IR 05000271/1986008

Insp Rept 50-271/86-08 on 860318-0505.Violation Noted: Failure to Completely Identify & Replace Namco Contact Blocks Installed in Plant Equipment Following Part 21 Defect Maint

ML20199L012
Person / Time
Site:	Vermont Yankee
Issue date:	07/01/1986
From:	Elsasser T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20199K999	List: IR 05000271/1986008 ML20199K991 ML20199L002
References
50-271-86-08-01, 50-271-86-8-1, NUDOCS 8607090292
Download: ML20199L012 (20)

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i U.S. NUCLEAR REGULATORY COMMISSION

REGION I

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Report N Docket N License No. DPR-28 Licensee: Vermont Yankee Nuclear Power Corporation

, RD 5, Box 169, Ferry Road Brattleboro, Vermont 05301 Facility: Vermont Yankee Nuclear Power Station i , Location: Vernon, Vermont '

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Dates: March 18 - May 5, 1986 Inspectors: Willia aymond, S ' r Resident Inspector

! Thoma ilk R dent Inspector Trainee

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-Approved by: / 7' I

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Thomas C. Elsassdh z gpnief, Reactor Projects Section 3C 'Date Inspection Summary: Inspection on March 18 - May 5, 1986 (Report No. 50-271/86-08)

_ Areas Inspected: Routine, unannounced inspection on day time and backshifts by i the resident inspectors of: actions on previous inspection findings; plant shut-l down operations, including pipe replacement and plant restoration activities; plant

physical security; concerns regarding worker fitness for duty; and, a worker's

, concern regarding the adequacy of personnel monitoring for radiation exposure.

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The inspection involved 213 hour0.00247 days <br />0.0592 hours <br />3.521825e-4 weeks <br />8.10465e-5 months <br /> i i Results: No violations were identified in 4 of 5 areas reviewed. Operational

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status reviews of shutdown operations identified no conditions adverse to safet A potential violation of 10 CFR 50.59 requirements discussed in section 5.1 con-l cerned the unanalyzed changes made to the seismic cooling tower cells in 1980.

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Licensee actions to replace the RHR " pup" piece and repair the cracks in the core

spray nozzles warrant further NRC review - section The licensee should assure

! contractor personnel closely follow security procedures for escorting visitors -

l section 7.0. One violation was identified related to the licensee's failure to completely identify and replace NAMC0 contact blocks installed in plant equipment

, following maintenance with defects of a type reported as unsafe in a Part 21 repor &

This item requires further NRC and licensee management attention - section 8.0.

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DETAILS 1. 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 Vermont Yankee Mr. J. DeVincentis, Engineer Mr. P. Donnelly, Maintenance Superintendent Mr. J. Pelletier, Plant Manager Mr. T. Trask, Engineer Mr. R. Wanczyk, Technical Services Superintendent Yankee Atomic Electric Company Mr. J. Hoffman, Engineer Mr. R. Oliver, Engineer Meetings were held with the Vermont State Nuclear Engineer on March 18, and April 1, 1985 in the NRC Resident Office to discuss NRC inspection of outage activities and recent events. The status of licensee and NRC actions con-cerning quality control problems with NAMCO parts, and the status of the NRC staff plans to review plant restart activities were also discusse . Summary of Facility Activities The plant remained in a maintenance outage during the inspection period while activities to replace the primary recirculation system piping continue Significant milestones achieved included the completion of installation of the recirculation and RHR system piping; completion of baseline NDE for the new piping; refilling the reactor vessel and removal of temporary equipment; and, the partial completion of induction heating stress improvement of re-circulation system welds.

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3. Status of Previous Inspection Findings 3.1 (Closed) Unresolved Item 85-25-10: Control Room Carpeting. The inspec-tor reviewed the carpet specifications in the Bill of Lading for Purchase Order No. 26356 dated January 20, 1986 and reviewed the markings on the carpet prior to installation in the control roo Geneva 2 carpeting provided by Tate Architectural Products, Inc. was installed, which was j tested for a fire rating in accordance with ASTM Standard E-84-70. The l licensee reported the test results to NRC:NRR by letter dated November 26, 1985. NRR reviewed the licensee's submittal and concluded that in-l

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. 3 stallation of the carpet would not craate a significant fire safety hazar The NRC safety evaluation was provided in a letter to the lic-ensee dated March 25, 198 This item is close .2 (Closed) Unresolved Item 85-25-08: Diesel Generator Fuel Oil Header Check Valve Replacement. This item was previously addressed in Inspec-tion Report 85-30. The licensee decided to not replace the check valves with ones having a seat material other than BUNA-N. Previous testing of the diesels has demonstrated that maintaining 85 inches of fuel oil in the day tank will keep the header filled and pressurized enough to achieve start times within the limits required by the Technical Specifi-cations. A day tank level of 85 inches provides 17 inches of oil head above the level of the engine injector headers. Adequate day tank level is presently verified 3 times daily as part of the auxiliary operator (AO) surveillance The next revision of OP 0150, which contains the A0's round sheets, will include a separate entry to verify at least 80 inches fuel oil in the diesel generator day tanks. This level is greater than the diesel gene-rator fuel oil header and is adequate to assure proper start tim Based upon the above, the decision to not replace the fuel oil check valve is acceptable. This item is close .3 (Closed) Unresolved Item 85-25-06: Drywell Electrical Penetration This item was previously addressed in Inspection Report 85-40 and 86-0 Corrective actions for the degraded electrical penetrations have been completed, which included the insertion of insulting material between the penetration conductors and the metal sleeves, or the retermination of bottom conductors to spare conductors in the center of the penetra-tions. The inspector reviewed completed work'on penetrations X103, X100A-D, and X104A- The inspector reviewed penetrations where insulation was added to verify insulating material was installed per the procedural instructions, and to verify that there was no visible damage to the conductors as a result of contact with the sharp end of the penetration sleeves. The inspection was conducted for two of the three penetration types receiving insula-tion - neutron monitoring and CRD position indication. The inspector verified that the retermination of conductors through penetrations con-taining 480 volt power, and control and indication circuits were com-pleted per the applicable installation and test procedures developed for the repair. The inspector noted that all reterminations received second person verification and 100% quality control revie The installation and test procedures developed for both the retermination and insulation processes were reviewed against AP 6001, Installation, Test and Special Test Procedures". The inspector noted that minor changes made to the installation and tests in the field were properly

