IR 05000271/1988006
| ML20197F296 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 06/01/1988 |
| From: | Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20197F293 | List: |
| References | |
| 50-271-88-06, 50-271-88-6, NUDOCS 8806100269 | |
| Download: ML20197F296 (39) | |
Text
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. , . . - . U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-271/88-06 Docket No.
50-271 License No. DPR-28 Licensee: -Vermont Yankee Nuclear Power Corporation RD 5, Box 169 Brattleboro, Vermont 05301 Facility: Vermont Yankee Nuclear Power Station Inspection At: Vernon, Vermont Inspection Conducted: March 22, 1988 - May 16, 1988 Inspectors: Geoffrey E. Grant, Senior Resident Inspector John B. Macdonald, Resident Inspector Approved By: / M d.
4/// M 4#u 4 Donald R. Haverkamp, Chief ReactorProjectsSectionNo.y Date 3C ' Inspection Summary: Inspection _on March 22, 1988 - May 16, 1988 (Report No. 50-271/88-06) > Areas Inspected: Routine inspection on daytime and backshifts by two resident , inspectors of: actions on previous inspection findings; operational safety; security; plant operations; maintenance and surveillance; engineering suppart; . radiological controls; licensee event reports; licensee response to NRC initiatives; and, periodic reports.
Results: 1.
General Conclusions on Adequacy, Strength or Weakness in Licensee Prog rams
The licensee exhibited technical competence and aggressive analysis in response to a failed fuel indication-(Section 6.1)
Licensee actions during the March 29, 1988 Unusual Event were professional and well coordinated. Activation of the Technical Support Center (TSC), implementation of emergency plan procedures, information turnover from the control room to the TSC and technical control of the event by the TSC were all excellent (Section 6.2).
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_ _ , . , . . , ' Inspection Summary (Continued)
Licensee delay in performing the vendor recommended maintenance on the B Diesel Generator represents a non-conservative approach to safety related equipment maintenance (Section 7.3) A number of recent problems with the surveillance program indicated the need for a comprehensive program review and validation.
Although the individual failures appear minor, they are indicative of a programmatic weakness in the focus and management of the system.
Licensee attention is necessary to address underlying program problems in addition to the already identified corrective actions for individual failures (Section 7.5).
After a troubled start, licensee development of a reliable and comprehen-sive IST Vibration Monitoring Program has shown progress and improvement (Section 8.2).
Licensee implementation of 10 CFR 50.59 requirements in revfew of diesel generator air start system changes was inadequate in that management of the issue was neither timely nor aggressive (Section 8.3).
2.
Violations A violation was identified by the licensee concerning failure te perform a TS 6.5.F required semi-annual inventory of a sealed source.
No notice of violation was issued (Section 9.1).
Another violation was identified by the licensee concerning failure to meet TS 3.9. A.1 requirements for service water system ef fluent radiation monitoring.
No notica of violation was issued (Section 9.2).
3.
New Unresolved Items Identified Licensee performance of a 100*; validation of TS surveillance requirements versus implementing schedules and procedures is unresolved oending com-pletion and inspector review (Section 7.5).
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, . , . . , . - . TABLE OF CONTENTS PAGE 1.
Persons Contacted...,
.................. 2.
Summary of Facility Activities...............
- 3.
Status of Previous Findings (IP 92701)*...........
3.1 (Closed) Unresolved Item 84-12-04: Environmental Qualification (EQ) of Two Way Solenoid Valves Installed in RHR Service Water Pump Motor Bearing Cooling System.
.......... 3.2 (Closed) Follow Item 84-18-04: Review Corrective Revision to Drawing G191175..... 3.3 (Closed) Unresolved Item 87-09-04: Review of Destructive Examination of the RHR Pump Impellers.....................
3.4 (Closed) Fo^ low Item 84-18-03: Submittal of Proposed Change to TS to Reflect the Intent of the Requirements.............
3.5 (Closed) Unresolved Item 88-03-01: Service Water System Effluent Radiation Monitor Inoperability.
............... .... 4.
Operational Safety (IP 71707, 71710,)....... ........
4.1 Plant Operations Review.
............... 4.2 Safety System Review.
................. 4.3 Feedwater Leak Detection System.
........... 4.4 Inoperable Equipment.
............ .... 4.5 Review of Lifted Leads, Jumpers and Mechanical Bypasses........................
4.6 Review of Switching and Tagging Operations.......
4.7 Operational Safety Findings....
......... 5.
Security (IP 71707).............
....... 5.1 Observations of Physical Security.
..........
5.2 Fitness for Duty.
................... 6.
Plant Operations (IP 71707, 93702, 82201, 94703)......
6.1 Failed Fuel Indications................
6.2 Notification of Unusual Event.
............ 7.
Maintenance / Surveillance (IP 71710, 61726, 62703, 61700)..
7.1 Uninterruptible Power Supply (UPS) Inoperability.
... 7.2 Valve CS-11B Failure.
................. 7.3 Diesel Generator Maintenance
......... ... 7.4 Emergency Core Cooling System (.CCS) Logic Testing...
..................... 7.5 Surveillance Program Review.
............. i
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Engineering Support (IP 73051, 73052, 73755)........
8.1 Core Spray Safe End Nozzle Weld Overlays........
8.2 Inservice Testing Vibration Monitoring Program.
.... 8.3 Diesel Generator Operability.
............. 8.4 Scram Discharge Volume Tubing Seismic Qualification..
9.
Radiological Controls (IP 71707)..............
9.1 Sealed Source Survey and Inventory.
.......... 9.2 Service Water Effluent Radiatior. Monitoring..
.... 10.
Licensee Event Reporting (LER) (IP 90712, 92700)......
10.' LER 88-01 Revision 1..................
0.2 LER 88-02 and Supplement 1...............
10.3 LER 88-03.......................
11.
Review of Licensee Response to NRC Initiatives: (IP 92703, 25590)......................
11.1 IE Bulletin 85-03: Request for Additional Information Concerning VYNPC Response.
..... .. 11.2 Wet Cell Storage Battery Adequacy Audit.
....... 11.3 NRC Bulletin 83-01: Defects in Westinghouse Circuit Breakers.......
............. 11.4 Region I Temporary Instruction 87-03, Storage of Transient Equipment in Safety-Related Areas......
12. Review of Periodic and Special Reports (IP 90713)......
13. Management Meetings (IP 30703, 40700)............
Attachment A: Request for additional information Attachment B: Wet cell storage battery adequacy audit
- The NRC Inspection Manual inspection procedure (IP) or temporary instruction (TI) or the Region I temporary instruction (R1 T!) that was used as inspection l
guidance is listed for each applicable report section.
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Persons Contacted Interviews and discussions were conducted with members of the licensee staff and management during the report period to obtain information per-tinent to the areas inspected.
Inspection findings were discussed periodically with the management and supervisory personnel. listed below, Mr. P. Donnelly, Maintenance Superintendent s
- Mr. R. Grippardi, Quality Assurance Supervisor
' Mr. S. Jefferson, Assistant to Plant Superintendent Mr. G. Johnson, Operations Supervisor Mr. R. Lopriore, Maintenance Supervisor Mr. R. Pagodin, Technical Services Superintendent
- Mr. J. Pelletier, Plant Manager Mr. R. Wanczyk, Operations Superintendent Mr. T. Watson, I & C Supervisor
- Attendee at post-inspection exit meeting conducted on May 19, 1988.
2.
Summary of Facility Activities Vermont Yankee Nuclear Power Station (VYNPS) continued full power opera-tions during this period except for pre planned power reductions to accom-plish required surveillances and a rod pattern exchange.
An unplanned power reduction to 96t, full power on March 31, 1988 was necessitated by , declaration of an Unusual Event.
The Unusual Event was secured and full power operation restored on the same day.
n An NRC Region I Operational Performance Assessment team inspection com-pleted a comprehensive review of overall VYNPC operations during the period of April 4-8, 1988 (Inspection Report 88-05).
3.
Status of Previous Inspection Findings 3.1 (Closed) Unresobed Item 84-12-04: Environmental Qualification (EQ) { of Two Way Solenoid Installed in the RHR Service Water Pump Motor ' Bearing Cooling System. During a 1984 EQ program review, the licen- _ see determi.1ed that four two-way ASCO (Automatic Switch Company) nor-
mally oper, type 8211 solenoid valves (SE-70-4A-D) installed in the.
