IR 05000271/1987012
| ML20237K509 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 08/19/1987 |
| From: | Elsasser T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20237K478 | List: |
| References | |
| 50-271-87-12, IEIN-85-071, IEIN-85-71, NUDOCS 8708270199 | |
| Download: ML20237K509 (21) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION I
Report N i Docket N License No. DPR-28 <
Licensee: Vermont Yankee Nuclear Power Corporation RD 5, Box 169, Ferry Road i Brattleboro, Vermont 05301 '
Facility: Vermont Yankee Nuclear Power Station Location: Vernon, Vermont Dates: June 2 - July 30, 1987 i
Inspectors: Donald R. Haverkamp, Project Engineer William J. Raymond, Senior Resident Inspector Geoffrey E. Grant, Senior Resident Inspector Designate Harold Eichenholz, Senior Resident Inspector, Yankee (Rowe)
Paul Bissett, Reactor Engineer Herbert J ,Kaplan,, ea actor Engineer Approved by: /W '_ W P//f Dale Thomas C. Elsasser y f, Reactor Projects Section 3C Inspection Summary: Inspection on June 2 - July 30, 1987 (Report No. 50-271/87-12)
Areas Inspected: Routine, unannounced inspection on day time and backshifts by resident and region-based inspectors of: actions on previous inspection findings; physical security; plant operations; control room habitability survey; maintenance activities; spent fuel pool reracking activities; and, construction / fabrication quality of new RHR pump impeller The inspection involved 201.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Results: One violation was identified concerning the failure to meet Technical Specification 3.2.7 requirements for the chlorine gas trip setpoints in the toxic l gas monitoring system (Section 6.0). Routine reviews of plant activities identi- I fied no conditions adverse to safe plant operations. A licensee-identified inci-dent concerned the discovery of low level radioactive material (contaminated rags)
outside the radiation control areas. Licensee management met with NRC Region I staff on July 22, 1987 to discuss actions taken for that incident and similar losses of control of low level radioactive materials (section 5.9).
8708270199 870821 PDR ADOCK 05000271 G PDR I
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i DETAILS l 1.0 Persons Contacted 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. T. Linn, Security Director Mr. R. Lopriore, Maintenance Supervisor Mr. M. Mete 11, Engineering Support Supervisor Mr. R. Morrissette, Plant Health Physicist Mr. R. Pagodin, Technical Services Superintendent Mr. J. Pelletier, Plant Manager Mr. J. Sinclair, Plant Administrative Supervisor Mr. T. Watson, Instrument and Controls Supervisor Mr. R. Wanczyk, Operations Superintendent 3 2. 0 Summary of Facility Activities The plant continued end-of-cycle power coastdown operations that began on May 18, 1987. The reactor shutdown for refueling / maintenance outage XIII is scheduled to occur on August 7, 1987. An NRC Region I specialist inspector ccmpleted a review during the period of July 13-16, 1987 of the radioactive effluents and waste programs (Inspection 87-10). NRC Region I and NRC con-tractor operator licensing examiners conducted written and operating (simu-lator and walk-through) exams during the period of July 14-17, 1987 of six reactor operator and one senior reactor operator candidates (examination OL 87-11). The licensee received two additional spent fuel storage racks during the period for possible placement in the spent fuel pool with racks of the same desig A change in the normal resident inspector coverage occurred during the perio In May 1987, William Raymond was selected as the incumbent senior resident inspector at Millstone Nuclear Power Station, Unit 3 in Waterford, Connecticut, and that assignment became effective on July 12, 1987. As a result of that reassignment, Mr. Raymond spent several weeks of the inspection period at the Millstone Station and the Haddam Neck Nuclear Power Station in Haddam, Con-necticut where he would provide, if needed, near-site response to an off-normal or emergency condition. In July 1987, Geoffrey Grant was selected as the incumbent senior resident inspector at Vermont Yankee, and that assignment is to become effective on September 27, 1987. Mr. Grant, currently the resi-i dent inspector at Millstone Nuclear Power Station, Units 1 and 2, provided one week of resident coverage in June July 198 Additional resident inspector coverage was provided for four days in June 1987 by Paul Bissett, a region-based operational programs section specialist, for two days in July 1987 by Harold Eichenholz, senior resident inspector at
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3 Yankee Nuclear Power Station in Rowe, Massachusetts, and for eleven days dur-
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ing three separate periods in June and July 1987 by Donald Haverkamp, region-based project engineer in Reactor Projects Section 3C, which has project over-sight responsibility for Vermont Yankee. Mr. Haverkamp was assigned to co-ordinate the resident inspector interim coverage activities at Vermont Yankee l until Mr. Grant's assignment becomes effective. During the current inspection )
period, nearly full-time resident inspector coverage was maintained at Vermont i Yanke l 3.0 Status of Previous Inspection Findings 3.1 (Closed) Unresolved Item 86-15-03: Review of Licensee Event Report (LER) ,
Concerning Missed Standby Liquid Control Tank Surveillance. This item l concerned the failure to sample the contents of the standby liquid con-trol (SLC) tank after addition of water as specified in the plant's 1 technical specifications. The licensee's corrective actions were re-viewed and found acceptable as documented in NRC inspection report 50-271/86-2 The issue remaining unresolved was review of the applicable l LE Licensee event report 86-14 was issued by the licensee on October l 28, 1986 and was subsequently reviewed by the inspector and discussed ,
in inspection report 50-271/87-02. This item is close l l
3.2 (Closed) Unresolved Item 85-23-02: Implementation of ASME Code Section l XI 1980 Edition Requirements. This item concerned inspector questions !
regarding the incorporation into plant procedures of inspection require-ments contained in IWV 3300 of the 1980 Edition of ASME Code Section X This edition of the code clarified the inspection requirements to verify j correct remote valve position indication of safety-related valves. The ;
inspector reviewed the list of valves designated by the licensee as re-quiring such inspection. The list was revised and updated in Revision 12 of procedure 0P 4102. Valve RV-40 has been added to the lis The !
inspector noted no discrepancies. This item is close l 3.3 (Closed) Follow Item 83-21-01: Followup of Double Blade Guide Shipmen The inspector reviewed plant information report 83-07 which was written i to address the incidents involving the double blade guide shipment to l Pilgrim and the HN-100-2 shipment to the Beatty burial site. The cor- j rective actions taken by the licensee were adequate to prevent recurrenc i The inspector noted that the corrective actions appeared to be effective in that no similar incidents have since occurred. This item is close )
3.4 (Closed) Follow Item 84-01-03: Followup of Glycol Intrusion Event. The inspector reviewed plant information report 84-03 which provided the licensee's evaluation of the January,1984 incidents where glycol from j the advanced offgas system closed cooling water heat exchangers entered l the primary system. The licensee's evaluation concluded that no adverse i effects occurred as a result of the incident. Short- and long-term cor- l rective actions were initiated which were effective to prtclude recur- l rence of the event. This item is close !
