewed).

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Additionally, the inspector observed portions'of.the work performed.on th reactor core isolation cooling (RCIC) system as part of the RCIC overspeed test conducted per OP 4121 on July 2, 198 No. inadequacies were identifie .0 Maintenance Activities The maintenance request log was reviewed to determine the scope and nature of work done on safety related equipment. The review confirmed: the_ repair of safety related equipment received priority attention; technical specifica-tion limiting conditions for operation were met while components were out of service; performance of alternate safety related systems was not impaired; and, the maintenance activity did not create an'unreviewed safety questio Maintenance activity associated with the fpilowing was reviewed to verify (where applicable) procedure compliance and cquipment return to service, in-cluding operability testin J

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MR 86-1432 UninterruptablePower;Su,pplyfUPS)A-TrippedWhileCycling RHR Valve '

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MR 86-1476 HPCI-15 - Breaker Trip When Cycling Close .MR 86-1750 Loss of Control Power to Diesel Generator B Exhaust Fa MR 86-1415 RCIC Amphinol Connector Repai MR 86-1474 Main Steam Isolation Valves 86.B&C Failed Closed During Test-in No inadequacies were identifie .0 Licensee Response to Selected Safety IssuesV(TI 2515/77)

During this inspection period, as requested by the NRC's Office of Inspection and Enforcement (IE), region-based inspectors conducted a survey of the lic-

ensee's response to selected safety issues (reference: IE Manual Temporary Instruction 2515/77). The primary purpose of the survey was to determine the

actions that licensees are taking to address a selected sample of safety is-l'

sues. These issues have been identified in IE bulletins, circulars, and in-formation notices and in the Institute of Nuclear Power Operations' (INP0's)

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significant operating event reports (SOERs). This information is needed to determine whether the NRC should take additional action on these items. A secondary purpose of the survey was to determine the actions that licensees are taking in response to INP0's SOERs. It should be noted that INP0's re-commendations are not regulatory requirements and that this survey did not j involve checking responses simply because they are INP0's recommendation Key items that are central to resolving safety concerns were selected for this

, survey.

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The selected safety issues applicable to the survey at Vermont Yankee included the reliability of high pressure coolant injection / reactor core isolation cooling (HPCI/RCIC) systems, and biofouling of cooling water exchangers. Re-garding the reliability of HPCI/RCIC systems, the inspector determined whether procedures have been changed, programs established, and management controls-implemented to improve system reliability. - For biofouling of cooling water heat exchangers, -the inspector determined whether equipment has been installed, surveillances are being performed,-and training and procedures are being pro-vided. The survey results were provided to IE for their review and assessmen No violations or safety concerns were identified as a result of this surve .0 Management Meetings Preliminary inspection findings were discussed with licensee management peri-odically during the inspection. A summary of finuings for the report period was also discussed at the conclusion of the inspection and prior to report issuance.

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