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documented and in accordance with AP 600 Based upon the above, the licensee's corrective actions should prevent recurrence of shorting of conductors inside the drywell penetrations. This item is close .4 (Closed) Unresolved Item 85-14-05: Diesel Generator Fuel Oil Header Check Valves. Potential Reportable Occurrence (PRO) 85-31 was issued to evaluate the reportability of the degraded conditions of the fuel oi check valves. The item was found not reportable under 10 CFR 50.73 since the diesel start times remained acceptable with the degraded check valve because day tank level was maintained greater than 80 inches (see para-graph 3.2 above). The licensee determined this event was not reportable under 10 CFR 21 because the diesel vendor, Fairbanks Morse, had not identified any previous similar occurrences. The vendor concluded that the check valve is not vital to the proper performance of the engine, provided sufficient " head" is available through the fuel oil day tan Fairbanks Morse is not planning to notify other users of the check valve i

because of the determination that the valve is not vital to the operation of the diesel. At Vermont Yankee's request, the vendor provided a new part number for the valve using Viton seats should they choose to replace them. The inspector had no further questions regarding the reportability of the degraded check valves. This item is close .5 (Closed) Unresolved Item 85-20-01: Inspection of SFP Check Valves. The licensee completed actions to inspect SFP check valves V19-21A and 8 for leak tightness. The V21A valve was closed, but V21B was found to be not fully seate Actions were taken to isolate the "B" SFP cooling supply header to preclude siphoning the pool in the event of a rupture of the piping upstream of the check valve. Licensee actions are described fur-ther in Section 8 beloe. This item is close .6 (Closed) Follow Item 85-08-05: Post Fire Doors. The licensee took timely actions to label the fire doors between the east and west switch-gear rooms with a message that the doors should be kept closed. The doors have been found closed by the inspector during routine plant tour This item is close .7 (Closed) Follow Item 85-30-06: Licensing Actions for Cycle XII Operatio The results of the licensee's core performance analysis showed that no changes to the technical specifications are required for Cycle XII opera-tion. This item is discussed further in Section 8 below. This item is close .8 (Closed) Follow Item 85-25-07: Diesel Generator Brush Rigging. The licensee conducted the inspection of the brush rigging assembly on the diesel generators to determine whether failures were present similar to that noted at another facility. Although similarities exist between the licensee's and the other brush rigging, the smaller length of the rigging on the VY diesels should lessen the likelihood of a similar failur _ _ .. .

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i l The licensee's inspection verified visually that there were no crackr

_ present in the brush rigging assemblie The licensee determined that l replacement of-this item is not necessary. This item is close .9 (Closed) Follow Item 85-30-04: Elimination of Potential HCU Leak Path Revision 15 of OP 2111 changed the procedure to isolate and/or flush a

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control rod drive. The procedure instructions now require that isolation .

valve HCU-107 be left closed following depressurization of the HCU to i

eliminate a potential leakage path from the CRD to the reactor buildin This item is close .10 (Closed) Violation 83-33-01: Failure to Have a Station Procedure for

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the YAEC Van Mounted Whole Body Counting Syste The licensee subse-

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quently stopped using a van mounted system and used the system permanently

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installed at the station. In his response to the item.per FVY 84-15

dated 3/2/84, the licensee stated that OP 0533, " Body Burden Counting",

will be reviewed as necessary prior to the future use of any equipment not already addressed in OP 0533. Revision 5 of OP 0533, 11/4/85, con-j tained such instruction. The commitment has been met and this item is I close i

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3.11 (0 pen) Follow Item 83-33-02: ALARA Program. A Memo from Plant Manager to 83-33 file stated that the VY ALARA program was further formalized by a 12/20/85 revision of the VY Radiation Protection Policy. The policy

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provides a commitment to the principle of ALARA. It defines the level

'1 of ALARA reviews to be performed based on the estimated job dose, in-cluding jobs that must be reviewed by the ALARA committee. New proce-

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dures/ procedure revisions as necessary will be drafted to implement the l provisions.

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! 3.12 (Closed) Follow Item 85-30-08: IRM "B" Cracking. The licensee's evalu-

! ation of the SRM/IRM dry tube inspection results was provided in a memo-

!- randum from the YNSD Core Components group dated March 10, 198 Five

of the ten tubes contained defects that were serious enough to warrant

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rejectio The licensee replaced nine of the ten SRM/IRM dry tubes with

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an improved unit recommended by the reactor vendo The new intended dry tube for IRM "B" was damaged during installation activities and was

not used. In its place, a new old-style dry tube was installe Even though the new model tubes are less susceptible to stress corrosion at-l tack, the old-style tube has been demonstrated to be acceptable for use <

. for several cycles of operatio The licensee's evaluation in regard

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to the " denting" observed on the lower sections of the IRM "B" dry tube was that the marks had been in place for some time and may have been 4 caused during the removal of core " poison curtains" that were used during the initial core operating cycle.

i The inspector reviewed the circumstances that resulted in damage to the j intended IRM dry tube during installation activitie The inspector

determined that the event occurred due to personnel error and that the l licensee's corrective actions were adequate to prevent recurrence. This

item is closed.