' RHR SW pump motor bearings cooling system did not meet the EQ cri-teria.
Ths licensee replaced the valves under EDCR 84-413 with qualified three-way solenoid valves, but found that the. replacement valves were prone to improper operation due to fouling of the pilot i caused by sediment in the service water.
The original valves were l subsequently reinstalled in accordance wita a revision to the EDCR ! ! . I
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with the intent of upgrading the valves to meet the EQ criteria. The licensee forwarded the valve and environmental data to ASCO via cor-respondence VYV 85-033 dated January 15, 1985.
On January 22, '1985 ASCO responded via correspondence, - ASCO P/N HPX8211C13. ASCO recon-mended replacing the existing plastic disk holder with a stainless steel disk holder and replacing Buna N' resilient parts with ethylene propylene -resilient parts to meet or exceed existing EQ criteria.
The proposed valve internals were replaced as recommended by ASCO un c-MR 86-0F on June 30, 1986.
The inspectors had no further ques.
is item is closed.
3.2 (Closea) Follow Item 84-18-04: Review Corrective Revision to Drawing G191175. A note on revision 33 to Drawing G191175 erroneously stated that the automatic control circuit for the vauuum breaker will over-ride a manual signal when a 0.5-psid between the torus and reactor vessel exists.
Corrective action 84-299 to revision 36 of Drawing . G191175 properly stated that the manual key lock switch will allow the operator to override automatic actuation. The inspectors had no further questions.
This item is closed.
3.3 (Closed) Unresolved Item 87-09-04: Review of Destructive Examination of the RHR Pump Impe11ers.
The NRC issued IEN 86-39, "Failure of RHR Pump Motors and Internals" on May 20, 1986, describing significant damage to a Bingham-Willamette pump.
The failure of the pump was cetermined to have been caused by intergranular stress corrosion cracking (IGSCC) of the pump impeller's wear. rings.
The Bingham-Willamette Type CVIC RHR pumps in service at VYNPS are similar to the type referred to in IEN 86-39. As a result of this information VYNPC performed inspections of the RHR pump impellers at~ VYNPS for wear ring damage. The inspections revealed no indications of IGSCC on the year rings, but both the 8 and D pumos had crack indications on the impeller casting.
Brookhaven National Laboratory was contracted to perform a failure analysis of the B RHR impeller which had a linear indication of approximately five inches located about three-eighths of an inch from the wear ring. The analysis consisted of the follow-ing destructive and non-destructive examination techniques: (1) Visual inspection, photography, optical microscopy (2) Hardness measurements (3) Scanning electron microscopy (SEM) and energy dispersive spectroscopy The examination concluded that the cracking was almost entirely asso-ciated with areas of weld repair. The probable cause of the cracking was hot cracks resulting from large manufacturing weld repairs of the original casting.
The inspectors had no further questions.
This item is closed.
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I 3.4 (Closed) Follow Item 84-18-03: Submittal of proposed change to TS to
- reflect the intent of the requirements to maintain both containment-i isolation and vacuum breaker functionality.
During review of a related operational event, an NRC inspector noted a conflict in the intent of TS 3.7.A.5 and TS 3.7.0 addressing required actions when a i torus vacuum breaker fails open.
The required. actions of TS 3.7. A.5.b dictate that the failed vacuum breaker be locked. open.
Containment isolation would be maintained by the normally closed ] check valve in the line.
Indefinite continued plant operation is-allowed by TS 3.7.D with only-one of two containment isolation valves-inoperable, provided the redundant valve is secured in the isolated position.
Strict reading of the TS requirements would not prohibit- ) indefinite operation with one vacuum breaker failed open. The licen-see has submitted to the NRC proposed change number 134 to the VYNPS TS to place a seven day limit for continued operation with a failed ,
vacuum breaker. The proposed change was submitted via correspondence FVY 86-80 dated August 26, 1986 and is under NRC review. The inspec-tors had no further questions, This item is closed.
3.5 (Closed) Unresolved Item 88-03-01: Service Water System Effluent i Radiation Monitor Inoperability.
This item was opened pending licen-see resolution of the reportability of the underlying event.
The licensee has submitted LER 88-01 Supplement 1 to address this event.
Section 9.2 of this report more fully describes the details.
This item is closed.
, 4.
Operational Safety 4.1 Plant Operations Review The inspector ob:erved plant operations during regular and backshift tours of the following a,eas: Control Room Cable Vault Reactor Building Fence Line (Protected Area) Diesel Generator Rooms Intake Structure Vital Switchgear Room Turbine Building ' Control Room instruments were observed for correlatten between chan-nels, proper functioning, and conformance with Technical Specifica-tions. Alarm conditions in effect and alarms received in the control room were reviewed and discussed with the operators. Operator aware-ness and response 'to these conditions were reviewed. Operators were
found cognizant of board and plant conditions.
Control room and ~
shift manning were compared with T.
ical. Specification require-ments.
Posting and control of radiation, contaminated and high radiation areas were inspected. Use of and compliance with Radiation Work Permits and use of required personnel monitoring devices.were j
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. checked.
Plant housekeeping controls were observed including control ' of flammable and other hazardous materials. During plant tours, logs and records were reviewed to ensure compliance with station proced-ures, to determine if entries were correctly ~ made, and to verify cor-rect communication of equipment status.
These records included var-ious operating logs, turnover sheets, tagout and jumper logs, and .Futential Reportable Occurrence Reports.
Inspections of the control room were performed on weekends and backshifts including March 22, 24, 29-31 and April 1, 7, 11, 12, 25-29 and May 9-13, 1988. Opera-tors and shift supervisors were alert, attentive and responded appro-l priately to annunciators and plant conditions.
4.2 Safety System Review ' The emergency diesel generators, residual heat removal, standby gas treatment, residual heat removal service water, and high pressure coolant injection systems were reviewed to verify proper alignment
and operational status in the standby mode.
The review included verification that (i) accessible major flow path valves were cor-rectly positioned, (ii) power supplies were energized, (iii) lubri-cation and component cooling was proper, and (iv) components were operable based on a visual inspection of equipment for leakage-and general conditions.
No violations or safety concerns were identified.
4.3 Feedwater Leak Detection System Status The inspector reviewed the feedwater leakage' detection system and the ' monthly performance summary provided by the licensee in accordance with VYNPC letter FVY 82-105.
The licensee reported thac, based on the leakage monitoring data for March and April 1988, there were no deviations in excess of 0.10 from the steady state value of normal-ized tnermocouple readings, -and no failures in the 16 thermocouples installed on the four feedwater nozzles.
No inadequacies were iden-tified and the inspector had no further questions in this area.
4.4 Inoperable Equipmeni Actions taken by plant personnel during periods when equipment was inoperable were reviewed to verify: technical specification limits were met; alternate surveillance testing was completed satisfac-
torily; and, equipment return to service upon completion of repairs J was proper.
This review was completed for the following items: March 29, 1988--B uninterruptible power supply (UPS) was declared inoperable due to a blown fuse and failed circuit cards.
Refer to Sections 6.2 and 7.1 for more detail.
March 31,1988--Core spray discharge M0V CS-118 failed during alter-nate testing.
Refer to Sections 6.2 and 7.2 for more detail.
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. - April 8, 1988--Reactor building fire detection panel ' was declared inoperable after it was inadvertently sprayed with water.
The panel was air dried and returned to service on April 12, 1988.
Compensa-tory firewatches were instituted as required by TS and approved procedures.
April 9-10, 1988--B UPS tripped three times as a result of failed circuit cards.
All circuit cards were subsequently replaced with vendor refurbished cards.
April 19 and May 11,1988--Reactor ' core isolation cooling system (RCIC) trip and throttle valve RCIC-1 failed to trip the turbine. On April 19, 1988 the valve cycled properly during troubleshooting. On May 11, 1988 troubleshooting identified valve stem binding caused by contact between the valve stem and actuator shaft coupling and the valve yoke. The valve actuator limit switch was adjusted to maintain a 0.030 inch clearance between the coupling and yoke. Troubleshoot-ing and maintenance were performed under MR 88-0870.
