IR 05000271/1985040
| ML20141M980 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 02/20/1986 |
| From: | Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18030B368 | List: |
| References | |
| 50-271-85-40, GL-84-07, GL-84-7, NUDOCS 8603030183 | |
| Download: ML20141M980 (23) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 85-40 Docket N License No. DPR-28
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licensee: Vermont Yankee Nuclear Power Corporation RD 5, Box 169, Ferry Road Brattleboro, Vermont 05301 Facility Name: Vermont Yankee Nuclear Power Station Inspection At: Vernon, Vermont l Inspection Dates: November 26 - December 31, 1985 Inspectors: William J. Raymond, Senior Resident Inspector T omas Silko, Resident Inspector
Approved by: #. [. RO F4 t. E. Tribp, Chief, Reactor Projects Section 3A ' 9 ate Summary: Inspection on November 26 - December 31, 1985 (Report No. 50-271/85-40)
Areas Inspected: Routine, unannounced inspection on day time and backshifts by I the resident inspectors of: actions on previous inspection findings; plant shut-down operations, including pipe replacement activities; plant physical security; HPCI and service water pipe support deficiencies, and the failure of a recircula-l tion pipe whip restraint; personnel changes; potential RHR pump problems; and,
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licensee event reports (LERs). The inspection involved 112 hour0.0013 days <br />0.0311 hours <br />1.851852e-4 weeks <br />4.2616e-5 months <br /> Results: No violations were identified in the areas inspected. Operational status reviews of shutdown operations identified no conditions adverse to safety. The evaluation, investigation and repair of pipe support deficiencies warrants further action by the licensee and followup review by the NRC staff (Section 6).
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DETAILS
! Persons Contacted l
l 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 i
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Yankee Atomic Electric Company Mr. J. Hof fman, Engineer :
Mr. R. Oliver, Engineer Meetings were held with the Vermont State Nuclear Engineer on December 6 and 20, 1985 in the NRC Resident Office to discuss NRC inspection of outage acti- ;
vities and recent events. The status of licensee and NRC actions concerning i the recently identified hanger and whip restraint problems were reviewed, and i the status of the NRC staff review of the radioactive material in the North *
owner controlled area was also discussed. The meeting was beneficial for the ;
review of items of mutual interes ; Summary of Facility Activities i
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The plant remained in a maintenance outage during the inspection period while i activities to replace the primary re: circulation system piping continue Significant milestones achieved included the completion of the removal of the old recirculation system piping; shipment of the old recirculation system '
piping to the offsite low level waste burial sites; and, the beginning of the installation of the new recirculation system piping. The licensee's program and procedures to install, weld and inspect the new recirculation system were
. reviewed by regional inspection personnel (reference NRC Region I Inspection Report 85-41). Licensee activities under EDCR 84-402 also continued during the inspection period to modify hangers and supports as part of the seismic reanalysis progra During the inspection period, two hanger discrepancies were identified by plant workers while modifying piping supports, and a recirculation pipe whip restraint failed during the removal of a recirculation spool piec These discrepancies are discussed further in Section 6 below, i
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3 Status of Previous Inspection Findings 3.1 (Closed) Follow Item 80-BU-13: Status of Vibration Monitoring Syste By letter FVY 85-50 dated September 3, 1985, the licensee proposed to NRC:NRR that the vibration and loose parts monitoring system be removed
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from service. The justification for this action was based on the visual
! inspection of the core spray sparger clamp conducted during each refuel-ing outage from 1980-1985, which demonstrated that no degradation of the clamp occurred. 8y letter dated November 19, 1985, NRR accepted the licensee's proposed action. Based on the above, and since the system had degraded to the point that it no longer serves any useful purpose, the licensee intends to remove the system during the present outag The inspector had no forther comments on licensee use of the syste This item is close .2 (0 pen) Unresolved Item 85-25-06: Corrective Actions for Containment Electrical Penetrations. The licensee's evaluation of the electrical conductor insulation degradation in containment electrical penetrations was provided in a memorandum to the Technical Services Superintendent l dated December 16, 198 A corrective action pla'i was developed for implementation during the present outage. This item is discussed further in Section 7 below. This item will remain open pending completion of
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licensee actions to correct the penetration deficiencies and further NRC j review on subsequent inspectio ,
3. 3 (0 pen) Unresolved Item 85-08-02: HCU Support Modification. By letter dated December 12, 1985, the licensee notified NRC:NRR that the design of modified supports for the hydraulic control units (HCU) was complete, but that installation of the modifications would directly interfere with recirculation pipe replacement activities. The licensee proposed de-ferral of the HCU modifications for the present pipe replacement outage, and to complete the installation of the insert / withdraw line supports within 5 months following startup from the outage. The NRR staff re-viewed the licensee's proposal and found that the combination of infor-mation provided by the licensee and previous actions taken to upgrade scram system supports was sufficient to address staff generic concerns regarding pipe breaks in BWR scram discharge system By letter dated December 9,1985, NRR concurred with the plan to defer the modification This item remains open pending completion of the HCU support modifica-tions following the outage and subsequent review by the inspector.
l 3.4 (Closed) Follow Item 85-30-05: LER Submittal. Licensee event report 85-09 was submitted to the NRC on October 30, 1985. The inspector re-viewed the report and verified the report accurately described the event and its consequences. This item is closed.
l 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 l
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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, authoriza-nn, badging, escorting, personnel and vehicle searches.