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3.13 (0 pen) Follow Item 86-01-02: Disposition of NCR 86-07 - Defects in NAMCO Contact Blocks. The licensee submitted a report on March 28, 1986, per 10 CFR Part 21 regarding the defective contact blocks supplied by NAMC0 Controls, In The vendor filed a letter with NRC Region I dated April 16, 1986 which indicated that, based on his review of the matter, it was not necessary to report the damage identified by the licensee under Part 21. Additional defects were discovered by the NRC inspector in contact blocks installed in plant equipment and on the Stores shelf. Resolution of the hardware discrepancies is upgraded to an unresolved item. Pro-grammatic considerations are separately addressed in Section 8 belo .14 (Closed) Follow Item IR 85-03: Staffing and Training. In item 4 on page 3 of the referenced inspection report, the inspector noted that the lic-ensee intended to include laboratory QC in the department training pro-gram. The inspector noted that Rev. 9 of OP 0632 dated 9/13/85 incor-porated laboratory QC in department training per item 12.5 of VYDPF 0632.06 - Trainee Qualification Tracking form. This item is close .15 (0 pen) Unresolved Item 86-01-11: Baseplate Installation Discrepancie The inspector met with licensee and YAEC personnel to review the licen-see's actions and evaluations in regard to baseplates with gaps between the plate and the concrete wall or ceilin Field Change Request (FCR)

400 was issued on 1/14/86 for EDCR 84-402 to provide additional engi-neering instructions to construction personnel to achieve full bearing contact between the baseplate and the concrete surface. This item is discussed further in section 8.0 below and remains open pending comple-tion of the licensee's evaluation and subsequent review by the NR .16 (Closed) Follow Item 86-05-06: SLC Incident Plant Information Repor The licensee issued Plant Information Report 86-01 on 3/17/86, which was approved with the recommendations in the draft report previously review by the NRC staff. The recommended changes to OP 4203 were as described in the meeting with the licensee in NRC Region I on 3/19/86, and there-fore acceptable. This item is close .17 (0 pen) Violation 86-05-01: Inoperable SLC Syste The status of licensee actions taken to return the SLC system to an operable status in accord-( ance with the the commitments made during the Enforcement Conference on March 19, 1986 were reviewed to assure the system would be properly i tested and returned to service prior to reloading the reactor cor Core

! reload was scheduled for the Week of May 12, 198 The inspector noted, l in particular, that OP 4203 had been revised and was in routing for ap-i proval. The revised test method for the Squib valves would assure de-fects of the type previously detected would be identified.

l The inspector noted that licensee actions had also been completed to is-l sue the plant information report, and to submit a Part 21 evaluation to

, the NRC. This item remains open pending completion of licensee actions

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to meet the remaining commitments, and subsequent review by the NRC.

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. 7 4.0 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; verification of physical barrier integrity in the protected and vital areas; verification that isolation zones were maintained; and implementation of access controls, including identification, authorization, badging, escorting, personnel and vehicle searche The inspector visited the drywell access point on 3/21/86 and noted that security controls were established by the licensee in accordance with previous discussions with NRC Region I to meet the requirements of 10 CFR 73.55(d)(8).

The controls had been established since 3/20/86. The inspector interviewed the guards to verify they were familiar with the general requirements of the post orders, and to verify that post orders were provided and covered during watch turnover. No inadequacies were identifie .0 Review of Outage Activities Plant tours were conducted routinely during the inspection period to review activities in progress and verify compliance with regulatory and administra-tive requirements. Lurs of plant areas included the Reactor Building and the Drywell. Radiation controls were reviewed to verify access control bar-riers, postings, and posted radiation levels were appropriate. Plant house-keeping conditions and the control of hot work were verified to be in accord-ance with the requirements of AP 004 Shift logs and records were reviewed to determine the status of plant conditions and changes in operational statu No inadequacies were identifie The inspector attended daily outage meetings to keep informed of the daily outage activities. Significant milestones achieved by the licensee included removal of the old recirculation piping and the start of installation of the new system. Plant activities and events that received further review are discussed belo .1 The licensee notified the NRC per 50.72 of his discovery on March 21, 1986 that structural modifications completed in May, 1980 on the West Cooling Tower had never been analyzed. Cells #1 and #2 of the West tower are a seismic Category I, safety class structure used as the ultimate reactor heat sink. Since the plant was operated in an unanalyzed condi-tion since 1980, the item was reportable under 10 CFR 70.72(b)(2)(i).

Operability of the ultimate heat sink was not required for the plant conditions at the time the item was discovered. The ultimate heat sink seismic capability will have to be assured prior to reloading the reactor with irradiated fuel and operating the plant above 212 degrees Routine corrective / preventive maintenance for the towers consists of re-placing the Douglas-Fir wood as required due to weakening or softening with age and the high inoisture environment. Additional modifications, L

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termed the E-fix, were completed in cell No. 1 and 2 in 1980 to take load off the outside support columns in the "C" bent which were weakene Companion columns were mounted adjacent to existing ones and lateral support bracing was added to shift load to outside columns. The old columns were left in place, and the combination of the old and new columns was stiffer than was previously assumed in seismic analyses for the structure. The 1980 work was done under a maintenance request and was not recognized as a modification. Thus no safety evaluation or analysis was complete The deficiency was discovered by YAEC personnel while doing preparation work for the current PDCR. Design changes on the tower are in progress during the present outage to add diagonal support braces and box in existing ones as required to account for changes in PVC fill material and provide for margins to load limits required by newer analysis method Analyses completed for this effort will account for the present structure with E-fix modifications plus intended changes. Work will be completed