4.5 Review of Lif ted Leads, Jumpers and Mechanical Bypasses Lifted lead and jumper (LL/J) requests and mechanical bypasses were reviewed to verify that controls established by AP 0020 were met, no conflict with the technical specifications were created, the requests were properly approved prior to installation, and a safety evaluation in accordance with 10 CFR 50.59 was prepared if required.
Implemen-tation of the requests was reviewed on a sampling basis.
The following LL/J requests were cancelled on April 4, 1988.
LL/J 87-0046 - LL/J 87-0072 LifJ 87-0132 - LL/J 87-0138 The '.mpers installed by the LL/J requests above were made permanent modi ications to motor operated valve actuator limit switches via ins a11ation and test (I&T) per EDCR 87-408, ECN-1.
4.6 Review of Switching & Tagging Operations The switching and tagging log was reviewed and tagging activities were inspected to verify plant equipment was controlled in accordance with the requirements ' of AP 0140, Vermont Local Control Switching Rules.
1he following switching and tagging orders were reviewed: 88-0362 -- issued and restored on April 19, 1988 to support RCIC-1 trip and snrottle valve cleaning and lubrication.
88-0439 -- issued and restored on May 11, 1988 to support RCIC trip and throttle valve troubleshooting and maintenanc... .-- .. _ _ _ - . . .. . , . . \\ .
. 4.7 Operational Safety Findings , Licensee administrative control of off-normal system configurations. by the use LL/J, mechanical bypass, and switching and tagging' pro-cedures, as reviewed in Sections 4.5 and 4.6, was in compliance with procedural instructions and was consistent with plant safety.
Licen-see efforts to minimize activa lifted leads, jumpers and mechanical bypasses is noteworthy.
5.
Security , 5.1 Observations of Physical Security Selected aspects of plant physical security were reviewed during regular and backshift hours to verify that controls were in accord-ance with the security plan and approved procedures. 'This review included the following security measures: guard staffing; vital and protected area barrier integrity; maintenance of isolation zones; and, implementation of access controls, including authorization, badging, escorting, and searches.
No inadequacies were identified.
5.2 Fitness for Duty Based upon reliable information obtained from a credible source, the licensee determined that probable cause existed to test 21 contractor personnel for drugs.
Four of these contract workers refused to take the test and were immediately dismissed.
The. remaining personnel ' submitted to the test on April 20, 1988.
Seven of these workers tested positive and were dismissed from the site on April 22, 1988.
All of these workers were from Custodis Ecodyne Company who were onsite for annual maintenance on the cooling towers.
All work was outside the protected area on non-safety related equipment.
The workers did not have unescorted access inside the protected area.
The licensee is reviewing the incident to determine if changes to fitness-for-duty procedures are necessary or advisable. The inspec-tor noted no deficiencies.
6.
Plant Operations 6.1 Failed Fuel Indications Radiation levels increased in the offgas system on March 21, 1988.
The inspector interviewed chemistry and reactor engineer personnel, reviewed control room strip charts and radiation monitors, and re-i viewed logs and records to independently assess the increased radia-i tion levels and their significance.
The increased radiation levels were indicative of a small fuel failure.
The reactor had been
operating at full power prior to the event.
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. The daily offgas ~ sample on March 21,1988 'showed a reading of 1270 uC1/sec, which was up from the 430 to 500 uC1/sec range experienced prior to March 21 based on daily samples. Confirmatory samples ver-ified the increase resulted from activity in the offgas system.
This increase coincided with a planned power reduction to 55% full power, control rod pattern exchange, and return to full power on March 19 and 20,1988.
The release rate. remained relatively constant. until. April 4,1988 when it increased to 3700 uCi/sec.
The increase coin-cided with a planned downpower to 84% full power, control rod adjust-ment and - return to full power on April. 3,1988.
The release rate again increased on April 11, 1988 to 5270 uCi/sec following a similar set.of planned evolutions on April 10, 1988._ The release _ rate grad-ually increased to approximately' 6500 uCi/sec by April 30, 1988.
Following ~ a planned power reduction to 52% full power, rod pattern exchange, single rod scram testing and return to full power, the release rate increased to 8240 uCi/sec on May 2,1988.
The -release rate remained relatively constant to the _end of the report period.
Steam jet air ejector and advanced offgas system. guard bed inlet radiation levels have shown corresponding increases.
, The current offgas release' rate of approximately 0._008 Ci/sec (8000 uCi/sec) remains well below the Technical Specification 3.8.K.1 limit
of 0.16 Ci/sec.
Reactor vessel isotopics for dose equivalent I-131 ' remain well below the Technical Specification 3.6.B.1 limit of _1.1 uCi/gm.
Environmental release rates, calculated by the licensee based on the Offsite Dose Calculation Manual (00CM), were only 0.025 > mrem /yr and remained well below Technical Specification 3.8.E.1 limits of 500 mrem /yr (whole body) and 3000 mrem /yr (skin).
Licensee evaluation of the event is ongoing. Measurements taken dur-ing the May 2 control rod pattern exchange to establish the location of the leaking bundle using the offgas flur. tilt monitor has narrowed the suspect bundle to one quarter of the core or less. The licensee preliminary conclusion in conjunction with NSSS vendor support is that one failed fuel pin exists and was possibly caused by random pellet-clad-interaction (PCI; or crud induced localized corrosion (CILC).
The two failed fuel pins discovered during_ the 1987 outage . were caused by CILC (see Inspection Report 87-21). The licensee will j continue to analyze the results of any future changes in release i rates or special testing.
The licensee has exhibited technical competence and aggressive analysis in response to this failed fuel indication.
Interaction with NSSS vendor was initiated at an appropriate time to aid in problem analysis.
Licensee continuing actions will be followed as part of subsequent routine inspection, along with reviews of plant status to confirm continued operation within license requirements.
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. 6.2 Notification of Unusual Event On March 29,1988 at 11:30 a.m., with Vermont. Yankee Nuclear Power Station (VYNPS) at 100% power, the B uninterruptible power supply
(UPS) was declared out of service due to a blown fuse-and failed
circuit cards. The B UPS is a subsystem of the low pressore coolant -
injection (LPCI) system in that it supplies emergency power to the ~ LPCI loop B discharge valves, RHR-258 and RHR-278, in the event of a i loss of offsite power.
Alternate testing of both core spray. (CS) j subsystems, the remaining LPCI and containment cooling subsystems and the standby diesel-generators was immediately performed on March 29, 1988 and daily, thereafter, in accordance with TS 3.5. A.4.
On March 31,1988 at 2:45 p.m. while performing the required daily alternate testing, core spray motor operated valve (MOV). CS-118 failed in the mid-stroke position. The valve was declared inoperable and a reactor shutdown was initiated in accordance with TS 3.5. A.6, which requires that an orderly shutdown be initiated and the reactor be in cold shutdown within 24 hours if the alternate testing criteria are not met.
The licensee declared an Unusual Event as of 2:45 p.m.
on March 31, 1988 in accordance with facility emergency action levels procedure AP 3125, Revision 6, "Emergency Plan Classification and~ Action Level Scheme". Procedure AP 3125 requires an Unusual Event to , be declared wnen the loss of a system function or engineered safety
feature requires a plant shutdown. in accordance with TS limiting conditions for operation. The NRC Operations Center and the appro-priate state and local notifications were made.
The Unusual Event and reactor shutdown were terminated at 7:15 p.m. on March 31, 1988 . after repairs and post maintenance testing of the B UPS were l completed.
Reactor power had been reduced to 96%. On April 2,1988 at 5:40 p.m. repairs and post maintenance testing of CS-11B were satisfactorily completed and the valve was returned to service. The repair of the B UPS and CS-11B are documented in detail in Section 7.1 and 7.2.
The licensee responded appropriately and accurately identified and-j classified the Unusual Event. The resident inspectors were present
in the control room at the time the Unusual Event was declared. On-shift personnel performed in a very controlled and professional man-
ner.
The inspectors observed the manning of the technical support center (TSC) and the shift of control from the control room to the TSC which were both prompt and orderly.
The B UPS return to ser-vice was enhanced by the consultation and expertise of a vendor representative dispatched to the site. The root cause failure deter- " mination and repair of CS-11B were comprehensive and complete.
The inspector had no further questions.