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The inspector also reviewed the compensatory measures taken for excavation work completed within the isolation zone along the North protected area fenceline for the installation of power conduits on December 16, 1985. 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-
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tive requirements. Tours 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 Shif t 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 outa0e meetings to keep informed of the daily outage activitie Significant milestones achieved by the licensee included removal of the old recirculation piping and the start of installation of the new syste Plant activities and events that received further review are discussed belo .1 Worker Injury A contractor fell from 252 ft elevation to the 238 ft elevation of the drywell at 11:15 A.M. on November 27, 1985, while removing recirculation pipe interferences. The worker's fall was broken by a pipe whip re-straint above the 238 ft. elevation. The worker was assisted out of the drywell while his anti-contamination clothing was removed. No spread of contamination occurred. The worker was treated for bruises and con-tusions on his back and side at the site. Transport to offsite medical facilities was not required. The inspector interviewed the worker re-garding the circumstances of his fall and determined that no further actions by the licensee were warranted to preclude a similar occurrenc No inadequacies were identified.
! 5.2 LSA Survey The inspector reviewed a radiation survey of LSA box VY-P-8 conducted i
in the Reactor Building by health physics personnel on November 27, 198 The LSA box was loaded with piping spools from the old recirculation system that were being made ready for shipment to an offsite burial sit The inspector interviewed the technician and observed the survey in pro-gress. Contamination smears of the exterior of the package were satis-
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factory. Radiation dose limits on the surface of the package and at two meters from its surface were less than the 40 CFR 173 limits. No in-adequacies were identifie .3 Consumable Purge Dams The licensee evaluated the potential impact on subsequent plant opera-tions from the use of consumable purge dams during the installation of the new recirculation system. Specifically, the licensee requested as-sistance from the General Electric Company to determine whether the con-trol rod drive (CRD) inner filter clogging problem which occurred at the Monticello facility was applicable to Vermont Yankee. The evaluation by GE, documented in letter GE-VY-85206 and accepted by the licensee, concluded that the VY CRD mechanism design at Vermont Yankee is not sub-ject to the failure mode experienced at Monticello. The evaluation is summarized belo The Monticello facility experienced degraded scram performance when the CRDMs with moveable type inner filters attached to the base of the spud became clogged with fibrous materials from the consumable purge gas dam The control rods at Vermont Yankee use a later model CRDM with i
a stationary inner filter fixed to the top of the stop piston Since these filters do not require the passage of reactor coolant during a scram, clogging of the filters will not adversely impact normal scram performanc Based on the above, the licensee concluded that use of the consumable dams would not create a safety proble NRC staff review of licensee actions to minimize the use and impact of i purge dams is documented in Inspection Report 85-4 The inspector re-viewed the bases for the licensee's evaluation regarding scrim perform-ance and identified no inadequacie .4 Drywell Radiological Controls 5. Followup of Worker Exposure Concerns A contractor worker contacted the resident inspector by tele- !
phone at 9:25 A.M. on December 5, 1985 to express his concerns '
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that he was required to complete a drywell work assignment in a manner that was in violation of the principle that personnel exposures be maintained as low as reasonable achievable (ALARA). ,
The worker was employed as a fitter for the licensee's general I contractor, Morrison & Knudsen Co., but was fired on December ,
5, 1985, after refusing to wait in a radiation area inside the {
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Reactor Building while scaffolding was erected at the intended work location inside the drywel , The worker stated that he was assigned to work on the N2E nozzle safe-end under Radiation Work Permit (RWP) 85-4114 on December 4, 1985. He signed in under the RWP at 10:55 l
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after dressing in protective clothing (PC's) per the require-ments in the permit. The worker stated he went to the N2E work area inside the drywell and noted that he could not begin his work assignment until carpenters completed the construction of scaffolding at the job sit The worker estimated that, based on previous experiences, it would take at least I hour to prepare the scaffolding. The worker stated that he exited the drywell since the job was not ready, but was instructed by his supervisor to remain dressed in his PC's in a holding I
area in the Reactor Building. The inspector noted that the worker did not determine what the dose rates were from Health Physics (HP) personnel stationed in the area. The worker stated he refused to wait in the designated area even though he did not know what the exact radiation levels were, since he knew the area was a radiation area. The worker stated he intended
' to redress in his street clothes and wait in the Containment Access Building since that approach was consistent with ALARA principles to avoid needless exposure. The worker stated he signed off on RWP 85-4114 at 11:00 P.M., was terminated from t the job by 12 midnight for refusing to wait in the designated area, and left the site after completing an exit whole body count at 2:00 A.M. on December 5, 1985. The worker stated his work partner remained at the job, waited in the designated holding area for one and one-half hours, and accumulated 60 milli-Rems (mRems) exposure while completing the assigned work on the N2E nozzle.