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under a PDCR with the associated safety evaluatio The licensee plans to complete an analysis of the West tower in the 1980-1985 configuration to determine whether it would have withstood an earth-quake with the companion columns. It appears that present analysis for PDCR shows some over stress condition would occur for diagonal member The changes made to the tower structural supports in 1980 changed the facility as described in Figure 12. 2-31 of the FSA The failure to analyze a change to the facility as described in the FSAR is contrary to the requirements of 10 CFR 50.5 Licensee actions on this matter were in progress at the conclusion of

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the inspection and will be followed on a subsequent inspection. This item is unresolved pending (i) completion of licensee actions to restore the safety class cooling tower cells to an analyzed, operable condition prior to plant startup; and, (ii) completion of an NRC review of the licensee's analysis results for the towers in the 1980-1985 configuration to determine the appropriate enforcement actions (UNR 86-08-01).

5.2 Shroud Head Bolt Examinations The inspector reviewed GE Service Information Letter (SIL) 433 dated February 7, 1986 concerning vendor recommendations to examine shroud head bolts for crack The SIL was received by the licensee and assigned to the Maintenance Department for review and dispositioning by April 15, 1986. During a discussion on March 20, 1986, the Maintenance Supervisor stated that UT examinations would be completed on the shroud head bolts during the present outage. Spare bolts are available onsite should cracked bolts be identified. The licensee subsequently completed an examination of the shroud bolts per the SIL recommendations and reported to the inspector that no cracks were identifie No inadequacies were identifie _ ._

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5.3 Drywell Work Activities General conditions and wo'rk activities,in progress were noted during an inspection tour of the drywell on 4/9/86. Housekeeping conditions were acceptable, but in need of attention.to improve cleanliness overal The inspector noted that this condition had been noted by licensee man-agement and actions were taken to improve housekeeping and trend its progress. - The inspector interviewed a fire watch covering hot work in progress on the 238 ft elevation for a support on the B recirculation discharge bypass valve. The fire watch was knowledgeable of his duties and the requirements of the fire control permit. No inadequacies were identifie .4 RHR System Piping Changeout The licensee notified the inspector on 3/17/86 that a section of stain-less steel piping, 18 inches long and 24 inches in diameter, was identi-fied on line RHR 28 just upstream of the RHR 46A check valve. Further nondestructive examination of the line was required to determine whether the piping section should be replaced due to susceptibility to inter-granular stress corrosion cracking (IGSCC).

The section of piping should have been included in the scope of pipe replaced during the recirculation pipe replacement outage. However, this spool was not recognized by the piping contractor as a candidate for replacement because the carbon steel-to-stainless steel transition weld was not shown on drawing 5920 FS 143B, which only showed field weld The licensee stated that the weld is scheduled for periodic inspection by inclusion in the ISI program. However, the bimetallic weld was not included in the augmented ISI inspection program completed per IE Bul-letins 82-03.and 83-02. The welds on either side of the bimetallic joint were excluded from inspection in the augmented program based on IGSCC weld susceptibility study. The licensee subsequently decided to replace the piping section with carbon steel material to eliminate its suscept-ibility.to IGSC This item is unresolved pending further NRC review of licensee actions to (i) replace the RHR spool piece; and, (ii) determine why the bimetal-lic weld was not included in the augmented ISI program (UNR 86-08-02).

5.5 Core Spray Nozzle Cracking The licensee notified the inspector on 4/28/86 of the results of non-destructive examinations on the core spray nozzles (NSA and N58) which showed crack indications. The inspections were done as a followup to a recommendations made by the NRC staff in IN 84-41. The first indica-tions of cracks were noted on the outside diameter of the nozzle follow-ing liquid penetrant examination. Subsequent UT examination using the P-scan methodology identified circumferential indications over about 25%

of the circumference of the 8 inch diameter B nozzle. Axial cracks were

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identified on each nozzle. The UT indications on the nozzle inside diameter align with the PT indications on the outside diameter within the uncertainty of the triangulation techniques. However, the licensee reported that visual inspection of the nozzles showed no evidence of through wall leakage or previous leakage (staining).

The joint consists of the following materials: P3 alloy carbon steel vessel nozzle with stainless steel cladding; inconel 182 weld butter; and an Alloy 600 inconel safe end welded to the butter pad with Inconel 82 weld wire. The NDE indications occurred in the Inconel 182 weld but-ter on the nozzle. The licensee stated that repair options were being reviewed with NRC:NRR and that a full strength weld overlay repair in accordance with the requirements of ASME Section XI, paragraph IWB-3641 was the most likely corrective action pla The inspector noted that the NRR staff had previously reviewed the same repair method with another utilit This item is unresolved pending NRC review of the licensee's identifi-cation of the problem and corrective action plan. and licensee actions to complete the core spray nozzle repairs prior to plant startup from the current outage (UNR 86-08-03).

5.6 MOV Wiring and Steam Tunnel Conditions During a review of work activities on 4/9/86 to reconnect the electrical power to RCIC valve 13-15 per control wiring diagram (CWD) 1188, the in-spector noted that the worker wired the incoming control power lead in a manner that was the electrical equivalent of the CWD specifications, but not exactly as shown on the CWD print. Additionally, the external cable connections to the open and closing torque switch was wired through the limitorque terminal block, rather than directly to the switches.