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Maintenance / Surveillance 7.1 Uninterruptible Power Supply (UPS) Inoperability On March 29,1988, at 11:30 a.m., the B UPS was declared inoperable i following the discovery of a blown leg fuse. Initial troubleshooting efforts under MR 88-0731 were somewhat uncontrolled in that, during the period of March 29-31, 1988, many circuit cards were replaced and.
tested unsatisfactorily evidently without utilizing maintenance guid-ance, "Procedure for Troubleshooting the UPS, One Inverter Leg Fuse Cleared."
After regrouping, maintenance personnel implemented the UPS gu, dance procedure and on March 31, 1988 ultimately detected the root cause failure mechanism, a failed capacitor on inverter leg 3.
The failed capacitor, circuit boards and fuses were replaced from existing stores.
The B UPS was post maintenance tested satisfac-torily and was returned to service on March 31, 1988 at 6:03 p.m.
7.2 Valve CS-11B Failure On March 31, 1988, at 2: 45 p.m., core spray motor operated discharge isolation valve CS-11B failed in mid-stroke during testing and was declared inoperable.
The valve was being tested in accordance with i the alternate testing requirement of TS 3.5.A.4, as the B UPS was out of service.
The valve actuator (Limitorque SMB-2) was removed. from the valve and maintenance personnel began a failure mechanism 'and I root cause evaluation.
The evaluation and subsequent repairs were i performed in accordance with the direction.of OP 5220, "Limitorque Operator Inspection", revision 9, under MR 88-0756.
The inspection of the actuator revealed that the motor pinion gear key was missing from the keyway and that the motor pinion gear had rotated 90 degrees from the proper keyway orientation.
The declutch fork was observed to have indications of metal scratching on the surface toward the motor pit: ion gear. The key was found in two pieces within the clutch housing.
Following further analysis of the results of the inspec-tion, maintenance engineers concluded that the root cause of the actuator failure was the improper staking or peening of the motor . pinion gear key to the motor shaft keyway allowing slight free motion , l of the key. During the course of repeated valve testing, the key had migrated approximately 3/4 of its length out of the keyway ultimately striking the underside of the declutch fork. This caused the key to break in two. The larger piece of the key remained functional in the keyway, the smaller piece fell and mixed with the grease. During the March 31, 1988 failed test of CS-118, the key segment that had fallen into the grease became wedged between the motor pinion gear and the ! motor helical gear preventing translation of motor shaft rotation to the worm shaft also causing the motor shaf t to rotate 90 degrees t - - - - - - - - - - - - -
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. without motion of the motor pinion gear. At this point, power to the
valve was secured locally by an auxiliary operator.
The inspectors observed a walkthrough of the failure mechanism and concur with the licensee conclusions. The metal scratching observed on the declutch fork correspond exactly to where the key impacted it and the scarring on the piece of key fit exactly into the scarring on the motor pinion , and m6 tor helical gears.
The licensee inspected and replaced as required the other actuator parts.
The actuator was reinstalled on CS-11B, post maintenance tested satisfactorily and returned to service at 5:40 ~p.m.
on April 2, 1938.
Vendor specialists from MOVATS were -brought on site to acquire baseline operational data.
The licensee inspection and repair of the CS-118 actuator was very well controlled and complete.
The licensee was deliberate in accur-ately identifying the root cause and failure mechanism.
The inspec-tors had no further questions.
7.3 Diesel Generator Maintenance Fairbanks Morse (Colt Industries), the manufacturer of the standby
diesel generators (SDG) at VYNPS, recommends that a major SDG over-haul be performed every refueling outage (12-18 months).
Major inspection items include checking: injection nozzles for operation and opening pressure -- fuel injection timing -- timing chain tension and timing gears -- lower crank strain, crank lead end float and all bearings -- cam shaf t bearing, cams and roller faces and torsional dampers -- condition of engine blower lobes and clearances -- vertical drive coupling and bearings l -- i This preventive maintenance was not performed during the 1987 refuel- ) ! i ing outage on either SDG due to unexpected delays in the removal of asbestos lagging from the SDG rooms. Licensee management elected not to extend tt.e outage to perform the SDG overhauls while shutdown but rather elected to conclude the outage as previously scheduled and to reschedule overhaul of the SDGs at power as expeditiously as possi-ble.
The A SDG was overhauled on December 7-11, 1987.
Several instances of component wear were identified including two wiped crankpin bearings and two worn fuel injector camshaf t cams.
(The A SDG overhaul is documented in Inspection Report 50-271/87-21, para-graph 10.2). On December 15, 1987, the B SDG was removed from ser-vice to begin the preventive maintenance overhaul but the effort was terminated approximately four hours leter after drywell spray dis-charge isolation valve RHR-26A failed during TS required alternate testing (details of this event are documented in inspection report 50-271/87-21, paragraph 9.4).
The B SDG was returned to service expeditiously on December 15, 1987 to increase emergency system-diversity during this event.
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.- Although the licensee responded appropriately by conservatively re-turning the B SDG to service on December 15, 1987, the licensee has not acted since to accomplish the already past due vendor recommended overhaul. The licensee has delayed the overhaul to avoid the poten-tial for another - event similar to the one on December 15, 1987, caused by the repeated daily cycling of safety related sy,tett. gom-ponents required by alternate testing.
This delay was-in anticipa-tion of NRC approval of'a licenseeLsut'..itted TS amendment which wauld delete alternate testing requirements.
Presently the licensee plans to inspect the scavenging air blower rotor. clearances and repair very minor air and jacket water leaks on May 24, 1988.
The licensee is not required by VYNPS TS or regulatory. commitment to perform the SDG overhaul. However, based on the component wear noted during the A SDG overhaul and generic concerns for increased scaven-ger air blower clearance inspections, the inspectors view the decis-ion of licensee management to continue to delay performance of the B SDG vendor recommended overhaul as non-conservative, 7.4 Emergency Core Cooling System (ECCS) Logic Testing
Performance of the ECCS logic system serveillance testing is cur-rently required by TS to be accomplished every six months.
The six month test frequency necessitates performance of the tests at power I and test methodologies require removal of certain safety systems from service for the duration of specific tests.
Manager of Operations (M00) Directive 88-01, datsc March 18, 1988, was issued by the licen-see to the plant te emphasize the concern for rendering safety' sys-tems inoperable to accomplish surveillance testing and to minimize the impact of such actions.
Prior to the performance of each logic system test, the I&C supervisor or lead foreman conducted a pretest-brief with the on-shift shift supervisor, the I&C technicians per-forming the test and the electrical maintenance personnel supporting the test. The briefings provided a thorough review of the test pro-cedures, amplified the conditions of M00 Directive 88-01, and ensured proper plant corditions.
Dedicated auxiliary operators and elec-tricians were identified whose sole responsibilities were to standby at the switchgear room and motor control centers to immediately restore the system in test to an operable status if required.
On March 24, 1988 OP 4349, "Core Spray Subsystems A/B Logic Test", revision 13 was performed and completed satisfactorily.
No except-ions were noted.
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. On April 5,1988, OP 4354, "RHR Subsystems A/B Logic Test", revision i 15, was performed and completed satisfactorily. One exception which had been identified during previous logic testing but had not been corrected in the current -procedure revision was noted. Upon receipt of a LPCI initiation signal, RHR heat exchanger bypass valves RHR-65A and B receive a 55 second (plus or minus 5 second tolerance) ~open signal in which the' valves should go full open. ' The purpose of the open signal is to ensure that the valves do not go closed. This is done by blocking a close signal vice ensuring the valves go full open. Valves RHR-65A and 8 take approximately 58 seconds to go from full close to full open indicated position depending on limit switch position indication setting. Therefore, it is po'ssible that the 55 second open signal relay will de-energize before the valves stroke . full open. This in fact occurred during the April 5,1988 test and . . during previous logic tests for the RHR-65A valve only.
Subsequent full open indication was achieved within approximately.2 seconds by , remote manual stroking of the valve from the control room.
This exception had been evaluated after the last logic testing by the ! engineering support department (ES0) as a satisfactory condition because the valve is approximately full open and would not affect ,
system performance and because the intent of the open signal is , actually to prevent valve closure.
The evaluation is documented in an ESD memo to I&C dated August 25, 1987.
The licensee plans to revise the appropriate section of OP 4354 to note that RHR-65A or B
not reaching the full open position indication prior to relay de-energization is acceptable based on the above evaluation.