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3.5 (Closed) Follow Item 84-18-06: Dryer / Separator Lift Problem. This item was last reviewed in NRC inspection 84-21 and was also addressed as part 1 of violation 84-21-07. The results of the licensee's evaluations of I plant operation with the dryer / separator not fully secured to the core shroud were addressed in plant information report 84-08. The root cause l of the event was adequately identified and addressed by the licensee.
! This item is close .6 (Closed) Unresolved Item 83-18-02: Appendix J Containment Testing. This i item concerns the issue of adjusting the 10 CFR 50 Appendix J containment l leak rate test results to reflect leakage improvements identified as a result o' Type 8 and C testing. The issue is also incorporated and .
tracked in unresolved item 86-14-01, and is discussed further belo !
Inspection item 83-18-02 is close .7 (0 pen) Unresolved Item 86-14-01: Appendix J. Testing. This item was open l pending revisions to improve test procedure OP 4029, which is used to complete containment integrated leak rate measurements, and pending com-pletion of licensee actions to include the methodology dercribed in NRC information notice (IN) 85-71 in the containment test progra The methodology described in IN 85-71 would require that the licensee add the " penetration leakage improvement" identified as a result of the l Type 8 and C testing to the Type A integrated test results to obtain the
" total" containment leakage. The licensee has declined to use this approach on the basis that the methodology is not explicitly required in any existing NRC regulation, and based on the current technical specification requirements that establish leakage limits for each type of Appendix J containment test. The issue was discussed with licensee management during a meeting on March 27, 1987 and addressed in a letter to the licensee dated June 16, 1987. As a result of this review, the NRC staff concluded that no further licensee actions are necessary at the present to change the licensee's. program to incorporate the IN 85-71 methodology. It was noted, however, that procedural revisions may later be necessary as a result of pending NRC action to revise 10 CFR 50, Appendix ..
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This item remains open pending completion of licensee actions to revise procedure OP 4029 to incorporate the other improvements noted in section 4.1 of NRC inspection report 86-1 .8 (Closed) Unresolved Item 82-11-02: Drywell Cooling Capabilities. This item was last discussed in NRC inspection report 82-16, section The licensee's reviews and followup evaluation regarding drywell tem-peratures and cooling capabilities were provided in plant information .
report 82-06 dated August 10, 1982. Actions were completed to evaluate drywell temperature monitoring (YAEC 1403 dated February 1,1984),
establish revised alarm setpoints for drywell temperature instruments (Setpoint Changes 85-18 through 85-21), revise surveillance and operating procedures to formalize plant actions on increasing containment tempera-
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i tures, and to improve the reliability of the drywell cooler The in-spector noted that performance of the drywell coolers had improved in that no failures had occurred during the last several operating cycle This item is close .9 (0 pen) Unresolved Item 85-25-11: NRC Review of the LLRW Facilit Con-
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struction of the on-site low level radwaste facility was completed in 1985, but plans for use of the facility were suspended due to continued availability of the LSA storage sites in Nevada, South Carolina and Washington. The licensee also deferred completion of a safety evaluation l to demonstrate that use of the storage area would not create health or l
safety hazard This item remains open pending completion of a safety evaluation by the licensee pending a decision to use the facility to store radwaste, and subsequent review by the NR .10 (0 pen) Violation 83-26-01: APPENDIX R Audit - Fire Protection Program I Deficiencies. Following a review of the findings from NRC inspection report 83-26, the NRC staff issued a Notice of Violation to the licensee on June 15, 1987 for the licensee's apparent failure to adequately re-assess fire protection program features in accordance with 10 CFR 50, l Appendix R requirements as explained in NRC generic letter 82-1 A l severity level III violation was issued with no civil penalty because of the apparent lack of clarity which existed regarding fire protection l regulation at the time the licensee took actions to meet the Appendix i R requirements. The licensee was requested to respond to the Notice by l July 15, 1987. By letter FVY 87-71 dated July 1, 1987, the licensee requested that an additional 30 days be allowed to prepare the formal ,
written response to assure that the staff's concerns are fully addresse I No inadequacies were identified. This item will be reviewed further by the NRC staff following receipt of the licensee's respons .11 (Closed) Follow Item 85-36-04: Control Room Ventilation System Isolation During Radiological Event. This item was addressed in a memorandum from G. Cappuccio to R. Wanczyk dated February 5,1986, which summarized the actions taken to address this issue and the issues discussed in Sections 3.12 and 3.13 belo The licensee concluded that no automatic isolation of the control room ventilation system for radiological conditions would be necessary for the spectrum of analyzed design basis events, so long as the plant as-built conditions match the 20 scfm analysis assumptions for inleakag The inleakage assumptions would be met if the control room kitchen and bathroom dampers were required to assure isolation in response to manual actuation of the ventilation system recirculation system control switch on panel CRP 9-25, as well as upon actuation of the toxic gas system.