, The torque switches were wired in a manner electrically equivalent to the CWD specificatio The inspector discused this observation with the Assistant Maintenance Foreman and noted that the licensee was aware of the practice and con-sidered it acceptable as an activity within the " skills of the trade",

and since independent review of the work and post connection functional testing per OP 5220 would verify the valve operated properly. The in-spector noted the above and identified no inadequacies in the licensee's position, but did express concerns regarding potential problems that could arise when field conditions are not exactly as reflected in tha control diagrams. The licensee stated that this matter would be reviewed further. The inspector will review this item further on a subsequent inspectio The inspector completed an inspection tour of the steam tunnel on 4/8/86 to note general conditions; observe work activities in progress and radiological controls; and, inspect the general condition of equipment

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and structural steel, including the adequacy of restraints on the steam and feedwater lines. No inadequacies were identified, except as noted belo The following apparent discrepancies were discussed with the Maintenance Supervisor on 4/8/86 for followup and corrective action, as necessary:

(i) bent and damaged louver panel on the steam tunnel access door:

(ii) questionable weld attachment on the end of one of two struts for the A feedwater line expansion guide; and, (iii) missing 1 foot section of seal around the outboard end of the penetration bellows for the D steam line. The licensee's evaluation and followup action for these items will be reviewed by the inspector on a subsequent routine inspectio .0 Concern Regarding Fitness For Duty The following item regarding the alleged consumption of alcohol prior to duty was identified to the NRC staff by workers at the plant and were reviewed to verify licensee activities assured worker safety and regulatory requirements were me NRC Region I received information from an anonymous caller on March 14, 1986 which alleged that contractor health physics (HP) personnel would arrive for duty at the site under the influence of alcoho The NRC Duty Officer noti-fied the licensee regarding the call so that measures could be taken to moni-tor personnel arriving for the March 14-15 midnight shift. The information relayed by the Duty Officer included a description of two suspect individuals as provided by the allege Licensee security personnel and a Health Physics Supervisor monitored con-tractor workers as they arrived for the mid-shift and identified no one who appeared under the influence of alcohol. Two individuals meeting the de-scription provided by the alleger were observed coming in for work. The in-dividuals were interviewed by the HP Supervisor and no evidence of alcohol was detected. The work activities of the technicians were monitored by the licensee and no unusual behavior was note The inspector met with security personnel on March 17, 1986 to review the licensee actions and findings on this item. The licensee concluded no per-sonnel arriving for work on March 14, 1986 were not fit for duty or under the influence of alcohol. The licensee instituted additional measures following March 14, 1986 to increase the monitoring of incoming workers for fitness for dut The inspector noted that the licensee has a policy regarding fitness for duty and a program for assuring personnel arriving for work at the plant are not under the influence of alcohol. The licensee's program includes the use of the guard force to monitor ircoming workers. The guards are trained on how to detect alcohol related problems, and are instructed on what actions to take if a worker is suspected of being under the influence. The licensee also has a program to monitor for alcohol and drug abuse that is implemented as part

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1 of the pre-employment screening program for prospective contractor employee The screening is done per AP 0800 in conjunction with pre-employment physical exams for contractors. The licensee also initiated a random drug and alcohol screening test program for contractors during the present outag No inadequacies were identifie .0 Concern Regarding Unmonitored Exposure The following item concerning an alleged violation of personnel monitoring practices for radiation exposure was identified to the NRC staff by a former worker at the site and was reviewed to determine whether licensee activities were appropriate to meet regulatory requirements and protect worker safet An ironmaker telephoned NRC Region I on March 14, 1986 to discuss concerns he had regarding his visit at the Vermont Yankee site on February 13, 198 While visiting the site on that date to take a welding test, the individual was admitted as a visitor and was taken to welding test booths inside the containment access building, and subsequently to a lunch room onsite. The individual received no General Employee Training (GET) and no personal radi-ation dosimetry was issued to him. The welding test was taken in a location that was about 15 feet from a fenced area that was posted as a radioactively contaminated area. The worker stated his escort also left him in a lunch room onsite from 2:45 P.M. until 6:45 P.M., from which he could see workers leaving the contaminated area. The worker was concerned because the control of visi-tors at Vermont Yankee was different than he had experienced elsewhere where GET and radiation badging were performed prior to entry onto the site area Additionally, the worker was concerned that he had received an exposure to radiation for which there was no recor The resident inspector interviewed the worker by telephone on March 21, 1986 to obtain additional details regarding his activities during the February visit and a description of the exact locations visited in the plant. The inspector also toured each area and conducted radiation surveys. Based on this review and the survey results, the inspector determined that the worker was not at any time in a radiation, contamination or airborne radioactivity area. Radiation dose rates measured by the inspector in the welding test area and the lunch room on March 21, 1986 were less than 0.1 mrem /h Based on this measurement, periodic tours of the area, knowledge of the licensee's use of the containment access building, and the observation by the worker on the day of his visit that there were no posted radiation areas within the test area, the inspector concluded that the worker most probably received no occu-pational exposure during his visit to the Vermont Yankee site on February 13, 198 NRC regulation 10 CFR 20.202 requires the licensee to supply personnel moni-toring equipment to individuals that enter the plant restricted area only if he is likely to receive an exposure in excess of 25% of the allowable occupa-tional limits specified in Part 20.101. Based on the above, the inspector