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Based on the worker's description of the intended holding area, I the inspector determined that the worker was asked to wait in the dress-out area due East of the drywell personnel air lock, where personnel normally remove their PC's upon completion of l
i drywell work activities. The inspector toured the work site at 10:30 A.M. on December 5, 1985 to review radiological con-ditions and the controls established by the licensee. The in-spector was accompanied during this review by Mr. H. Bicehouse, a Radiation Specialist from NRC Region I, who was onsite during the week to review radiological controls for outage activitie The inspectors performed radiation surveys of the outer drywell ,
access areas with an NRC R0-2A survey instrument, and with l
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licensee PIC-6 and R0-2 survey instruments. The health physics controls established by RWP 85-4114 were reviewed and found r to be appropriate for and commensurate with the :ddlation hazards associated with the work assignment. The personnel ,
exposures and work times recorded on the RWD were consistent with the facts as stated by the worker. The radiological con-ditions in two general holding areas were assessed by the in-
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Area 1 - dress-cut area at East exit from drywell - General area dose rates were less than 1 mRev/br throughout the area, except for one location on the Worth end, which had a ticld l
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of 2 mrem /hr from an RHR system process line located out,ide the dress-cut area. Dose rates on the RHR line were 6 to 8 mrem /hr on contact.
l Area 2 - material holding area at equignent hatch Southeast
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exit from drywell - General area dose rates were 1 to 2 mrem /br, !
with contact dose rates of 10 and 20 mrem /hr on two items in transit out of the drywell and temporarily stored in the hold-ing area. The items were clearly marked as radicactive mate-rial and dose rates in the general area around them were in the 1 to 2 mRe:n/hr rang ,
Based on the above survey results, the inspector noted that the dose rates in the areas of interest were very low, and radiological conditions in the area did hot constitute a i
" radiation area". In contrast with the designated waiting
! area, the dose rates at the k25 job site were considerably ;
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hi0her (2500 to 8000 mrem /hr), and contributed most of the 765 ;
mRems exposure to J3 individuals who completed the work !
covered under RWP 85-4114. While it is true that no radiation exposure is better than even a little exposure, it is the in- ,
spector's conclusion that the directions to the worker to re-main in Area 1 for up to 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> did not cause a radiological concern, and was not inconsistent with good ALAM practice This conclusion was based on the very low dose rates in the waiting area, the amount of exposure eventually required to ;
l complete the work, the NAC quarterly exposure limits, a con-sideration of the reed to maintain production activities, and a recognition of the time is that usually lost in undressin0 '
and redressing in FCs. The inspector's curvey results and !
conclusions were discussed with the worker when he contacted the insgt. tor by phone on January 6,198 '
i No violations were identifie .4.2 Drywell Access Contro's The inspectors reviewad the rindiation controls and work prac-tices established by the licensea in the drywell access control t
area. During interviews on December 5,1985, the inspector ,
noted that not all workers questioned knew what the dose rates were in the watting area, but they did know that the informa-tion was available fron heelth physics and supervisory person-nel in the area. Th9 inspector interviewed health physics personnel and noted they were cognizant of dose rates and general radiological conditions at the job sites, the holding .
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area, and in the dress-out cre The need to advise workers regarding radiation hazards even in low dose rate areas was discussed with health physics personnel at the control poin The control of work activities in high dose ratt areas such as the drywell was reviewed during the period of 12/2-6/85, as documented in NRC Region I Inspection Report 85-3 Based on interviews with a sompling of 12 workers during that period, the inspector determined that health physics personnel in-structed workers regarding dose rates at the drywell job sites, as well as nhere to locate themselves during work activities to minimi:e exposures. Additionally, based on a review of work j
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activities inside the drywell on 12/16/85 and through periodic monitoring of drywell activitics using the CCTV system set up in the Reactor Building, the inspector noted that workers did not linger or spend idle time in areas with high dose rates, i
However, this matter will bc reviewed further on subsequent routine NRC inspections to verify drywell work activities are
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sufficiently controlled to minimize worker " slack time" in high radiation areas and to keep worker exposures as low as reason-ably achievable (IFi 85 40-01).
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5.5 Drywell Epoxy Coating
i During inspection tours inside the drywell, the inspector noted that peeling of the drywell epoxy coating in the upper elevations had pro-gressed to some extent beyond that observed previously (see Inspection Reports 83-14 and 83-17). The inspector estimated during this inspection that the epoxy coating had peeled away from about 20% of the drywell surface area between the upper containment spray header and the top of the biological shield, exposing the underlying drywell primer. The in-spector noted further that the Operations Superintendent had opened an Action Item to review the condition of the coating based on a recent
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licensee inspection of the condition of the epoxy coating and the mirror insulation.
' During a meeting with the Operations Superintendent on December 23, 1985,
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the inspector stated that this item would be followed on a subsequent inspection to further review the licensee's evaluation of the accept-ability of the condition of the coatin Specifically, the inspector requested that the licensee's evaluation address whether the following previous conclusions were still valld: (i) the underlying primer coating alone provides sufficient protection for the drywell shell; and, (ii) the peeled epoxy coating will not adversely affect the operation of safety systems or other functions important to pla'nt safety. The inspector further requested the licensee to consider the feasibility of removing the segments of peeled epoxy coating during the present outage
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to eliminate / reduce the amounts of material that can becono loose and transported within the drywel This item will be reviewed further on a su'a sequent routine inspection (UNR 85-40-02).
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6.0 Pipe Support Deficiencies The licensee made a four hour notification to the NRC Daty Of ficer per !