On April 15, 1988, OP 4368, "RCIC System A/B Logic Test", revision 16, was performed and completed satisfactorily.. One procedural ' exception was identified pertaining to the expected relay responses to the return to service of the RCIC steam line pressure switches.
By design of the ene out of two twice logic, return to service of pressure switches PS13-87A or PS13-87B and PS 13-87C or PS 13-870 will de-energize relay 13A-K10.
Procedure OP 4368 'did not identify
13A-X10 as a de-energizing until both pressure switches PS 13-87A and PS 13-878 or PS 13-87C and PS 13-870 were returned to service. The system responded as designed during the test, the exception was noted
and a procedure change was generated.
On April 22, 1988, OP 4360, "HPCI System Logic Test", revision 14, and OP 4361, "HPCI System Isolation A/B Logic Test", revision 16, were performed and completed satisfactorily.
No exceptions were identified.
, The inspectors attended the pre-test briefings and observed the per-formance of the tests.
The briefings were very comprehensive, ad-dressed all questions and stressed the importance of minimizing the unavailability of safety systems. The tests were performed in a well controlled and ef ficient manner. All precautions were observed and proper communications between disciplines were maintained.
The inspectors had no further question. _ _, __ , d e . . h
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. On November 30, 1987, VYNPC submitted TS proposed change 142 to the NRC.
The proposed change would increase the surveillance period for various logic system tests from every six months (which requires - testing at power) to every operating cycle such that testing would
be required only when the. reactor is shutdown, at which time a more comprehensive test could be performed.
7.5 Surveillance Program Review On April 12, 1988, the licensee determined that functional testing of-the scram discharge volume (SDV) high water level trip had not been - tested in accordance with TS Table 4.1.1 NOTE 1 requirements. Note 1 ' to Table 4.1.1 requires that testing be performed monthly initially and between one to three month intervals thereafter based on compo-nent reliability studies. It was determined that the SDV high water level trip had not been tested monthly initially but rather hcd always been tested on a three month frequency since implementation.
. ! This missed TS required surveillance was reported in LER 88-03.
The root cause of this event was the failure of the. licensee to en-sure that the surveillance interval required by the TS amendment issued following completion of EDCR 82-06 was properly observed by
plant procedures.
Prior to the implementation -of EDCR 82-06 the SDV
level instrumentation surveillance interval was three months.
How-I ever, TS Amendment 76, issued following the completion of EDCR 82-06, required that the SDV level instrumentation be tested monthly-initially in accordance -with Note 1 to Table 4.1.1 of the VYNPS TS.
, The surveillance interval as recognized by the surveillance program j remained three months.
In January 1985 an I & C procedure review identified that all other analog instrumentction except the SDV level instrumentation was being tested monthly in accordance with Note 1 to Table 4.1.1.
No action was taken at that time based upon the accept-ability of the SER for the superseded three month TS surveillance interval.
On April 12, 1988, this issue was readdressed.
Plant management at that time properly determined that the requirements of Note 1 to Table 4.1.1 technically had not been' met, in that the SDV level instrumentation installed per EDCR had never been tested monthly.
In recent months the licensee has identified several other missed TS surveillances. On December 15, 1987, the licensee identified a fail - ure to perform the LPCI reactor vessel shroud level permissive for containment spray monthly functional test for the months of October and November 1987 (LER 87-19).
Initial corrective actions to LER 37-19 resulted in the review of the surveillance test tracking system which identified on January 14, 1988 a missed TS required secondary centainment valve exercise for the first quarter of 1988 (LER 87-19,
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revision 1).
On Fe'ruary 16, 1988 the licensee identified that the once per cycle TS required surveillance testing of containment isola-tion valves RHR 32 and 33 on the terminated and non-operable reactor vessel head spray system (RVHS) had not been accomplished since the system was deactivated via PDCR 81-12 (PR0 88-012) (IR 50-271/88-03, paragraph 6.4).
The inspectors conducted an independent assessment of the VYNPS surveillance program in December 1987, following the identification of the initial missed TS surveillance.
Several pro-grammatic weaknesses were identified by the inspectors (IR 50-271/ 87-23, paragraph 6) including; licensee management failure to ensure that the surveillance test coordinator was fully trained in the responsibilities of his position; and the accountability and instruc-tion of the surveillance program controlling procedure, AP 4000, were inadequate to ensure full compliance with program requirements.
The inspectors also identified to the licensee what appears to be a fundamental misdirection of focus pertaining to surveil!ance program philosophy and practice. It appears that at the implementing depart-ment level the focus of ensuring that each TS surveillance reqaire-ment is satisfied is redirected to ensuring that each procedure referenced to include the surveillance requirement is performed. The missed surveillance reported in LER 87-19 is an example of this prac-tice.
The performino department was only ensuring the referenced procedure was completed as opposed to ensuring the TS surveillance requirement was satisfied.
The missed surveil:ance reported in LER 87-19, revision 1 occurred because the surveillance was inadvertently omitted from an updated revision of the 1987 annual schedule.
The missed surveillance of PRO 88-012 was a result of the removal of specific steps to test the valves of the deactivated RVHS from the test procedure although the valves were still required by TS to be tested.
Licensee attention is required to ensure TS surveillance require-ments are the focal point for every aspect of the surveillance pro-gram, including TS amendments, schedular changes, procedure revisions and plant modifications which affect surveillance.
Inherent in this concept is the need to verify the existing accuracy of the master surveillance list from which surveillance program scheduling evolves.
At present, the licensee performs biennial reviews of all procedures to ensure that the specific procedures properly satisfy the specific TS surveillance requirements.
However, currently there is no inde-pendent prog ram to review existing TS requirements to ensure that they are recognized by the master surveillance list and properly implemented by approved procedures.
Prompt licensee attention is necessary to address these concerns from a programmatic standpoint.
To ensure that all TS requirements are properly implemented and that positive control of the surveillance program is realized, a more global approach to problem resolution is require. . ' , . . , '
The licensee has committed to the performance of an indepe.ndent com-prehensive audit of TS to ensure that all TS surveillances a,e being properly implemented by the surveillance program.
Completion of the audit and any resultant corrective actions is an unresolved item pending inspector review (UNR 50-271/88-06-01).
8.
Engineering Support 8.1 Core Spray Safe End Nozzle Weld Overlays By letter dated May 9, 1988, NRC:NRR responded to the VYNPC request of March 1, 1988 te continue operation beyond cycle 13 (current oper-ating cycle) with the current configuration of weld overlays on the core spray nozzles (see Inspection Report 88-03 Section 10.0).
The NRC:NRR found the licensee program to confirm evaluation results and document acceptable continued operation through performance of periodic ultrasonic examination in accordance with NUREG-0313 and Generic letter 88-01 to be acceptable providing inspection results continue to be satisfactory.
The inspector had no further questions in this area.
8.2 Inservice Testing Vibration Monitoring Program Plant technical specifications require that the structural integrity and operability of safety-related systems and components be periodi-cally verified through performance of inservice inspection (ISI) and inservice testing (IST) activities performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code (the Code) and applicable Addenda.
The VYNPC IST Vibration Mor,itoring Program is a facet of the overall program which uses pump vibration measurements to determine operability, confidence in continued reliability and early indication of pump degradation.
The program requirements are currently implemented by administrative procedure (AP) 4024, "Inser-vice Testing" as well as other application oriented guidance.
The sum of this guidance yields a minimal but workable approach to pro-gram proceduralization.
However, gaps in the program definition allowed a misinterpretation of Code requirements which resulted in a Notice of Violation being issued (see Inspection Report 86-25 Section 5.2).
The correct interpretation was debated by VYNPC and NRC on numerous occasions and has been concluded in this inspection report (see cover letter).
The violation precipitated a licensee review of possible program improvements including upgraded vibration monitoring equipment, revised test practices, and a revised corrcctive action process.
The initial result of the review was adoption of a new vibration measurement test system.
Implementation of the new system proved to be dif ficult due to problems with instrument sensitivity, . - - - - - - - - - - - - - - - - - - - - - - - . - - - - - - - .
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, data repeatability, instrument physical deterioration and operator unfamiliarity.
Inspector observations made during this period included a lack of definitive program requirements / guidance, a lack of program leadership, and premature implementation of the new instrument system.