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determined that no regulatory requirements regarding the control of radio-logical exposure were exceeded in regard to the licensee's treatment of the worker as a visitor on February 13, 198 However, based on the worker's account of his activities onsite on February 13, 1986, the inspector determined that security requirements possibly were not met in the manner that the individual was escorted as a visitor. The worker stated that he was escorted to the welding test area, the Coke lunch trailer and the lunch room. However, his assigned escort left him while in the lunch room and the Coke trailer. Although other badged personnel were in the areas with him, the worker was not aware of who had responsibility for him as an escort for those periods. This matter was referred to the licen-see's security organization on March 25, 1986 for review and followu The licensee reviewed this mar,ter with the contractor responsible for the conduct of the welding test. The inspector reviewed the results of the lic-ensee's and the contractor's reviews, which were documented in memoranda dated March 26, 1986 and March 27, 1986. The contractor's procedure for handling visitors was reviewed. If followed, the procedure would keep an individual under escort control at all times and thereby meet the requirement of the security plan. The contractor personnel who conducted the test on February 13, 1986 could not recall the specific worker in question, who was one of five people who took tests on that day. The contractor could identify no reason to indicate that the normal escort procedures were not followed for the worke The inspector discussed the review results with the licensee's Security Super-visor. The inspector concluded that it could not be determined conclusively whether the individual was properly escorted at all times while onsite on February 13, 1986. It is probable that a one-on-one escort control was not maintained for the four hours the worker spent in the lunch room, but his actions and whereabouts were possibly under surveillance by other badged per-sonnel. In any case, the worker did not enter either security sensitive or radiologically controlled areas. The inspector expressed his concerns re-garding the need to assure visitors are properly controlled. The licensee acknowledged the inspectors comments. The control of visitors will be re-viewed further by the inspector as part of future routine inspections of the security at the sit .0 Followup on Previous Inspection Findings 8.1 Followup of Item 85-30-06: Licensing Actions for Cycle XII Operatio By letter FVY 86-23 dated March 13, 1986, the licensee reported the re-sults of the Cycle XII Core Performance Analysis summarized in YAEC-150 The licensee stated that his review pursuant to 10 CFR 50.59 determined that Cycle XII operation does not involve an unreviewed safety question and there was no need to change the station technical specification The inspector reviewed YAEC-1507, " Vermont Yankee (VY) Cycle XII Core Performance Analysis Report", and discussed the analysis methodologies and conclusions with the Reactor and Computer Engineering Superviso The analysis was performed using techniques and computer codes previously

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reviewed and accepted by the NRC staff. The inspector verified that the analysis results showed that the margins of safety provided by the limiting condition for operation in the present technical specifications were not reduced. Specifically, the present technical specification requirements (through licensee Amendment No. 92) for shutdown margin, linear heat generation rate and MCPR operation limits were conservative when compared to the analysis results. However, no stability analysis was completed for Cycle XII. This matter is addressed further belo No inadequacies were identifie The inspector noted that the station technical specifications were changed by Amendment No. 92 issued on March 27, 1986 to delete requirements re-lated to the recirculation systems header cross-tie value These values were deleted from the system design during replacement of the recircula-tion system in the present outag The inspector noted further that licensee actions this outage to replace the recirculation system per EDCR 85-01 and 10 CFR 50.59 met the require-ments of NRC Confirmatory Order issued on August 28, 1984. Replacement of the recirculation system using material resistant to IGSCC obviated the need for enhanced leakage surveillance measures. By letter FVY 85-121 dated December 13, 1985, the licensee summarized his basis for and plans to discontinue use of the moisture sensitive tape leakage de-tection systems and the interim, restrictive coolant leakage limit No inadequacies were identifie Inspection item 85-30-06 is considered closed based on the above action The inspector noted that the present technical specifications allow reactor operation only when both recirculation loops are operabl By letter dated March 12, 1986, the licensee submitted a proposed change to the technical specifications to allow single loop operatio The proposed specifications are presently under review by the NRR staf Based on discussions with Reactor Engineering personnel, the inspector noted that the licensee anticipates having the revised specifications in place in time for startup from the current outag Revisions to plant operating and surveillance procedures for single loop operation are in progress but not yet complete. In letter FVY 86-23 dated March 13, 1986, the licensee committed to administratively implement core stability monitoring requirements even if the proposed technical specifications are not issued in time for Cycle XII startu This action is necessary to meet the NRC staff positions issued by Generic Letter 86-02 dated January 23, 1986, " Technical Resolution of Generic Issue B-19 Thermal Hydraulic Stability". This item is unresolved pending completion of licensee actions to obtain technical specifications for single loop operation prior to starting Cycle XII, to issue single loop operating proce-dures, to establish controls for stability monitoring for Cycle XII operation, and subsequent review by the NRC (UNR 86-08-04).

8.2 Followup of Item 85-20-01: Spent Fuel Pool (SFP) Check Valves. SFP check valves V19-21A&B were radiographed to verify leak-tightness of the valve seats, and thus assure that the valves would act as a siphon break to preclude draining of the SFP in the event of a break in the pump dis-