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The original plant design as described in the FSAR did not incorporate dampers on the control room kitchen and bathroom. The dampers were added as part of the EDCR 82-33 and -38 design changes to provide the toxic gas isolation and pressurization systems. The controls for the control room dampers were required to provide the desired isolation function as
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part of design changes completed per plant alteration request (PAR) 85-05 on January 15, 1987. Plant operators switched control room ventilation from the " normal" to " emergency" mode at.the request of the inspector on July 22, 1987. The insp'ctor verified that the kitchen and bathroom dampers isolated as required in response to the manual actuation. The licensee further assured that plant procedures required the operator to manually place the control room ventilation system in the recirculation l mode for events involving excessive r6diation level No inadequacies were identifie The NRC staff followup of the results from "e control room habitability survey is discussed further in Section .io .12 (Closed) Follow Item 85-36-05: Clarify Procedures for Control Room Ven-tilation System Operation. See item 3.11 above for further information ,
on this issue. The licensee identified the procedure changes required I as a result of the PAR 85-05 modifications. Changes were made to proce- l dures OP 2192, Heating, Ventilating and Air Conditioning System, and ON 3153, Excessive Radiation Levels that assure that operation of the con-trol room kitchen and bathroom dampers in response to use of the CRP ventilation system recirculation switch is adequately describe The inspector verified that procedures OP 2192, Revision 12 and ON 3153, Re- i vision 2, contained the appropriate changes to reflect the present desig The inspector noted based on discussions with operations personnel during previous routine inspections that the operators were familiar with the expected operating characteristics of the present design. No inade-quacies were identified This item is close .13 (Closed) Follow Item 85-36-06: Chlorine Gas Detector Setpoin See Item l 3.11 above for further information on this issue. This item was open !
pending licensee clarification of the proper setpoint for the chlorine gas detector in the toxic gas monitoring system. The licensee verified that the FSAR value of 5 ppm (reference FSAR Table 7.19.1) was the ap-propriate setpoint and that this was the trip settin The licensee noted in a memorandum dated February 5,1986 that calibration procedure 0P 4329 specified that the setpoint be at 7 ppm and initiated action per Job Order #86-04 to revise the procedure and set the chlorine monitor setpoint at less than or equal to 5 ppm. The inspector reviewed Revision 2 of OP 4329, Toxic Gas Monitoring System Calibration dated June 26, 1987, and the most recent calibration for both toxic gas monitor channels com-pleted in accordance with VYOPF 4329.06 on July 1, 1987. The procedure had not been changed as of July 23, 1987 and the trip setpoints estab-lished for both monitors was 7 ppm. This trir value was contrary to the Technical Specification 3.2 requirement that the trip setpoint be less than or equal to 5 ppm. This issue is discussed further in Section belo The licensee also verified that the alarm response procedure specified the correct actuation setpoint at 5 ppm and contained the correct in-struction for the operator to respond to a toxic gas monitor system alar .. . _. .- - _ - _ _ _ . _ _ _ _ _
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l The inspector reviewed the alarm response procedure and identified no inadequacies. This item tracking the need to clarify the toxic gas monitor setpoints is close .14 (Closed) Unresolved Item 80-15-04: GE Electrical Penetration Assemblie The inspector reviewed Qualification Documentation Review (QDR) Package 15.3 for GE electrical penetration assemblies, which provided the licen-
! see's verification that the drywell penetrations are environmentally l qualifie The inspector noted by review of the associated QDR drawings
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that GE penetration Types NS04 (low voltage), NS03 (medium voltage / power),
and NS02 (high voltage / neutron monitoring signal cable) were the type installed in the plan This item is closed.
l 3.15 (Closed) Unresolved Item 80-15-01: MSIV Qualification Documentatio The inspector reviewed qualification documentation review (QDR) package 9.2 which provided the licensee's verification that components associated with the main steam isolation valve (MSIV) reactor protection system (RPS) and position indication switches were environmentally qualifie The components described in the qualification documentation included the position switches, the RPS switches, MSIV mounted terminal boxes, inter-connecting wires, terminal strips and lugs, flexible and rigid conduit, and penetration connection boxes and covers. This item is close .16 (Closed) Unresolved Item 80-15-05: HPCI Qualification Documentatio The inspector reviewed qualification documentation review (QDR) package 3.1 which provided the licensee's verification that components associated with the high pressure coolant injection (HPCI) system were environment-ally qualified, and specifically, the inboard steam line isolation valve MOV-23-1 The components described in the qualification documentation included the valve motor, terminal boxes and lugs, penetration connection boxes and covers, flexible conduit, and interconnecting wires and cable The inspector noted that, contrary to the NRC inspection report 80-15 l discussion, there are no RPS limit switches associated with MOV-23-15,
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and that the position and torque limit switches are built into the valve motor. This item is close .17 (Closed) Unresolved Item 81-05-09: SRV Pressure Switches. The inspector reviewed qualification documentation review (QDR) package 36.1, which provided the licensee's verification that components associated with the safety relief valve (SRV) position indication were environmentally quali-fied. The components described in the qualification documentation in-cluded the Static-0-Ring pressure switches for PS-2-71-1A through 10 and PS-2-71-2A through 2 The QDR package 8.4 provided documentation show-ing environmental qualification of the accelerometers used in the posi-I tion indication circuits for the safety valves. The Babcock & Wilcox l units were replaced with ENDEVC0 accelerometers and charge converters.
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8 1 3.18 (Closed) Unresolved Item 82-23-09: MSIV Position Indication. The in-spector reviewed the QDR package for the outboard MSIV RPS and position indication switches as noted in section 3.15 above. The documentation showed that the switches were qualified for the harsh environments as-
.sociated with a high energy line break. The inspector noted that, fol-lowing repair of the switches in 1982, there have been no subsequent problems with the position indication circuits, including periods of operation with steam leaks in the steam tunne This item is close .19 (Closed) Unresolved Item 84-18-02: RHR-18 Environmental Qualificatio The inspector reviewed section 5.0 of the licensee's environmental quali-fication manual and the electrical component matrix report and its as-sociated technical bases (TB) notes. Technical bases notes TB17 and TB39 described those portions of the residual heat removal (RHR) system that are required to be environmentally qualified. The licensee determined that since RHR 18 only opens for the .:old shutdown operating mode of RHR, and since " safe shutdown" could be provided by the RHR system in the ,
torus-to-reactor vessel-to-torus (via break) flow path following a high '
energy line break (HELB), there was no basis to qualify the RHR 17 to 18 valves for EQ purposes. The valves are normally closed when the reactor is pressurized and would be in the closed position at the start of a HELB even To further assure this assumption, the licensee in-stalled a key lock switch in the RHR valve shutdown cooling interlock circuitry to assure that the RHR 18 valve is deenergized in the closed position during normal operations. Based on the above, environmental qualification of the motor and brake for the RHR 18 valve is not required.