In meetings held with the licensee during this period, the inspector stressed the need for a comprehensive approach to re-establishment of a technically sound and procedurally well-supported program.
Licensee efforts during.this transition period culminated in establishment of confidence - in data generated by an i improved measurement instrumentation system, increased program ' standardization, and improved performance.
The licensee plans to continue program improvement through a reorganization of responsi-bilities and revised procedures. Completion of these activities will be tracked under the original violation (50-271/86-25-01).
Although initially fitful and unfocused, the licensee development of reliable and comprehensive IST Vibration Monitoring Program, that a additionally supports licensee predictive maintenance initiatives, has shown progress and improving results.
Completion of remaining action items should successfully conclude program transition.
8.3 Diesel Generator Operability Plant PRO 87-52 was opened on September 25, -1987 to address the potential reportability of a discrepancy in the diesel generator (DG) air start system. Quality Control (QC) observation of the setting of pressures switch OG-PS-1A in the diesel generator air start system noted that the actual setting differed from that described in the facility Final Safety Analysis Report (FSAR) Figure 10.14.7.
This pressure switch controls the motor driven air compressor which main-tains the required air pressure in the dual air receivers for each of the diesel generators.
The QC inspection report 87-0059 addressed - this discrepancy. The DG-PS-1A was set at 220 3 psig vice the FSAR described value of 225 psig.
Section 8.5.3 of the FSAR also states , that the dual air start receivers are maintained at an air pressure of 225 to 250 psig.
Thus, the licensee initiated change to the DG-PS-1A setting constituted a change to the facility as described in
the FSAR and therefore required a safety evaluation to determine if the change represented an unreviewed safety question per 10 CFR , 50.59.
) The original disposition of PRO 87-52 focused on the operability of the diesel generators with an air start system pressure set lower than described in the FSAR.
The determination was made that the diesel remained operaale and the occurrence was not reportable , because the diesel was still capable of fulfilling its intended func- ' i tion and start in under thirteen seconds per technical specifica-
tions.
Final approval of the PRO by the Plant Manger on September
29, 1987 tasked the Engineering Support Department (ESD) with making ' l any necessary FSAR changes.
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l ' . Ouring a review of various PR0s, the inspector noted that PRO 87-52.
! analysis did not address the required safety evaluation implicit in ' the OG-PS-1A setpoint change. This item was brought to the attention of the Engineering Support Supervisor (ESS) on February 10, 1988 and was discussed on several occasions with the ESS, -the Technical Services Superintendent, Operations Superintendent and Plant Manager.
On May 9,1988 the Plant Operaticns Review Committee (PORC) reviewed and approved the setpoint change request (87-26) for DG-PS-1A/B and its attendant safety evaluation.
The span of time from February 10 to May 9 to develop the required safety evaluation was excessive. This is especially so because the - ' condition had existed since September 1987.. Additionally, safety evaluations for determination of the - existence of potential unre-viewed safety questions are required prior to the actual change..The t licensee demonstrated a lack of continuity and some confusion in the ' generally unsatisfactory response to this situation.
Although the , technical aspects did not represent a safety concern, management 'of
the issue was neither timely nor aggressive. While this item appears to be an anomalous occurrence, the inspector will continue to review licensee execution of 10 CFR 50.59 requirements.
8.4 Scram Discharge Volume (SDV) Level Transmitter Capillary Tubing l Support Seismic Qualification.
! This PRO addressed the Muestion of the seismic qualification of. the SDV instruments with one instrument tubing clamp inactive (missing or . ! loose), as was observed an December 14, 1987.
Engineering support department personnel revieved EDCR 84-429 in which the clamps were installed and held telephone conversations with vendor representa-i
) tives from Rose.aount Inc.
Two support clamp designs were accepted ' as seismically qualified by EOCR 84-429, Unistrut P-2558 clamps and i Unistrut P-1117 clamps.
The P-1117 clamps were installed but they have exhibited problems remaining properly engaged with the Unistrut.
Otscussions with Rosemount, Inc.
representatives indicated that spacing of the clamps up to five feet apart is seismically accept-i able.
The EDCR 84-429 limits clamp spacing to no more than three
feet apart. An actual system walkdown of the tubing indicated that all clamps were less than two feet six inches apart.
Therefore, it , was concluded that if one or even every other clamp were to become inactive the required seismic qualification would be maintained, and > j therefore, such an event would not be reportable.
This PRO was approved by the plant manager on February 29, 1988.
The engineering review performed in support of PRO 88-13 concluded that the Unistrut clamp P-2558 design was superior to the Unistrut i clamp P-1117 design for the specific application and recommended that the P-1117 clamps be replaced by the P-2558 clamps. The recommenda-l tion was accepted by operations management and will be implemented by MR 88-0348 and MR 88-0349.
The inspectors had no further questions.
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9.
Radiologi al Controls 9.1 Sealed Source Survey and Inventory Technical specification (TS) 6.5.F requires semi-annual surveys for contamination / leakage of licensed radioactive sealed sources.
Plant TS 6.6.A 8 requires that results of these surveys be recorded in units of microcuries (uCi) and retained for at least-five years.
Administrative Procedure AP 4500, revision 9, requires that these sources be inventoried at least every six months.
Contrary to the above, the source survey performed on August 7, 1987 logged the results in disintegrations per minute (dpm) vice uCi; results were logged as ">"(indicating greater than) 1000 dpm (indicating contam-ination was present) vice "<" (indicating less than) which was the actual condition; and, a 24 millicurie (mci) Cobalt-60 (Co-60) source was not logged as inventoried. Although the licensee maintains that the 24 mci Co-60 source was surveyed and inventoried later in the day on August 7,1987 af ter the error was noted, no documentation of the activity exists.
None.7 the sources exhibit any leakage, no case of personnel con-tamination occurred during handling of the sources, and the 24 mci Co-60 source was routinely used from August to February (rnost recent inventory). Thus, no hazard existed due to the logging errors.
This was an isolated case of poor administrative control of the survey / inventory process.
The licensee has taken effective corrective measures including AP 4500 changes for clarification.
Because the violation of TS requirements noted above was identified by the licensee, will be reported to the NRC by special report, was of a low severity level, had prompt corrective actions taken, and was not related to corrective actions for a previous violation, no notice of violation will be issued in this instance and the item is con-sidered closed (50-271/88-06-02).
9.2 Service Water Effluent Radiation Monitoring As a result of further investigation into PRO 87-38 and LER 88-01 (see Inspection Report 88-03, Sections 6.2 and 7.0) the licensee determined that there were potentially a number of occasions in the past when the plant service water (SW) ef fluent stream was not ef fec-tively monitored. The cases arise from the fact that the SW effluent radiation monitor alarm setpoint is set at three times normal back-ground per the Offsite Oose Calculation Manual.
Derivation of this setpoint, however, assumes mixing / dilution with other clean plant effluent streams (notably the circulating water (CW) system) during
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! "open cycle" (once-through discharge of CW after cooling the main condenser) plant operations. During "hybrid" or "closed cycle" oper-ations and when the CW system is secured, this dilution effect is not present and the alarm setpoint exceeds 10 CFR Part 20 limits.
Thus, discharges in excess of these limits could possibly occur without alarm indication or licensee knowledge. This condition does not meet the intent of TS 3.9. A.1 and should be compensated for by obtaining and analyzing grab samples on a daily basis.
, Interim corrective measures identified by the licensee include review of the radiation monitoring system design, review of detector cali-i bration guidelines, review of applicable procedures, and obtaining grab samples of the SW effluent during any period when the plant is not operating open cycle, d Because the violation of TS requirements noted above was identified " by the licensee. was reported in LER 88-01, Supplement 1, was of a low severity level, had prompt corrective actions taken, and was not related to corrective' actions for a previous violation, no notice of violation will be issued in this instance. This item is considered open pending completion of final licensee corrective actions (50-271/ 88-06-03).
'] Discovery of this condition was the result of a probing analysis by-the Engineering Support Department and represented a philosophy of excellence in operations.
10.
Licensee Event Reporting (LER) J The inspector reviewed the below licensee event reports (LERs) to deter-mine that with respect to the general aspects of the events: (1) the report was submitted in a timely manner; (2) description of the events was accurate; (3) root cause analysis was. performed; (4) safety implications were considered; and (5) corrective actions implemented or planned were sufficient to preclude recurrence of a similar event.