50.72.b(2)(1) on November 27, 19a5 to report two support failures identified by plant workers during ruutine work activities. A third pipe support failure reported to the resident inspector on December 23, 1985, involved a deadweight support on the service water system. The inspector observed each deficiency, inspected other supports and restraints of similar design, and held several meetings with licensee personnel to review the licensee's evaluation of the failures and the statas of the corrective actions planned. The pipe support deficiencies are discussed in detail in Attachment I. Summaries of the in-spection status on these deficiencies are described belo .1 Recirculation Whip Restraint On November 23, 1985, the licensee found that whip restraint R108, lo- ,
cated on the 28 inch diameter discharge pipe of the E recirculation line, '
broke away from its support plate when the piping was removed. Examina-tion showed that the restraint had not been properly weltied during in- l itial plant constructio The licensee identified 32 whip restraints installed on the recirculation piping. NRC inspectors visually examined 21 installed restraints and found additional discrepancies. None of the discrepancies were judged to ha significant enough to cause failur Licensee inspectors also examined the installed restraints. There wss i generally good agreertent between the two inspections, although each in-spection found some discrepancies which were u.11gu The licensee had previously completed a reanalysis cf the recircul.ation system piping under revised pipe break assumptions. This analysis con- !
cluded that 19 of the 32 whip restraints, including rt:str.sint R100, were not needed. Based on this result, the licensee has concluded that the '
improper installation of R10B would not have had an adverse effect on !
safety if a pipe break had occurred. Further, uaing as-butit infcrma-tion, the licensee perforced cn 90gineering evaluation of the whip re-straints, and 24 of the 32 restraints were found to be acceptable as 1 i An additioral evaluation of the 8 unacceptable restraints found that only restraint R108 would have exceedel allowable stress limits under pipe >
break conditions. Of the 8 restraints unacceptable as is, 4 will be re-paired and 4 will be removed and not be replaced (as previously planned).
Additional inspection for this area will be followed under the following ntabers:
IFI 85-40-03, Review of new pipe break study (GE Report 23AS478).
IFI 85-40-04, Review of licensee's as-built inspectio . . - . _ __ _ i
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IFI 85-40-u5 Review of resolution of whip restraint problem, including disposition of unacceptable restraints, repair of re-l straints, review of whip restraints in other systems, and review of other welding work by the same welding contracto .,
l 6.2 Concrete Eribedded Base Plate
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The base plate on pipe support MS-HD 22E, located on the HPCI discharge line, fell from the building column on which it was mounted when workers removed the support during modifications. Based on inspection, the lic-ensee concluded that the 2 shear lugs and 5 ancher bars had been cut off '
the back of the base plate and the base plate had been tack welded in place prior to pourim; the concrete column during initial constructio l At som tirte later, the tack welds faile ,
t The licensee analyzed the HPCI piping stresses with the as-found pipe si.pport information and concluded that the HDCI system would have re- l mained operable for all anticipated normal and abnormal operational load :
Dased on this analysis, the inspector concluded that no technical speci- '
ficatica LCO violation had occurred and no adverse safety impact had resulte t The licensee was reviewing the impact of the base plate failure on other base plates. The licensee deternined that 60 simil,ir pipe supports were installed with ettachaents, and a visual examination of them revealed j that none of these base plates had the concrete spalling around the base
- plate that the failed base plate had. The licensee was reviewing other l cvaluation techniques, including static load testing and infrared scan- l l nin The resolution of this base plate failure will be reviewed under uv i resolved item (85-40-06). -
6.3 Service Water Deadweight Support l
On December 23, 1985, deadweight support R5W-HD 164E, attached to 20 inch !
diameter service water line SW-12 in the reactor building, was found '
broken and not supporting its intended load. The licensee's preliminary l assessment concluded that there were no apparent adverse ef fects on the ;
service water line or other sdjacent supports. Inaccessible during plant i operation, the support's titra of failure was not know !
The resolution of this deadweight support failure will tia reviewed under l unresolved item (85-40-07). ,
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j 7.0 Followup of Previous Inspection Findings 7.1 Drywell Electrical Penetrations
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- Licensee investigation of the electrical problems with the A recircula- i
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tion pump suction valve identified a short between the suction and dis- {
) charge vs1ve position indication cables as they passed through drywell -
! penetration X105C. Since there were no protective insulators on the GE t l Type 237X680G016 penetration assembly, the interaction between the cables
- and the metal edge of the assembly was deemed the most probable cause '
1 of the observed breakdown in insulation properties. The inspector re- t j viewed the licensee's proposed corrective actions to address the issue, ,
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as documented in a December 16, 1985 memo to the Technical Services l
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! Licensee inspections of the six penetration types identified the same !
j conditions as that observed en X105C, and the same potential for insula- ;
{ tion degradation in the control rod position indication (X104A-C), 480 ,
i volt power (X1058, 0), neutron monitoring (X100A-D), and control and in- i
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dication (X105A, X105C, X101D, X102) penetrations. Five-KV power pene-trations X101A and X101C, and-thernoccuple penetration X103 are not sus-i ceptible to the failure mode and no corrective actions are required.
The inspector reviewed the conditions of penetrations X104A, X105A and '
} X105C during a drywell tour on December 18, 1985, and noted on each one '
l a large bundle of cables that were supported on the drywell side by only
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a closure box. This configuration allowed most of the weight of the 1 I cable bundle to press the lower conductors into the metal edge of the j
{ penetration sleev j i
l Licensee review of the four penetration types susceptible to the failure !