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This item is close .20 (Closed) Unresolved Item 87-09-02: Exemption From Appendix J Test Re-quirements. Following further revi a of this item with the NRC Region I technical staff, the inspector det. ermined that torus electrical pene-tration X-215 and the associated thermocouple are of the type of pene-tration addressed in the NRC staff SER dated August 19,.1983, and are therefore exempt from the testing requirements of 10 CFR 50, Appendix J. Based on the above, the licensee is not required to complete a con-tainment leak rate test following the scheduled replacement of the TE 16-19-34 thermocouple during the 1987 refueling outage. This item is close .21 (Closed) Unresolved Item 86-18-01: Licensee to Further Investigate and Evaluate the Failure of Valve DV-301. The licensee had an Automatic Sprinkler Corporation representative assist in performing preventive maintenance on the valve on March 27, 1987. No abnormalities were note Maintenance request 87-0241 and the associated valve test report were reviewed by the inspector. Test report results were found acceptabl The licensee intends to initiate an annual maintenance contract with Automatic Sprinkler to ensure the continued operability of DV-301. This item is close _ _ _ _ _ _ - _
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1 3.22 (Closed) Unresolved Item 85-26-02: Quality Assurance Technician and Senior Quality Assurance Technician Positions to be Filled. Since the i time this issue was identified, the licensee has filled the QA technician positions, however, they have upgraded the senior QA technician position to a receipt inspection supervisor. Discussions with the engineering )
support supervisor indicated that the job interview process had been l completed, the selection had been made, and that corporate approval was I all that was needed prior to filling the position. The supervisor stated l that the position would be filled within a matter of week Based upon l the above review, this item is close I i
3.23 (Closed) Unresolved Items 85-26-04 and 86-20-02: Licensee to Clarify the l Requirements and Intent of Signing the " Inspection / Hold Point Completed l By" Block on the Maintenance Request Form. The inspector reviewed ad- !
ministrative procedure AP-0021 " Maintenance Requests" and AP-6025 " Qual- j ity Control / Independent Inspection" and associated department instruc-tions which clarified the intent of inspection hold point They also provided additional guidance for operation quality and independent in-spection personne These items are close )
3.24 (Closed) Unresolved Items 85-26-05 and 85-20-01: Adequacy and Timeliness i of Corrective Actions Associated with Audits VY-85-11 and VY-85-08 Both audits, VY-85-11 and VY-85-088, were reviewed during NRC inspection 86-20. Corrective action for audit VY-85-08B was found to be adequate ;
and timely. However, the corrective action for audit VY-85-11 was not !
timely. The licensee subsequently initiated action to resolve this problem. The inspector reviewed various 1987 memoranda dealing with actions instituted to preclude untimely corrective action. Also, various meeting minutes were reviewed which discussed the audit program in gene-ral and specifically the importance of timely corrective action. To verify the effectiveness of licensee actions in this regard, the inspec-tor held discussions with quality assurance personnel and reviewed the
" Summary Status of Corrective Action" for various audits. It appeared, from this review, that corrective actions were being addressed and/or implemented in a timely fashion. The above items are close .25 (Closed) Unresolved Item 86-20-03: Peer Inspection to be Addressed in the Licensee's General Employee Training Program. The inspector held discussions with the Training Supervisor and reviewed the lesson plan utilized during general employee training that addresses the peer in-spection progra This item is close .0 Observations of Physical Security 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 measures: guard staffing; vital and protected area barrier integrity; main-
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tenance of isolation zones; and, implementation of access controls, including ,
authorization, badging, escorting, and searches. No inadequacies were iden- l tified, except as discussed belo The inspector reviewed the licensee's response actions following a moderate loss of security effectiveness due to equipment failures at 10:36 a.m. on July ,
22, 1987 and at 7:43 p.m. on July 28, 1987. The inspector verified that the !
compensatory measures established during the equipment outages were appropri- l ate, and that the licensee reported the events to the NRC Duty Officer as 'I required by 10 CFR 50.73. No inadequacies were identifie ;
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5.0 Operational Status Reviews l Plant tours were conducted routinely to review activities in progress and to verify compliance with regulatory and administrative requirements. Tours of accessible plant areas included the control room, reactor building, ',urbine building, diesel generator rooms, and the protected area. Radiation controls were reviewed in areas toured to verify access control barriers, postings and radiological controls were appropriate. Plant housekeeping conditions and shift staffing were reviewed. Shift logs and records were reviewed to deter- i mine the status of plant conditions and the changes in operational statu .)
The NRC Region I Administrator, the resident inspector and the plant manager I completed a tour of the facility on July 7, 1987. Several items noted during j the tour warranted followup, which are identified in Attachment I to this [
report and were identified to the licensee for review and action as necessar Licensee actions on the items will be reviewed on a subsequent routine in-spection (UNR 87-12-01).
5.1 Safety System Review The residual heat removal, core spray, residual heat removal service l water, high pressure coolant injection, service water, reactor core I isolation cooling, standby liquid control and standby gas treatment l systems were reviewed to verify the systems were properly aligned and fully operational in the standby mode. The review included verification that (i) accessible major flow path valves were correctly positioned; (ii) power supplies were energized, (iii) lubrication and component cooling was proper, and (iv) components were operable based on a visual inspection of equipment for leakage and general conditions. No inade-quacies were identifie .2 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 l letter FVY 82-105. The licensee reported that, based on the leakage j monitoring data reported as of June 26, 1987, there were no deviations
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I I in excess of 0.10 from the steady state value of normalized thermocouple readings, and no failures in the 16 thermocouple installed on the 4 feedwater nozzles. No unacceptable conditions were identified.