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. 15 charge lin The section of piping containing the valves is seismically qualifie The inspection was done because the valves are not part of a periodic inspection program and had not been previously examine Both check valves were radiographed on 3/13/86 and the developed films were reviewed onsite on 3/17/86. Vl9-21A was closed tightly against its seat and thus acceptable. V19-21B was found stuck open by about 1 inch off its seat, and thus could potentially allow back leakage through the valve and draining of the spent fuel pool. An evaluation of this sce-nario and the licensed operators' actions to mitigate such an event are provided in Inspection Report 85-2 The licensee took additional corrective actions on April 8, 1986 by closing the header manual isolation valve, V19-208, that is located physically adjacent to the check valve. This action was accomplished through Switching & Tagging Order 86-494 issued for the shift superviso A memorandum attached to the tagging order from the Senior Operations Engineer identified the reason for the order and the need for the system configuration. The licensee plans to operate the spent fuel cooling system with the "A" header only, pending completion of a design change in the spent fuel pool that will raise the discharge spargers above the top of the fuel. The design changes are tentatively planned to be com-pleted later in 1986 in conjunction with actions to expand the capacity of the spent fuel poo The licensee reviewed the item for reportability and concluded it was not reportable since the pipe line containing the valve is seismically qualified, and credit can be taken for pool level instrumentation and operator actions to mitigate an event. The licensee is planning addi-tional inspections of valve V19-21A to assure its continued operabilit The licensee evaluated the adequacy of SFP cooling with only one sparger in service and concluded it would be acceptable. The inspector noted that the spent fuel pool temperatures were maintained well below the technical specification limit of 150 degrees F with only the "A" header in opera-tion. This item is close .3 Followup Inspection Item 86-01-02: NAMCO Contact Blocks. The resident inspector reviewed the receipt inspection documentation for P0 20145 completed when the material was received in 1983. The documented in-spection results contained weaknesses previously identified and corrected as programmatic inadequacies (reference NRC Region I Inspection Renart 85-11). Based on a review of the forms and an interview with the re-ceiving inspector, the inspector determined that the 35 contact blocks from line item 5 of the order, along with material from 7 other P0 line items, were inspected on a sampling basis, and no discrepancies were identified. In particular, none of the 15 contact blocks with discre-pancies were identified by the random selection method used to inspect the material. The blocks were inspected through at least one plastic wrapping, a sealed closed plastic ba , . _ ,

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The inspector reviewed other records of receipt inspections and deter-mined that the blocks were not inspected again following the initial inspection in 1983. Specifically, the blocks were not included in the program for "second" receipt inspections completed in response to the Inspection Report (IR) 85-11 findings, since they did not meet either of the following two criteria: (i) material was selected for "second" receipt inspection only if the material appeared to be sealed closed in the original (opague) shipping packages - the blocks were in transparent packages on the shelf and not in the original cardboard box shipping containers; and (ii) material was selected for a second inspection if -

it was withdrawn from stores during the period from April - June, 1985 for use in the plant. The blocks were first withdrawn from stores for use in 1986. However, receipt inspection records show that materials from P0 20145 line item #2 (top cover gasket kit) and line item #4 (con-tact carrier) were receipt inspected a second time on April 12, 1985 apparently due to the sealed packaging of the material on the shel No discrepancies were identifie No other inspections were performed until the blocks were remcved from stores for maintenance on the MSIV switches. The inspector concluded that, if only a few of the 35 blocks were inspected in 1983, it is pos-sible that none of the 15 with defects wer', inspected. Further, defects of the type later identified could have been missed since the receipt inspections were done through the plastic bag Nonconformance Report (NCR) 86-07 was issued on 1/16/86 to address the material discrepancies and apparent vendor QA problem. The NCR was closed on 3/19/86 following actions to remove the defective blocks (15)

from Stores and the MVIV switches; return the defective parts to the vendor; return the damaged parts to the vendor for credit; and, request YNSD to perform a vendor QA evaluation and a Part 21 evaluation. Un-damaged blocks were accepted by the plant following inspection by I&C Q The licensee notified the resident inspector on March 26, 1986 of a con-cern associated with NAMC0 contact blocks that was reportable onder 10 CFR Part 21. Licensee technicians identified defects (insulator cracks and chips) on NAMCO contact block kits withdrawn from stores to perform routine preventive maintenance on the main steam isolation valve (MSIV)

position indicating and RPS limit switches during the present refueling outage. Subsequent examination of all kits in stores identified that 15 of the 35 contact blocks received under purchase order 20145 were  ;

defective. It was particularly notable that the broker, pieces from the chipped contact blocks were not in the sealed shipping bags, which indi-cated that the kits were not damaged by handling during shipment, but were shipped by the-vendor with the defect The licensee's Part 21 evaluation of this item concluded that, if the contact blocks had been installed without discovery of the defects, a potential safety hazard could have been created by failure of the

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switches in service, which in turn could have resulted in a loss of RPS trip inputs or loss of position indication on the MSIVs. The defects could have resulted in a failure of an RPS scram function when require Additionally, the licensee concluded that, without knowing the cause of the defects, it could not be stated with certainty that apparently non-defective contact blocks would not develop similar defects in servic The licensee concluded that other facilities using the contact blocks should be notified of the defects and the potential safety hazar The above findings by the licensee raised concerns regarding the NAMCO QC program. Actions were initiated by the licensee to request a vendor inspection by the YAEC QA organization. A vendor visit was completed under QA surveillance 86-038 on April 9-10, 1986. No inadequacies in the vendor QC program were identified by the licensee's QA organizatio At the licensee's request, the vendor reviewed 1018 contact blocks from the manufacturing process at a point prior to final QC inspection and packaging, and identified one block with the type of defect identified by the licensee. The vendor's review included samples from several manufacturing lots. By letter dated April 16, 1986 to the NRC Regional Administrator (Region I), NAMC0 Controls stated that, after reviewing the information from Vermont Yankee, they believed the contact blocks in question were damaged either in handling after packaging, or at in-stallatio The vendor further stated that there was no need to notify other users of the blocks, nor was there a need to report the damage identified by Vermont Yankee under Part 21. In regard to the vendor's conclusion that the blocks could only have been damaged after final packaging, the inspector noted that this assertion was not supported by the NRC inspector's observations in January, 1986 of damaged blocks in the manufacturer's sealed closed packaging with no broken pieces in the packages. The acceptability of the NAMCO QC and final packaging controls is a matter that requires further NRC review (UNR 86-08-05).