- identified a large quantity of spare conductors in the center of the i j penetrations. The number of spares will allow abandoning the conductors
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in the lower regions of the penetration, while leaving an adequate number i i of spares for future use. The abandoned conductors will also provide a ,
! buffer (Insulator) between the center conductors and the metal edge of I i the penetration sleeve. The abandoned conductors will be marked inside l the terminal boxes and on drawings to indicate they should not be used, j Switching from the outer to the center conductors will require reter-
! minating cables on both sides of the penetration. For example, 75 con- ,
ductors on both sides of penetration X102 will require retermination '
Seven installation and test procedures will be prepared per AP 6001 to ;
j direct completion and testing of the reterminations. The installation '
- procedures and licensee actions to reterminate conductors will be re- !
l viewed on a subsequent routine NRC inspectio !
The licensee reported this item under 10 CFR 50.73 as LER 85-1 The
! licensee was still reviewing the item for reportability under 10 CFR Part i
) 21. This item remains open pending subsequent NRC review of the licensee .
j actions as noted abov ;
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8.0 Staffing Changes The licensee announced the following staffing and organizational changes on December 17, 1985. Mr. D. Bauer was selected for the new position of State l Liaison officer at the corporate office. The vacated position of Engineer Assessment at the plant site would be filled following posting of the position.
! Mr. J. Babbitt was the former Security Supervisor and was selected as a Operations Training Instructor. Mr. J. Sinclair was the former Assistant to i the Vice President and was selected for the position of Security Superviso The inspector noted that the announced changes would not affect the organiza-i tional plan listed in the Technical Specifications. The staffing changes were l reviewed with the Plant Manager and the need to maintain continuity in the l
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Engineer Assessment function was discussed. Measures will be taken to assure continuity is maintained. No inadequacies were identifie .0 Review of Licensee's Event Reports The licensee event reports (LERs) listed in Attachment IV were reviewed in
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the NRC resident and regional offices. The reports were reviewed to verify the details of the event were clearly described; safety significance was identified; the event cause was identified and corrective actions taken (or planned) were appropriate; and, the report satisfied the requirements of 10 CFR 50.73. A review was also completed to determine whether generic implica-tions were indicated, and whether further followup review was warranted. The inspector had no further comments on these items, except as noted below.
l LER 85-13 was issued on December 26, 1985 to describe the inoperable pipe i
support and whip restraint identified on the HPCI system and recirculation
! systems, respectively. The licensee's description of the final corrective '
! action plan for the identified deficiencies was incomplete in the LER since the evaluation of these events were still in progress. During meetings with )
the Plant Manager and the Technical Services Superintendent, the inspector stated that the supplemental report for LER 85-13 should describe the correc-tive action for the deficiencies along with the re:ults of any further reviews of pipe supports and restraints. The inspector stated that the supplemental report should be submitted for NRC staff review of the completed actions at least 6 weeks in advance of the startup from the current outage. The licensee noted the inspector's comments. This item is unresolved pending submittal of the supplement LER and subsequent review by the NRC (UNR 85-40-08).
10.0 Potential RHR Pump Problems The inspector informed the licensee on December 5, 1985 of information re-ceived from NRC Region I regarding problems noted with residual heat removal (RHR) pumps manufactured by the Bingham Company. Inspections completed by utility personnel at another facility in November, 1985 identified wear ring l failures on two Model 18X24X28 CVIC Bingham pumps with about 10 years of ser-vice history. Failure of the wear rings can degrade pump performance and
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adversely affect the fulfillment of the safety function. Routina surveillance testing at the other facility failed to detect the degraded wear ring condi-tio The RHR pumps at Vermont Yankee are Bingham Model 16X18X26 CVIC pumps, which have been in service since initial plant licensing in 1972. The inspector requested the licensee to review this information for applicability to the ,
Vermont Yankee pumps to determine whether further actions are required to avoid a similar failure. This item is open pending completion of the licen-see's review and subsequent review by the NRC (IFI 85-40-09).
11.0 Management Meetinas 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 issuanc l l
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l ATTACHMENT I PIPE SUPPORT DEFICIENCIES l 1. Failed Recirculation Whip Restraint The first support deficiency reported by the licensee concerned the failure of a whip restraint on the B recirculation discharge line on November 23, 198 Whip restraint R10B, located at elevation 249 feet - 2 inches and azimuth 180 degrees of the drywell, became detached from its support plate as the 28 inch diameter discharge pipe was being removed from the drywell.