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5.3 Switchgear Room Carbon Dioxide System Activation l
At approximately 5:40 p.m. on June 30, 1987 the west switchgear room carbon dioxide (CARD 0X) system initiated. The initiation had been pre-ceeded by two separate spurious detector alarms ten minutes and twenty-five minutes prior to initiatio Each of these alarms was acknowledged and investigate In each case, local examination of room detectors did not identify any in the alarm state and the panel alarm was reset. In
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l l the third instance, the first detector alarm was followed by the second detector alarm in approximately two minutes. This caused the CARD 0X System to initiate and discharge to the west switchgear roo Post-initiation licensee investigation and recovery actions were taken. No fire nor cause for detector alarms was present. The switchgear room was ventilated, access restored and a fire watch was stationed in the room, An unidentified leakage path allowed carbon dioxide (CO2 ) to escape from the switchgear room and enter the control room. C02 build-up depressed the oxygen concentration in the control room from a normal level of 21%
to a low level of approximately 19.5%. Although low, this level is still considered safe for personnel. The C0 2 entry into the control room also caused the toxic gas monitor to alarm (approximately 800 ppm C02 ) and initiate the toxic gas system. This resulted in damper closings isolat-ing the control room ventilation system and discharge of oxygen into the control room from rack mounted bottles. This system operated correctly and was returned to stand-by status after normal conditions were estab-lishe Several deficiencies were noted during this event and are actively being addressed by the license These include:
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Multiple spurious switchgear room detector alarms.
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Premature discharge of the CARD 0X system into the switchgear room
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(occurred in 10 seconds vice design 30 seconds)
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Failure of the switchgear room ventilation exhaust fire damper to automatically shut (possible cause of C02 in control room).
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Leakage of CO 2 into the control roo Licensee corrective actions concerning these deficiencies will be re-viewed during routine inspection activities. Because the unidentified leakage path of CO 2 into the control room presents a potential hazard to the operators, this item will remain unresolved pending licensee identification and correction and NRC review (UNR 87-12-02).
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5.4 Inoperable Equipment Actions taken by plant personnel during periods when equipment was in-operable were 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. This review was completed for the items listed in Table A, which appears at ;
the end of this repor '
The inspector noted, in particular, the failures of various local power range monitors, which were all inputs to average power range monitor E, '
that occurred in July 1987 (see section 5.7 for failure review details).
Also, the inspector noted the repeated occasions of toxic gas monitors (TGM) being declared inoperabl In most cases the inoperable condition was the results of TGM removal from service for chlorine channel cali-brations. As noted in Table A, certain occasions of components becoming inoperable are discussed further in sections 5.3, 5.8, 6.0, and 7.0 of this report. Except as described in section 6.0, no inadequacies were identified during the period regarding inoperable equipmen .5 Review of Jumpers and Lifted Leads i Jumper and lifted lead (J/LL) requests 87-26 to 87-40 were reviewed to l verify that controls established by AP 0020 were met, no conflict with l the technical specifications were created, the requests were properly improved prior to installation, and a safety evaluation in accordance with 10 CFR 50.59 was prepared if required. Implementation of the re-quests was reviewed on a sampling basis. No inadequacies were identified.
l 5.6 Review of Switching & Tagging Operations The switching and tagging log was reviewed and taggir.g activities were inspected to verify that plant equipment was controlled in accordance with the requirements of AP 0140, Vermont Local Control Switching Rules.
l Action completed under orders87-366, 87-382,.87-402,87-422, 87-468,87-484 and 87-486 were reviewed'and no inadequacies were identified.
! 5.7 Failed APRM Inputs l
The inspector reviewed actions taken during the inspection period re- i garding average power range monitor (APRM) channel 'E' when it was de- I clared inoperable with less than the minimum number of operable local l power range monitor (LPRM) inputs as specified by the plant technical i specifications. The channel was declared inoperable and bypassed at 12:15 a.m. on July 15, 1987 upon loss of LPRM 08-250. The channel was returned to service at 4:00 p.m. on July 15, 1987 after plant personnel returned LPRM 16-41C to service. The channel was again declared in- .
operable at 1:45 a.m on July 18, 1987 when plant operators noted that I the signal from LPRM 32-33C was drifting out of allowable values. Chan-nel 'E' was declared operable at 1:3A p.m. on July 21, 1987 after plant
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personnel restored LPRM 08-098 to an operable status. The channel was l declared inoperable at 12:28 p.m. on July 22, 1987 when the input from !LPRM 16-41C failed. The channel was returned to service at 3:10 i on July 23, 1987 when LPRM 24-25B was determined operable. Channel 'E' I was again declared inoperable at 8:10 p.m. on July 29, 1987 when opera- l tors observed LPRM 08-098 spiking high. In addition,LPRM 24-25B was declared inoperable at 1:50 a.m. on July 30, 1987 when operators observed I that it was ' spiking hig The APRM channel remained bypassed at the end
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of the inspection perio Periodically during the inspection period, the inspector verified (i)'the proper number of LPRM inputs were operable in the associated APRM chan- i nels as required by Technical Specification 3.1.1, Note 5; (ii) inoper- '
able APRM channels were tripped or bypassed as required; and (iii) the number of APRM channels met the minimum operability requirement The licensee plans to repair or replace inoperable LPRM channels as a routine maintenance activity during the 1987 refueling outage. No inadequacies were identifie .8 HPCi System Overspeed Trip Failure j The licensee declared the high pressure coolant injection (HPCI) system inoperable at 10:15 a.m. cn June 10, 1987 when it was determined that
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the Terry turbine mechanical' overspeed trip mechanism was inoperabl The plant was operating at 95% of rated power in end-of-cycle coastdown
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at the time. The discovery was made d., ring routine monthly HPCI testing per procedure OP 4120, which was revised in May,1987 to include a check j
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of the trip assembly freedom of vovement in accordance with the recom- l mendations of GE RICSM 004 (SIL #353). The licensee determined that I the trip assembly tappet was frozen in place and therefore not capable of initiating a turbine trip iri response to'an overspeed condition. The .