10.1 LER 88-01 Revision 1
This revision to LER 88-01, "Plant Service Water Effluent Stream Not
Monitored Due to Procedure Deficiency" updates the LER to include ! information relating to a subsequently discovered condition of inade-quate monitoring (see Section 9.2).
The additional information was generated as a result of LER 88-01 analysis.
The revised LER ful- , filled the above criteria and no deficiencies were noted.
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. . - 20- - . 10.2 LER 88-02 and Supplement 1 The LER 88-02 and subsequent Supplement 1, "Main Steam Relief Valve Above Setpoint Due to Steam Cuts on the Pilot Seat and Safety Valve Binds Due to Misalignment" addressed the results of TS 4.6,0.2 required testing of relief and safety valves.
Confusing statements in the original LER 88-02 were discussed with.the licensee and sub-sequently clarified in Supplement 1.
The original and revised LER fulfilled the above criteria and no deficiencies were noted.
10.3 LER 88-03 The LER 88-03, "Missed Surveillance on High Water Level in. Scram Discharge Volume Trip Channel and Alarm Due to Programmatic Defici-
encies" addressed surveillance program problems that led to -a tech-i nical violation of technical specification requirements-(See Section 7.5).
The LER was comprehensive and generally acceptable with the exception of event cor ective actions.
Licensee corrective actions do not address the fac' that the condition was identified in 1985 but a conclusion based u on faulty logic allowed the situation to per- , s petuate. Also, corrective actions do not address the potential for ' this condition to exist in other areas.
This lack is noteworthy in light of recent problems identified ~ in the surveillance program.
These points were discussed with the licensee and are more fully described in Section 7.5 of this report.
No violations of reporting j requirements were identified.
j 11.
Review of Licensee Response to NRC Initiatives 11.1 IE Bulletin 85-03: Request for Additional Information Concerning VYNPC Response > l The IE Bulletin 85-03, "Motor Operated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings" requested l licensees to respond to various information requirements _ contained
in the Bulletin. By letter dated May-14, 1986, VYNPC responded to the Bulletin. Upon subsequent review of the VYNPC submittal, NRC:NRR determined that further inforr.1ation was needed to clarify portions of the response.
By memorandum dated March 30, 1988, the NRC Senior Resident Inspector forwarded a Request for Additional Information
(RAI) concerning IEB 85-03 to the VYNPS Plant Manager.
The RAI ap-pears as Attachment A to this inspection report. Based upon a mutu-ally agreed upoi scope and schedule, the licensee submitted a re-sponse to the RAI on May 5,1988.
This response is currently. under NRC review. The inspector had no further questions.
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. 11.2 Wet Cell Storage Battery Adequacy Audit ! Region I Temporary Instruction 87-07 requires a review to determine if licensees assure that storage batteries will, in accordance with the current licensing basis, remain operable under a variety of con-ditions. The NRC will perform an audit to assess the adequacy of licensee control over storage battery operability and compliance with existing NRC requirements.
Attachment B of this inspection report was provided to the licensee to facilitate a coordinated approach to information gathering and to define the scope of the review.
" Pending review of this information, this Temporary Instruction re-mains open.
11.3 NRC Bulletin 88-01: Defects in Westinghouse Circuit Breakers The subject Bulletin requested licensees to determine if Westing-heute, Series DS, circuit breakers were used in Class 1E applica-tions, and if so, to perform pole shaf t ' weld and breaker closing meetanism alignment inspections.
The licensee determined that the subject breakers are not used in any application at VYNPS nor are they planned to be used.
By letter dated April 8,1988, VYNPC com-municated this response to the NRC within the required timeframe.
The inspector identified no deficiencies.
11.4 Region I Temporary Instruction 87-03, Storage of Transient Equipment in Safety-Related Areas l Region I TI 87-03 addresses a concern for the potential adverse ef-feet that improperly stored temporary equipment may have on nearby safety related equipment.
. Previously IEN 80-21, "Anchorage and Support of Safety-Related Elec-trical Equipment" identified the potential for ancillary items such as dollies, gas bottles, block and tackle gear etc. to dislodge, i impact and damage safety related equipment during a seismic event.
A document search by the licensee assessment coordinator could not locate an internal VYNPC response to the notice.
i The inspectors reviewed the administrative procedures that control , temporary scaf folding erection and storage of transient equipment, j, and conducted plant tours to determine proper implementation.
The
i procedures, AP 0019, "Control of Temporary Load on Piping, Equipment and Structures" and AP 6024 "Plant Housekeeping", provide sufficient
instruction to ensure control over transient equipment.
Routine inspector plant tours indicate that the above procedures are properly , implemented. Gas bottles are properly stowed and secured, gang boxes are stored well away from safety related equipment, and staging and ' scaffolding are erected in accordance with AP 0019.
No significant i
deficiencies were identified.
The inspectors had no further- ] questions.
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Review of Periodic and Special Reports l ' Upon receipt, the inspector reviewed periodic and special reports submit-ted pursuant to Technical Specifications..This review verified, as appli- ! cable: (1) that the reported information was valid and included the NRC-i required data; (2) that test results and supporting information were con- ' sistent with design predictions and performance specification; and
(3) that olanned corrective actions were adequate for resolution of the problem. The inspector also ascertained whether any reported information i , should be classified as an abnormal occurrence.
The following reports were reviewed: Monthly Statistical Report for plant operations for the months of -- ' March and April 1988.
i ! Annual Radiological Environmental Surveillance Report for 1987.
& -- , The inspector reviewed the licensee 1907 annual report of the Radiological
l Environmental Monitoring Program.
The program is designed to: ! , Provide early indication of the appearance or accumulation of radio- -- , active material in the environment caused by the operation of the } d nuclear power plant.
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Provide assurance that the plant's environmental impact is known and -- within anticipated limits j
Verify adequacy and proper functioning of station effluent controls --
j and monitoring systems.
Provide standby monitoring capability for rapid assessment of risk --
to the general public in the event of unanticipated or accidental j releases of radioactive material.
The report provides the following information: Summary of the Radiological Effluent Technical Specifications (RETS) -- required surveillance program.
) Analytical results of environmental samples.
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i Results of laboratory analysis quality assurance programs.
-- Results of the required Land Use Census.
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As a result of the review, the inspector determined that the licensee has complied with the technical specification requir?ments for sampling fre- - quencies, types of measurements, analytical sensitivities, and reporting schedules.
Exceptions to the program were adequately explained and jus-tifiable.
The analyses of environmental samples indicated that dose to humans from radicnuclides of station origin continue to be negligible and have no significant impact on the environment.
13. Management Meetings At parfodic intervals during this inspection, meetings were held with senior plant management to discuss the findings. A summary of findings for the report period was also discussed at the conclusion of the inspection and prior to report issuance.
No proprietary informd. ion was
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ATTACHMENT A . .
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. REQUEST FOR ALDITIONAL INFORMATION (RAI) RE: Review of Responses to Action Item e of IE Bulletin 85-03 ,
Licensee Unit (s): Vermont Yankee 1 j Vermont Yankee Nuclear Power Corp.
Date of Response: 05-14-86 RD 5, Box 169 Ferry Road Respondent: Brattleboro, Vt 05301 Warren P. Murphy, . Vice President and . Manager of Operations , The information provided in your response to Action Iten e of IE Bulletin 65-03 was found to be deficient in some areas.
Provide the additional information necessary to resolve the following comments and questions: ' 1. If MOVATS is planned for application to some MOVs which are not included in its data base, commit to and describe an alternate
j method for determining the extra thrust necessary to overcome
the pressure differentials for these valves.
2. Revise Table 1 of the response dated 05-14-86 to indicate
whether the tabulated differential pressures apply to-opening , the valve, closing the valve or both opening and closing.
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3. Noting per Item 2 above that separate values of differential < pressure for opening are not specified clearly, assume inadvertent operations of the following MOVs.
This assumption is required by Action Item a of the bulletin and is mentioned
in the second paragraph of the response dated 05-14-86.
! J (a) HPCI MOV V-17 is shown normally open in Zone D-11 of Drawing G-191169 Sheet 1 Revision 23, and as MOV 3 on Page 68 of BWROG Report NEDC-31322 dated September 1986.