The inspector interviewed licensee personnel and reviewed the status of

licensee actions in Report to NCR 86-07 as of March 26, 1986. Work on the 16 switches installed on the inboard valves (MSIV 80A-D) was com-pleted per MR 85-1118 on January 20, 1986 to refurbish 13 of the switches and install 3 new switch assemblies from stores. Work was in progress per MR 85-1117 on the outboard valves (MSIVs 80A-D) when the defects were first noted and work activity was halted pending further review of the replacement contact blocks in Stores and those already installed on the outboard valves. Two contact blocks with defec'.s were identified and removed from valve V2-86B. Thirteen (13) adM tional defective contact blocks were identified in Stores, for a total of 15 defective blocks that were sent back to the vendor. Twenty (20) of the original 35 obtained under P0 20145 were accepted for use. Thirteen of the 20 were used to complete the work under MR 85-1117 for the outboard valves. The 7 re-maining contact blocks were returned to stores. As of the end of March, the licensee had completed the work on the switches. The switch assem-

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blies had been reinstalled on the outboard valves, and the switch assem-

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blies for the inboard values were stored in the Reactor Building pending completion of the recirculation pipe replacement activities inside the drywell .

The inspector examined materials from P0 20145 in Stores which included 7 contact blocks and contact carriers. Additional defects of the type reported in the Part 21 report were identified. Based on this finding, the inspector requested that the housing covers be removed on the MSIV switches to allow an inspection of the contact blocks. The findings, based on inspections completed during the course of the inspection period were as follows:

1. of Components Stores with defects / type 4 contact carriers 1/ crack 7 contact blocks 1/ crack 4 switch assemblies 0 Field 16 blocks - inboard valves 6/ chips 16 blocks - outboard valves 2/ chips The types of defects noted by the inspector were either minute cracks or small chips in the bakelite material None of the defects noted on installed switches appeared to affect the integrity of the contract blocks or the contact function. Additionally, none of the defects in the bakelite occurred near energized terminals. Further licensee evalu-ation was required to determine the appropriate actions to disposition the finding The inspector met with the I&C Supervisor, the Maintenance Supervisor and the Plant Manager at various times during the period to discuss the inspection findings and to follow the licensee's action The inspector expressed his concerns regarding the failure by plant personnel to iden-tify all defects of the type reported as unacceptable, and the failure to increase the scope of the review to all similar material (i.e., bake-lite components supplied by NAMCO) in the plant and in Stores. The in-spector requested that the following items be addressed by the licensee:

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an evaluation of the bakelite blocks / carriers presently installed in components with safety relates applications to determine whether the parts are acceptable for use or should be replaced to assure the intended functio reinspection of material in stores from P0 20145 to assure all defective material is identified and removed as necessar .

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obtain assistance from the vendor INAMCO) to identify the root cause for the observed defects. and,

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obtain further information from YAEC engineering regarding the basis

.for the conclusion presented in the Part 21 report that (i) the de-fects were caused by rough handling before packaging, and (ii) the -

observed defects would probably result in loss of safety functio Licensee actions on this item were in progress at the conclusion of the inspection period and included working with the vendor to disposition the identified defects, and to complete an engineering assessment of how the blocks would have functioned in the as found condition. Followup actions also included re-inspection of all contact blocks and carriers used in safety related applications at the site, including the main steam isolation valves and the RPS switches in the turbine stop valves. The inspections will be completed pending receipt of replacement parts from the vendor. This item remains unresolved pending completion of licensee actions to assure the RPS and position indication switches are operable prior to plant startup. These actions will be tracked under inspection item UNR 86-01-0 The failure by the licensee to complete adequate corrective actions to identify and control potentially defective material to prevent use or installation is a violation of 10 CFR 50, Appendix B, Criteria XV and XVI (VIO 86-08-06).

8.4 Followup Item 86-01-11: Baseplate Discrepancies. The inspector met with licensee and YAEC engineering personnel on 3/21/86 and 4/9/86 to review the status of the licensee's evaluations and actions regarding installed baseplates with less than full bearing contact (gaps) between the plate and the concrete surfac The licensee completed a walkdown of plates with gaps identified by NRC inspection. Instructions had been issued to construction to fix some of the defects (e.g., RSW H160). A review to determine the basis for accepting some amount of gap on different hanger configurations was still in progress. Preliminary results indicated that, for hanger baseplates which handle deadweight loads only (struts), spacing up to 1/8 inch is of no concern for either compression or tensile loads acting along the axis of the support. The acceptability of gaps on base plates that see moment loads in addition to deadweight is dependent on how the plate was designe For those base plates that were designed with hand calculations, there was enough conservatism in the design such that the bolt pull-out loads are insensitive to gaps up to 1/8 inch. However, for plates designed by finite element computer modeling, further review by NSD was required to see if the models take credit for the stiffness afforded by the por-tion of the baseplate that may not have full bearing contact with the concrete. Depending on how the plate was modeled, for those plates de-

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signed by computer analysis, it cannot be concluded whether the 1/8" gap is acceptable, since the margins normally available from the hand calcu-lation are no longer available. Either additional analysis or construc-tion fixes would be required for these case The inspector discussed with the licensee the need to assure field per-sonnel are aware of the type of information presented at meeting, so that workers are aware of need to report as discrepancy any gaps in frame type support The licensee acknowledged the inspector's comment This item remains open pending completion of licensee actions as noted above and subsequent review by the NR The inspector informed the'lic-ensee that a specialist in this engineering discipline from NRC Region I would review this item during a subsequent inspection at the sit .0 Management Meetings Preliminary inspection findings were discussed with licensee management peri-odically during' the inspection. A summary of findings for the report period was also discussed at the conclusion of the inspection and prior to report issuance.

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