overspeed trip function was last tested satisfactorily on July 2, 1986 -
during an uncoupled run of the turbine as part of the start-up program from the last refueling outag Plant operators isolated the HPCI system for maintenance and completed alternate testing per Technical Specification 4.5.E of the reactor core isolation cooling (RCIC) system, the automatic depressurization system (ADS). logic, and the low pressure core cooling systems by 10:00.p.p. on June 10, 198 The technical specifications allow continued plant" operations for seven days without the HPCI system as long as the RCIC-system and ADS logic is tested daily and proven operable. Surveillance testing was completed daily as' required until the HPCI system was re-turned to service. Actions were taken on June 11, 1987 to inspect and repair the tappet assembly and to restore the 10- to 14-mil cold clear-ances recommended in RICSIL 004 between the tappet assembly and the valve body inner diamete ,
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.I Following installation of the repaired trip assembly, an overspeed trip ^
test was performed during an uncoupled run of the turbine at 3:25 on June 11, 1987. The test failed when the turbine tripper' at 4600 rpm instead of 5000 rpm. The licensee completed actions with vendor assist-ance to adjust the turbine overspeed trip setpoint. The turbine over- ,
speed test was completed satisfactorily at 2:37 p.m., and the HPCI system '
was returned to an operable status at 6:05 p.m. on June 13, 1987. No inadequacies were identifie :
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5.9 Release of Contaminated Material from the RCA - 6/15/87 1 On June 15, 1987 an incident occurred involving bags of insulation found in the turbine building dumpster, which is located outside the Radiation
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Control Area (RCA). The bags were determined +, contain radioactive l l materials above the licensee's RCA administrax.i /e release limit. The l
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licensee's investigation of this incident revealed the following infor-matio During performance of a routine survey of the turbine dumpster, two bags
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were found to be reading 2000 to 4000 ccpm on an RM-14 with an HP 210 I i
prob Inside the bags several rags were found to be the source of activit The bags read 0.1 to 0.2 mR/hr with an E-520. The licensee's investigation concluded that the bags originated in the RHR corner room Three bags were brought to the trash frisking area sometime on Friday, June 12, 1987. Both of the technicians assigned to frisking trast re-
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called seeing only one bag, but somehow over the weekend the bags found their way over to the dumpste As a result of the incident, the licensee closed several of the exits from tN RCA. In addition, frisker watches were posted at several exists to ensure proper frisking of personnel and equipment. These actions were described in a memorandum dated June 19, 1987 from the radiation protec-tion manager to plant, corporate, and contractor personnel. The plant manager reported the incident to NRC Region I on June 16, 1987, reviewed the matter and corrective actions at a department heads meeting, and then with all personnel on site on June 17, 1987. At various times during the period, the inspector observed that licensee actions were being taken as described above and appeared to be effectiv The inspector reviewed the details of this incident with NRC Region I radiation specialists. The specific incident did not create a signi-ficant hazard or safety concern and no regulatory limits were exceede However, the incident was similar to previous losses of control over low t level radioactive materials that occurred on February 26 and April 21, 1987, as discussed in NRC inspection reports 87-04 and 87-09. These incidents and licensee corrective actions taken or planned were discussed during a meeting on July 22, 1987 between licensee management and NRC Region I staff, as described in an NRC letter to the licensee dated July 29, 198 During the meeting, licensee representatives committed to com-
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plete certain actions, as identified in enclosure 2 to the NRC lette These actions will be tracked as part of inspection item 87-04-03, which remains unresolved pending further NRC review of the licensee's controls on a subsequent inspection by an NRC health physics inspecto .10 Review of Potential Reportable Occurrences The inspector reviewed the potential reportable occurrences listed in l Table B, which appears at the end of this report. The inspector deter-mined that the licensee's review and assessment of the occurrences ap-peared appropriate for the stated condition, and that the items are not ;
reportable for the reasons stated in the licensee's c' determinations. No '
inadequacies were identifie I l
5.11 Worker Injury '
l A medical emergency was declared at 1:20 p.m. on June 4, 1987 when a l contractor was injured while working on the refueling floor. While moving a steel I-beam lifting fixture on a dolly, the worker lost control of the load and the beam crushed the middle finger of his left han The worker was escorted to the health physics check point, undressed from l his protective clothing, frisked and found to be free of contamination, and transported to trattleboro Memorial Hospital via ambulance for treatment. The resident inspector reviewed the licensee's followup actions and identified no inadequacie .0 Control Room Habitability Surve_y The NRC staff completed a survey of the control room ventilation system in October,1985 to verify that design features were incorporated as necessary l to satisfy the habitability requirements of Item III.D.3.4 of NUREG 073 l The preliminary findings from the NRC survey were issued in NRC inspection report 85-36 dated January 6, 1986 and by NRR letter dated May 20, 1987. The survey results determined that the ventilation system at Vermont Yankee was adequate, but noted some areas for further review were required, including the need to issue technical specifications that: demonstrates the control room '
isolates on toxic gas and radiological incidents; demonstrates the control room envelope is maintained such that inleakage is maintained less than 20 cfm; and, limits equipment temperature Licensee actions to address some of these issues is discussed further in sec-tion 3.0 above. Actions were taken to assure the 20 cfm inleakage limit is maintained, and to assure plant procedures call for the operator to manually isolate the control room in response to a radiological event. The NRC staff did issue technical specifications in Amendment 96 dated August 11, 1986 which addressed operability requirements for the toxic gas monitoring system, which will provide for automatic isolation of the control room in response to a toxic gas event. No actions were taken by the licensee to address limiting equipment temperatures or to add radiation monitors to the control room or !