How would suction from the CST be ensured if this MOV were to be (a)
' actuated inadvertently to the closed position upon intended initiation of the system or (b) left closed inadvertently? (b) HPCI MOV V-20 is shown normally open in Zone R-7 of Drawing G-191169 Sheet 1 Revision 23, and as MOV 8 on Page 68 of the
BWROG Report.
How would discharge to the reactor vessel be
ensured if this MOV were to be (a) actuated inadvertently to
the closed position upon intended initiation of the system or (b) left closed inadvertently? I .,- ,.-,. - _ - - _. _, -, _ . ,,., _ - -,. - -.. - -. . - -..,, _ -. -
< . , ,' '* g g.,; jggg Pego 2 of 2 ,, RAI for Vermont Yankee 1 (c) RCIC MOV V-18 is shown normally open in Zone E-14 of Drawing G-191174 Sheet 1 Revision 18, and as MOV 3 on Page 72 of the BWROG Report.
The question in Item 3(a) above applies here also.
(d) RCIC MOV V-20 is shown normally open in Zone G-10 of Drawing G-191174 Sheet 1 Revision 18, and as MOV 8 on Page 72 of the BWROC Report.
The question in Item 3(b) above applies here also.
(e) RCIC MOV V-1 is shown normally open in Zone E-14 of Drawing G-191174 Sheet 2 Revision 13, and as Trip and Throttle MOV I on Page 74 of the BWROG Report.
How would steam supply to the RCIC Turbine be ensured if this valve were to be (a) actuated inadvertently to the closed position upon intended initiation of the system or (b) left closed inadvertently? 4. The proposed program for action items b, e and d of the bulletin is incomplete.
Provide the following details as a minimum . (a) commitment to a training program for setting switches, maintaining valve operators, using signature testing equipment and interpreting signatures, (b) commitment to justify continued operation of a valve determined to be inoperable, (c) description of a method possibly needed to extrapolate valve stem thrust determined by testing at less than maximum differential pressure, i (d) justification of a possible alternative to testing at maximum differential pressure at the plant, and (e) consideration of pipe break conditions as required by the bulletin.
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ATTACHMENT B c. .. =e e WET CELL STORAGE BATTERY j ADEQUACY AUDIT . <
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General Battery Information
Document the below information for batteries which carry vital loads.
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(1) Qualified, or design, seismic life, t
' (2) Qualified, or design, electrical life.
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!' (4) Time in service.
(5) Plans for replacement.
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. I f [, ,2.. Previous Licensee Actions f , Identify actions taken on the following IE Information Notices: 83-11, ! Possible Seismic Vulnerability of Old Lead Storage Batteries; 84-83, , . Various Battery Problems;~ 85-74, Station Battery Problems; and 86-37, . Degradation of Station Batteries.
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Soismic Lifetime and Qualification - ' .. .. ..
- For batteries supplying vital loads, identify the following information.
, . ' (I) Licensee and/or manufacturer's establishment of seismic lifetime.
This maybe through documentation allowing verification by competent personnel other than the qualifiers and containing design specifica- ' tions, the qualification method, results, and justifications (ref: IEEE 535-1986).
(2) Seismic qualification maintenance.
Identify how the criteria for assuring that the battery and rack will maintain seismic qualifica-tion are defined, available, and used for periodic inspections and cell replacements.
Identify the criteria for determination of - seismic end of life based upon the in-service condition of the ' battery, i ! i \\ l
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, Electrical Sizing and Qualification . For batteries supplying vital loads, identify the following information.
' - (1) Confirmation that the battery size is sufficient to handle the load profile with a suitable margin.
- (2) The means of tracking and control _ of battery loads such that the batteries and their replacements will have sufficient capacity throughout design life, if worst case electrolyte temperature and other worst case conditions exist when the battery is called upon to perform its design function.
. (3) lhe provisions for consideration of the effect of jumpered out cells >' upon the ability of a battery to perform under worst case conditions.
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Battery Ventilation and Protection From Ignition Hazards ., " ' For batteries carrying vital loads, identify the following.
, (1) The provisions for assuring adequacy battery ventilation during ,, normal operation, outages, charging, and discharge.
(2) Adequacy of checks of battery ventilation flow.
(3) Adequacy of controls over battery ventilation impediments such as enclosing the battery space or its ventilation with plastic sheeting, or any other ventilation obstructions, during outages and other periods.
(4) Adequacy of hydrogen detection equipment and its calibration and use, or of the technical justification for not using such equipment.
(3) Knowledge of the hydrogen hazard on the part of plant management, operating shift management, and personnel who access the battery j spaces.
f l (6) Prohibition of hot work and smoking in battery spaces, including I checking the spaces for the residue of such activity.
(7) Assurance that battery cells are secured, with post-to-case and top-to-jar seals tight.
Thermometers should not be left in cells after temperatures are measured.
Caps on the filler openings should be properly secured when not required to be off.
(Cells should be vented only through the flash arrestors.)
i (8) The means of assuring proper elimination of water-carrying pipes (e.g., HVAC lines) from battery spaces, especially those which may carry salt water.
(9) The means of positive control over the quality of water added to the batteries to assure that the manufacturer's recommendations or an appropriate licensee standard are met or exceeded.
(10) The assurance of elimination of combustibles, and loose equipment and conductors, from battery spaces.
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Electrolyte Temperature Control
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- ' For batteries supplying vital loads, identify the adequacy of the
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following.
(1) Avoidance of localized heat sources such as direct sunlight,
radiators, steam pipes, and space heaters.
(2) That the location / arrangement provides for no more than a SF difference in cell temperature, as confirmed by measurements representative of operating conditions.
If this is not the case, then the licensee and manufacturer should have identified the consequent impact on expected battery and individual cell capacity-and life, and surveillance-procedures should reflect the ' additional allowable temperature variation.
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'For batteries carrying vital loads, identify the adequacy of the following.
' (1) provision for a freshening charge after more than 3 months of being on open circuit, unless determined by the manufacturer to be , unnecessary to assure rated capacity throughout life.
(2) Accomplishment of equalizing charges at 18 month intervals, and when the corrected specified gravity (SG) of an individual cell is more than 10 point (0.010) below the average of all the cells, and when the average corrected SG of all cells drop more than 10 points below the average installation value, and if any cell voltage is below 2.13V.
(Specific manufacturer's provisions and assessment may allow the non-performance of some of these recommended charges, or may provide different criteria.)
(3) Control over battery water quality such that specified purify is confirmed before addition, that water added just prior to charging is added only to bring the electrolyte up to the prescribed minimum (to prevent overflow during charging), and that water added after and between charges not bring the level above the prescribed maximum (unless manufacturer's instructions provide for other water addition measures).
That routine float and final end of charge SGs not be taken before 72 (4) hours of float operation after completion of the charge and the last water addition, unless the manufacturer's instructions provided otherwise.
(The need is for measurement of representative cell levels and average them.)
(5) Establishment and maintenance of float voltage on accordance with the manufacturer's instructions.
(6) Assurance that single-cell charger use does not violate Class 1E independence from non-class 1E equipment, i _ J
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Performance Tests and Replacement Criteria
. , . ..,. " * For batteries carrying vital loads, identify the following.
' , (1) Initial acceptance testing which demonstrates the ability of the batteries to meet the manufacturer's rating.
. (2) Service testing which demonstrates the ability to carry the load profile with an appropriate margin for worst case conditions, including end of life loss of capacity under the worst case electrolyte temperature.
(3) Accomplishment of a performance test (capacity test discharge) within the first two years of service and at 5 year intervals until signs of degradation are evident or 85% of the qualified service life is reached.
(4) Annual performance testing of batteries which show signs of degradation or which have reached 85% of the qualified service life is reached.
(5) End of electrical life criteria which consider the rapid end of life drop-off in capacity, worst case state of charge during float service, worst case electrolyte temperature, current DC loads, and the time needed to replace the battery while it can still handle worst case conditions.
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. '. '. $. Other Safety-Significant Wet Cell Batteries . e s.
s For safety-significant wet cell batteries not used for vital loads, show - . how the maintenance program periodically determines the ability to perform the design function and provides for. timely replacement of ' batteries and for maintaining associated equipment (e.g., chargers), . ! [ l l i .
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