i I Examination of the whip restraint and its support mount revealed that the re-straint was not properly installed in accordance with the design drawings during original plant construction (reference Drawings 5920-424, Revision 5
, and 5920-425, Revision 5). As shown in Attachment II, the bracket plate de-
! sign included the use of a rectangular standoff, which was constructed of 1 l
inch X 1-1/2 inch bar stock, and used to allow field adjustment of the whip restraint hoop around the recirculation pipe during installation. The overall dimensions of the restraint base is 38 inches by 6 inches. The restraint failed because it was improperly welded to the bracket plate due to an incom-plete weld "B", as shown on Attachment II. The restraint was welded to the bracket plate along the right side and about two-thirds the length of the bottom side, and it was tack welded at two locations along the top side. The construction drawings required weld "B" to be a 1-1/E inch weld around the perimeter of the restrain The recirculation pipe whip restraints were de-l signed and furnished by the General Electric Company and installed by the Hartwell Company under contract with the plant Architect Engineer, Ebasc In order to determine the significance of the event, the licensee initiated plans to inspect all 32 recirculation system whip restraints, document and ;
review the as-built conditions, and evaluate the significance of the R10B 1 failure. The results of the licensee's and NRC examinations are discussed i further belo l The function of the 32 whip restraints initially installed on the recircula-tion system was to prevent the pipe from impacting the drywell shell in the i
' event of a rupture in the recirculation system. The piping analysis used when l the plant was first constructed assumed that pipe breaks were as likely to '
happen anywhere in the system. The 32 restraints were located on both loops such that the piping segments between the restraints could not impact the containment wal The pipe break analysis requirements subsequently changed in recognition of the toughness of austenitic stainless steel and the conclu-sion that a double ended pipe break is considered incredible. The licensee performed a new pipe break study for the recirculation system based on the requirements of NRC Generic Letter 84-07 and NUREG 0800 (NRC Standard Review Plan), Section 3.6.2. This analysis resulted in a reduction in the number of required whip restraints from 32 to 19. Restraint R10B was one of the re- i
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straints scheduled for removal, since it was not located at a nozzle connec-tion, high stress locction, or high fatigue location. Based on the above,
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I Attachment I 2 the licensee concluded that the improper installation of R10B could not have had an adverse impact on plant safety. The inspector reviewed the licensee's evaluation and identified no inadequacies. However, this item is open pending
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inspector review of the licensee's new pipe break ' study (GE Report 23AS478 -
Enclosure J to EDCR 85-01) for conformance with SRP Section 3.6.2 (IFI 85-40-03).
The resident inspectors and a specialist inspector from NRC Region I (refer-ence Inspection Report 85-41) conducted an inspection on 12/18/85 of 21 re-circulation whip restraints for conformance with the construction drawing In general, most restraint attributes on all those reviewed were found to conform to the drawing requirements; however, some discrepancies were noted, as described below. None of the discrepancies were as significant as that noted on R10B, and in the opinion of the inspectors, none would have resulted in a failure of the restraint.
The inspectors also noted that other hign energy lines inside the drywell are not restrained a a manner similar to the recirculation system. However, the feedwater and main steam lines are generally restrained by structural steel at two locations, near the 266 and 252 ft. elevations. Additionally, hoop-type restraints are attached to the four steam lines at the 252 ft. elevation, i and are bolted to cdjacent structural steel. The inspector reviewed the overall condition of the restraints and structural steel and noted that the l condition of bolting and welding was satisfactory. The items noted below were identified by the NRC inspectors and discussed with the Technical Services Superintendent on December 19, 1985, and/or with representativeR from the l recirculation task forc R6A&B - top side weld missing or weld size inadequate from restraint to standoff (and from standoff to bracket on R6B) R11A&B - top of restraint welded di ectly to bracket, top standoff bar missing i R4A - weld appeared undersized along bottom side of restraint R4A&B - 2 of 6 stiffening fins along the bottom side of both supports h'ad been notched out to provide an opening for jet pump in-strumentation lines R8A&B - weld size on all four sides appeared adequate, but the st6ndoff was missing and the restraint was welded directly to the bracke The R8 as-built conditions noted by the inspectors deviated from the required design details specified by Revision 1 of Drawing G191710 dated August 27, 1969. However, subsequent licensee review of construction records identified an approved change for the design to the present condition through Field Sketch 5920-FS-1330 dated August 20, 197 Drawing G1917710 was not updated following the redesign. Following an extensive review of available construc-tion field changes, licensee representatives were unable to identify docu-mented field changes to the approved design for the other discrepancie _ _ - _
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Attachment I 3 The above apparent discrepant items were discussed with licensee personnel for followup review and dispositioning. The inspector reviewed the licensee's initial inspection results documented in a December 19, 1985 memorandum. The inspection scope was comprehensive and well documented, and the results in-cluded findings not noted by the NRC survey. All NRC survey findings were compared with licensee inspection results. The inspector noted that there was generally good agreement between the licensee's and the NRC finding However, the licensee's inspection failed to note the discrepancies with R4A&B (item D above), and the missing weld on R68 (item A above). The licensee noted these findings for followup review. The adecracy of the licensee's inspection of the as-built restraint conditions will be further reviewed by the NRC on a subsequent inspection (IFI 85-40-04).
An engineering evaluation of the licensee's inspection results was provided in Recirculation Task Force Document #006820 MEM-PT-MIS dated December 23, 1985. NRC review of the evaluation was in progress at the conclusion of the inspection. In order to verify the adequacy of the Ebasco designs and the as-built conditions, the restraint mounting details were re-engineered using the same design requirements as the original. The design assumption was that the force on the whip restraint was obtained from the product of the pipe cross sectional area and the system internal pressure of 1050 psig. The re-quired weld size needed to withstand the pipe break force for each pipe size (12, 22 and 28 inches) was computed. The as-built conditions were then com-pared to the calculated required weld sizes, with safety margins in the ac-ceptance criteria established by AISC code requirements and other engineering considerations. The evaluation showed that 24 of the 32 restraints were ac-ceptable as is, with 8 being unacceptable when compared to the established criteria. Four (4) of the 8 that were unacceptable are required for the new recirculation system and will be repaired. Of the 8 unacceptable restraints, only R10B failed the minimum weld size criteria that would have assured code allowable stress requirements were me The following items are considered open in the NRC review of this area:
(i) completion of the NRC review of the licensee's engineering evalu-ation to disposition the as-built restraint conditions; (ii) completion of licensee actions to repair restraints R6A, R7A, R9B and R6B for subsequent use with the new recirculation system; (iii) completion of VY actions to review the conditions of any other whip restraints installed on piping inside or outside the drywell;
'
(iv) completion of an evaluation by VY of other welding work completed by the same contractor to verify no other discrepancies exist on other systems and component The above items collectively are considered unresolved and will be followed on subsequent routine NRC inspections (UNR 85-40-05).