the ventilation system fresh air intake, l'
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The survey noted further that there appeared to be a breakdown in the transfer of information from those personnel doing safety evaluations to those imple-menting the safety evaluation assumption through operations and procedure An example of this problem was noted Juring the present inspection and is discussed further below. The stafi's May 20 letter stated that the survey results would be incorporated in a future formal report and would be used in considering further regulatory activity. The NRC staff review of the survey results is in progress and will include a determination of whether enforcement actions are appropriate as a result of the survey findings. This item is unresolved pending completion of the NRC staff review of the licensee's ac- 1 tions in response to the October, 1985 control room habitability survey (UNR l 87-12-03), i i
The following discrepancy was identified by the inspector duaing a review on July 23, 1987 of the licensee's actions in response to the Lctober, 1985 sur- i vey. The NRC staff noted during the 1985 survey conflicting requirements in !
plant references for the chloride gas trip setpoint. At the time of the 1985 survey, proposed technical specifications for the toxic gas system were under review by the NRC staff. In followup to inspection item 85-36-06, the licen-see noted in February,1986 that 5 ppm was the proper chlorine trip setpoint, I as described in the FSAR and in the proposed technical specifications. The i licensee further noted that the actual trip setpoint established by calibra- '
tion procedure OP 4329 was 7 ppm and that the procedure should be changed to establish a 5 ppm trip setpoin The procedure revision was tracked as a follow item for PAR 86-05, which was used to implement the design change ,
needed to isolate ancillary control room dampers in response to a manual i actuation of the " emergency recirculation" mode of the ventilation syste l The Technical Specifications were revised on August 11, 1986 through Amendment 96 to incorporate limiting conditions for operations and surveillance require- d ments for the toxic gas monitoring system. The trip setpoints specified in Table 3.2.7 require the systim to isolate the control room at concentrations of 5 ppm for chlorine gas. As of July 23, 1987, no actions had been taken l to mise the trip setpoints established by 0P 4329, and basM on a review of the OP 4329 calibration records, the inspector verified tnat 7 ppm had been used in the periodic calibrations for both the redundant trains up to and 4 including the most recent calibrations on July 1, 1987. Although the 2 ppm )
difference in trip setpoints is not significant from a safety perspective, '
the inspector noted that both trains of the toxic gas system were technically inoperable due to the non-conservative trip setpoints. The failure to have the toxic gas monitoring system operable with trip setpoints less than or equal to 5 ppm is contrary to the requirements of Technical Specification
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i Table 3.2.7 (SL4 87-12-04).
The inspector noted further during the review of OP 4329, Revision 2, that I the trip setpoints for the other four toxic gas channels (methane, vinyl l chloride, carbon dioxide, ammonia) were specified in the calibration procedure I
with a range that equaled +/- 1% of the desired trip setting. For example, the calibration procedure established an acceptable trip setting range of 792 to 808 ppm for vinyl chloride, as compared to a specified trip setting of less
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l than or equal to 800 ppm in the technical specification. The licensee stated that a review was completed of the toxic gas system to verify that the re-maining four trip setpoints were correct. The licensee stated, however, that OP 4329 would be revised to lower the acceptable setpoint range to at least 10% b91ow the specified trip setting to assure that the technical specifica-tion 1 bit would not be exceede The inspector discussed this finding with operations and instrument & control personnel and the technical services superintendent on July 23, 1987. The operations shift supervisor took actions to declare both channels of the toxic gas system inoperable. The technical specifications allow continuea reactor operation with an inoperable system and require that the system be restored to an operable status within seven days. The licensee also took actions to revise OP 4329 to establish chlorine trip setpoints at 5 ppm and to recali-brate both chlorine monitoring channels. Also, on July 24, 1987, department instruction DI 87-09 was issued to OP 4329 to correct the setpoint discre-pancies of all toxic gas channels. For example, the instruction established an acceptable trip setting range of 712-728 for vinyl chloride. The licen-see's subsequent actions on this matter to prevent recurrence of the above violation, on a generic basis regarding amendments to v.chnical specifications, will be reviewed during future a routine inspection. No further inadequacies were identified, t
7.0 Maintenance Activities The inspector reviewed the corrective maintenance actions taken on July 13, 1987 when the recirculation pump motor generator (M-G) rat A slip ring brushes were found arcing. A power reduction was initiated at 3:15 p.m., and when .
the reactor was operating at about 60% of rated load, operators secured the !
M-G set at 3:50 p.m. Upon completion of brush replacement, the recirculation I pump was started at 6:25 p.m. and was operating at maximum recirculation flow I at 7:30 p.m. The inspector observed the condition of the six generator brushes in the running M-G set on July 14, 1987, and reviewed the previous occasions of brush replacement as a result of corrective or preventive main-tenanc In orde to minimize unscheduled power reductions that could result from cor-rn.tive maintenance on M-G sets to replace arcing brushes in late 1984, the '
licensee instituted a new preventive maintenance actitity to replace the i brushes, based on wear measurements, during a scheduled power reduction or outage of sufficient duration. The brushes on M-G set A had last been re-placed in February 1987, and those brushes were expected to perform satisfac-torily until the refueling outage in August 1987. However, the licensee noted during their routine brush inspections that some brushes were exhibiting wear at on accelerated rate, and that such wear had become progressively worse until the arcing of one brush occurred on July 13, 1987. The inspector da-l termined that the licensee had been closely monitoring the degrading brush a condition as the outage neared, and that maintenance personnel had prepared the replacement brushes by grinding their surfaces to properly fit the gene- {
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18 rator rotor contour. When the arcing occurred, operations and maintenance personnel activities were well-coordinatsa to minimize the power reduction yet assure proper M-G set maintenance. No discrepancies were identifie .0 Spent Fuel Pool Reracking Activities The inspector reviewed licensee actions in progress on July 17, 1987 to com-plcte neutron testing of the boral loading in a high density spent fuel stor-age rac The storage rack was the first of two 10 x 10 PAR racks scheduled to be delivered to the site, and represents one alternative being pursued by the licensee to increase the storage locations in the spent fuel pool so as to preserve full core offload capabilities upon startup.from the refueling outage scheduled to begin on August 7, 198 The PAR racks would be installed in the pool per EDCR 87-406 if the PAR option is selected by the license The inspector witnessed the testing and reviewed the results for PAR Rack l Serial #21025. The tests were completed per OP 2430, High Density Fuel Rack l Boral Testing, Revision 5 dated June 5, 1987. Test results recorded on data l sheet VYOPF 2430.09 and strip chart recordings from Eberline RM-19 neutron
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detector channels showed that alternate cells in the 10 x 10 storage array were loaded with boron. The inspector ver:fied that test prerequisite condi-tions were met as required by the procedure, and that the requirements of radiation work permit 87-500 and fire control permit 87-258 were met. The
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inspector also reviewed the neutron testing activities and test results of the second PAR rack on July 29, 198 No inadequacies were identifie The licensee informed the inspector on July 17, 1987 that a stop work order had been issued by the licensee to the Nuclear Energy Services (NES) Company of Danbury, Ccanecticut to sta fabrication on the 18 x 20 NES storage rack being prepared for shipment to the site during July, 1987. The stop work order was issued based on the results of an inspection of the rack completed by NES personnel at the US Toal and Die fabrication facilities which identi-fied inadequate welds on a seam joining two individual storage cells. The stop work order was addressed in a letter from the licensee to NES dated July 17, 198 The licensee's stop work order and letter also identified the lic-ensee's concerns that need to be resolved prior to acceptance of the rac Additional licensee inspection plans of the rack will be formulated after further review of the process and the responses from NE The inspector reviewed the licensee's actions and found them to be appropriate to assure that a quality product is delivered to the sit This area will be reviewed further on subsequent routine inspections to verify an acceptable j resolution of identified quality concerns is obtaine !