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Attachment I 4 Concrete Embedded Base Plate The licensee reported on November 27, 1985 that during ne O ! cations to pipe support MS-HD 22E located on the 14 inch diameter HPCI discharge line to the reactor (HPCI ISB), the associated embedded base plate fell from the wall when it was cut away from the support. MS-HD 22E is an anchor support that was scheduled to be modified as part of the seismic reanalysis program during the present outage per EDCR 84-402. As an " anchor", the support provided the structural boundary between two piping analysis models, and the deadweight load of HPCI 158 was held by other supports. The planned modification of MS-HD 22E consisted of replacing the embedded base plate with a new base plate mounted over the embed and bolted to the reactor building wall. The modifi-cation was required since the embedded base plate was found by analysis to be inadequate to support the intended seismic load The embedded plate was nounted in a column in the Southeast corner of the reactor building. There are 7 columns in the reactor building. Concrete walls are filled with rebar for added structural integrity, and the columns especially have a high density of rebar. Inspection of the base plate and its imprint in the concrete revealed that the 2 shear lugs and 5 anchor bars (see Attachment III) had been cut off from the back of the plate during in-itial plant construction. A significant amount of spalling in the concrete around the plate was also evident, which indicated that the plate had moved due to vibrations in the HPCI line. The licensee concluded that construction workers encountered interference with the rebar while installing the base plate, and the embed bars were cut off to remove the interference. Based on the appearances of the plate, it was evident that the modified embedment plate was then tac welded in place until the concrete was poured to form the walls and column It is not known exactly when the support failed. The licensee assumed that the base plate became inoperable some time ago, including during periods when the plant was operating and the HPCI system was required to be operable. YNSD engineering performed an analysis of the line in the as found condition. The results of a preliminary analysis using the ADLPIPE computer code with ANSI B31.1 code criteria and Regulatory Guide 1.60 seismic input data were docu-mented in memorandum VYM 322/85 dated 12/18/85, which showed the maximum pipe stresses were as listed below. (See Inspection Report 82-23 for a summary of previous NRC staff review and acceptance of the licensee's ADLPIPE analysis capabilities.)
Equation 11 (deadweight + pressure)
Maximum stress = 0.11 allowable stress Equation 12 (deadweight + pressure + SSE seismic)
Maximum stress = 0.42 allowable stress Equation 13 (thermal)
Maximum stress = 0.12 allowable stress Equation 14 (deadweight + thermal + pressure)
Maximum stress = 0.09 allowable stress
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Attachment I 5 The license determined from the above results that HPCI ISB, and hence the HPCI system, would have remained operable for all anticipated normal and ab-normal operational loads. Based on the above, the inspector concluded that the deficiency with MS-HS 22E did not have and adverse impact on plant safety and no technical specification LC0 violation occurred. However, see the sum-mary section below regarding further information requested from the licensee on this ite The licensee and YNSD personnel completed a review of base plate installations and plant records to address whether similar conditions existed for other embedded base plates. The three types of embedded base plates in use at the plant are wall mounted, ceiling mounted, and speciality. Based on a review of drawings G191509, 511, 513 and 493, the licensee determined that there are 143 other base plates of the same design as MS-HD 22E installed in the Reactor Building. Based on a field inspection of the embedded base plates, the lic-ensee determined that 84 of the 144 have no attachments on them, leaving 60 with pipe support or conduit attachments. Eleven of the 144 are installed on building columns, and of these, 3 have deadweight supports attached, 3 have seismic attachments, and 5 have no attachment The licensee determined that the embedded plates with deadweight loads at-tached are most likely installed correctly since the support design load is the load applied to the plate. The largest deadweight load applied to an embedded base plate is about 4200 lb The licensee concluded that supports with deadweight loads could not be held in place with tac welds, since the maximum static load a tack weld could support was estimated to be 1000 lbs, and a tack weld would readily fail at a load much less than 1000 lbs if the load is cycled (as is the case when piping systems vibrate while operating).
Once the tac welds fail, concrete spalling at the edges of the plate would follow as the plate was free to move with the piping system. The licensee reviewed each base plate with attachments for evidence of spalling and found none attributable to plate movement (two base plates had some spalling that was attributed to modification activities associated with the plate).