9.0 Construction / Fabrication Quality of New RHR Pump Impellers Pursuant to the NRC Region I staff's request in a te'lephone conversation on June 2, 1987, the licensee submitted documents and correspondence covering the four new Bingham Willamette (BW) residual heat removal (RHR) pump impel-1ers for in-office review by an NRC Region I lead reactor enginee The new I
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impellers replaced the original impellers that were found eroded and cracke The cracks in the old impellers were evaluated by BW and believed to be due to original defects in the castings. Documentation for these original cast-ings was unavailable and there were no records of material supplier, mill test reports, repair welds or NDE testin j i
To assess the quality aspects of the new BW impellers the inspector reviewed the information provided by the licensee. The information contained: 1) mill test reports that showed chemistry and mechanical properties in accordance with ASTM A-743-CA15 and contractual requirements; 2) heat treatment of the castings to 1750 F normalize and 1275 F temper; 3) casting weld repairs and stress relief in accordance with appropriate procedures MSD-1 Rev. O and MSD-3
! Rev. 0; and, 4) magnetic particle inspections of the repair weld i l
Based on the telephone discussions with licensee personnel, the inspector i determined that the number of defects per casting ranged from two to five with '
the largest defect measuring 3" long x 5/8" wide x 5/8" deep. The inspector also determined that the weld repairs of three of the four impellers were made in the presence of the Yankee Atomic welding metallurgis The inspector concluded that the material traceability, the weld repairs and stress relief in accordance with procedures and the NDE testing provides i assurance of improved quality of the new impeller .0 Management Meetings Preliminary inspection findings were discussed with licensee management peri-odically during the inspection. A summary of findings for the report period
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i was also discussed rt the conclusion of the inspection and prior to report l 1ssuanc ,
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I TABLE A INOPERABLE EQUIPMENT DilRING INSPECTION PERIOD (see discussion in section 5.4) !
Component .Date Inoperable Toxic Gas Monitor A June 2, 1987
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Standby Gas Treatment A June 9, 1987 High Pressure Coolant Injection Pump June 10, 1987 (see section 5.8 for review details)
Recirculation Pump Motor-Generator Set B June 15, 1987 ;
Combustible Gas (H 2 -0 2 ) Monitor A June 16, 1987 l Toxic Gas Monitor A June 16, 1987 Toxic Gas Monitor A June 19, 1987- .
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Toxic Gas Monitor B June 22, 1987 Drywell Atmosphere Temperature Recorder June 24, 1987
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Toxic Gas Monitor A June 29, 1987 l Residual Heat Removal Pump D June 30, 1987 l l
Standby Gas Treatment B June 30, 1987 Toxic Gas Monitor B June 30, 1987 ;
West Switchgear Room CO 2 Fire Protection System June 30, 1987 (see section 5.3 for review details) i
Main Steam Line Isolation Monitor July 7, 1987 Local Power Range Monitor 08-09B July 8, 1987
Local Power Range Monitor 24-25B July 11, 1987 l Recirculation Pump Motor-Generator Set A July 13, 1987
.(see section 7.0 for review details)
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Tuxic Gas Monitor A July 15, 1987 Toxic Gas Motors A and B July 23, 1987 1 (see section 6.0 for review details)
Various LPRM Inputs to APRM "E" Various l (see section 5.7 for review details) ;
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TABLE B i
POTENTIAL REPORTABLE OCCURRENCES (see discussion in section 5.10)
Potential Reportable Occurrence Description Date l 87-22 Recirculation pump A tripped due to low oil June 3, 1987 l pressure as result of apparently. vibrated shut !
lube oil pump C suction isolation valv A&B Main steam drawer A failed to calibrate per June 3, 1987 procedur Breaker for control room chiller SCH-1 needs to June 12, 1987 be replaced as it passed its QE lifetime of 15 years in April 198 J 87-25 High steam line d/p instrument 13-84 as-found June 13, 1987 trip setpoint was 197.5 inches of water, which
exceeded the 195 inches of water limit of Tec Spec. Table 3. '
87-26 Latest revision of Offsite Dose Calculation June 18, 1987 Manual omitted the section referring to steam jet air ejector calculation High pressure coolant injection pump overspeed June 19, 1987 trip assembly failed freedom of movement tes Toxic gas monitor A vinyl chloride analyser June 25, 1987 failed to trip system during performance of logic tes Spurious initiation of west switchgear carbon July 9, 1987 dioxide fire protection syste Pending Missed fire hose surveillanc July 29, 1987 Pending Toxic gas monitor alarm limits set above Tec July 30, 1987 Spec. limits.
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