Based on the above, the licensee concluded that failures of the type observed on support MS-HD 22E could be detected by visual inspection. The inspector reviewed the licensee's findings and concluded that the presence of spalling could aid in the identification of a failed support. However, the lack of spalling alone could not provide complete assurance that an embedded plate was installed in accordance with the design drawings, and further evaluation of the plates with other techniques would be require The licensee was reviewing alternate evaluation techniques at the and of the inspection period, including use of static load testing and an infared scanne The inspector observed a preliminary demonstration of the infrared technique on 12/31/85 on an embedded plate installed in the Reactor Building, and noted that the scanner was capable of detecting the presence of an enbed bar for a distance of about 10 inches into the concrete (the design length of_the embed bars is 15 inches). Efforts to refine the infrared scanning technique and procedure were in progress to enhance its detection capabilities. The in-
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Attachment I 6 spector reviewed the licensee's preliminary proposals for conducting the static load-test program. The inspector stated that a sample selection tech-nique acceptable to the NRC staff would be one which has a statistical basis and provides assurance of detecting defective base plates in the tested popu-lation at a 95% confidence level. The licensee noted the inspector's comment The licensee stated further that the acceptability of supports randomly selected for review would be evaluated through either infrared scanning, pull testing, and/or consideration of the applied deadweight load, as applicabl NRC review of the licensee's actions were in progress and the following items were considered open at the end of the inspection period:
(a) The licensee was requested to provide the actual load values and stress limits used to address the acceptability of the HPCI 15B line with MS-HD 22E in the failed condition; (b) The licensee was requested to tabulate the normal and peak loads on wall embedded base plates with attachments; (c) Development of techniques and procedures for infrared scanning and pull testing of base plates. Demonstrate that the sample selection and testing technique provides the desired confidence level for detecting defects; (d) Address the adequacy of the embed plates for those supports for which the deadweight load is much less than the design peak load; (e) Develop and document the basis for the conclusion that tack welds would not withstand cyclic loads attributable to operating vibrations in the process lines; and, (f) Address the acceptability of the ceiling and speciality embedded anchor
! plates.
! The above items are collectively considered an unresolved ite.a that will be reviewed further by the NRC on a subsequent routine inspection (UNR 85-40-06). Failed Service Water Deadweight Support The licensee notified the inspector on December 23, 1985 that deadweight sup-port RSW-HD 164E, attached to 20 inch diameter service water line SW-12, was found broken and not supporting its intended load. Line SW-12 is located in-side the reactor building, is classified as safety class 3, and forms a por-tion of the service water discharge return header from the Reactor Building Closed Cooling Water Heat Exchangers and the 480 volt uninterruptible Power Supplies. This portion of the service water system experiences normal service loads during routine power and shutdown operations. The licensee's prelimin-ary assessment concluded that there were no apparent adverse affects on the service water line or other nearby supports attached to SW-12 due to the in-
- - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _
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w Attachment I 7 operable hanger. Based on the above, the licensee concluded when the item was identified on December 21, 1985 that the discrepancy did not meet the reporting criteria in 10 CFR 50.7 The support was installed during original plant construction and it is not known when the support became inoperable. The support is not readily access-ible for inspection during normal plant operations. The support failed due to the failure of fillet welds on the support structural members. The de-ficiency was identified by contractor pt rsonnel while they were preparing to modify a nearby support as part of upgraies in progress per the seismic re-analysis program. Approximately 200 of 696 support modifications have been completed per the program during the current pipe replacement outage and no other failures of this type have been identifie (See Section 2 above re-garding a failed concrete embedded anchor plate). Support RSW-HD 164E was addressed as part of the seismic reanalysis program and was to become "in-active" upc? completion of modifications to adjacent deadweight and anchor support Initially, the licensee planned to simply " tag" the support as in-active and leave it connected to SW-1 The licensee now plans to remove the support from the lin The licensee will complete a reanalysis of line SW-12 in the unmodified, as-found condition to determine whether any stress limits were exceeded on the piping or adjacent supports with the inoperable support. Reportability of this item under 10 CFR 50.73 will be completed by the licensee pending com-pletion of the analysis. The inspector requested the licensee provide con-struction records for NRC review showing the intended design for support RSW-HD 164 This item is considered open pending completion of: licensee reviews of the support failure; calculations to address whether an overstress condition occurred; an assessment of reportability per 50.73; and NRC review of the as-built condition for conformance with the design drawings (UNR 85-40-07).
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ATTACHMENT II RECIRCUIL ATION WHIP RESTR AINT BRACKET PLATE
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ATTACHMENT III EMBEDDED BASE PLATE DESIGN FOR HPCI SUPPORT MS-HD 22 .
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) ~ ATTACHMENT IV
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LISTING OF LERs REVIEWED.
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LER N EVENT DATE REPORT DATE- SUBJECT
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85-01 01/02/85 02/11/85 Inoperable Sample Pump at Environmental Station >
85-02 01/28/85 02/26/85 Failure to Complete Quarterly Instrument Calibration 85-03 02/31/85 03/" ~ '85 Environmental River Samples Collected Late
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85-04 02/06/85 03/LE '8i5 Reactor Scram on 02/06/85
] 85-07 09/26/85 10/28/85 Type C LRT Test Failures
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85-09 10/07/85 10/30/85 Inadvertent Scram Signal While Shutdown '
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- 85-10 10/03/85 10/30/85 Containment Electrical Penetration Conductor
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Insulation Degradation l
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85/13 11/27/85 12/26/85 Inoperable Pipe Supports on HPCI Discharge and Recirculation Discharge Lines
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