IR 05000213/1987012

Insp Rept 50-213/87-12 on 870506-0609.No Violations Noted. Major Areas Inspected:Plant Operations,Radiation Protection, Fire Protection & Security.Unresolved Item Open Re Effect on Safety Limits

ML20215L840
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	06/17/1987
From:	Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20215L785	List: IR 05000213/1987012 ML20215L782
References
TASK-03-06, TASK-3-6, TASK-RR 50-213-87-12, IEIN-86-096, IEIN-86-96, NUDOCS 8706260189
Download: ML20215L840 (13)

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U.h NUCLEAR REGULATORY COMMISSION

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l REGION 1 Report N /87-12' '

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License No'. DPR-61 ,

Licensee: Connecticut Yankee Atomic Poker Company, P. O. Box 270  ;

Hartford, CT 06101 )

Facility: Had m Neck'P16M., Haddam, Connecticut j

Inspection at: Haddam Neck Plant I Inspection Period: May 6 - June 9, 1987

Inspectors: Andra A. Asars, Resident Inspector Paul D. Swetland, Senior Resident Inspector {

i Approved by: OSc bM 6/8747 E. C. McCabe, Chief, Reactor Projects 3B Date Summary: Inspection 50-213/87-12 (5/6/87,- 6/9/87)

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i Areas Inspected: This was a routine safety inspection (136 hours0.00157 days <br />0.0378 hours <br />2.248677e-4 weeks <br />5.1748e-5 months <br />) by the resident )

inspectors. Areas reviewed included plant operations, radiation protection, fire !

protection, security, maintenance, surveillance, licensee events, open items from previous inspections, NRC Information Notices, Emergency Preparedness Drills, and l switchgear building constructio Results: No violations were identified. One previous violation related to inade-quate procedure controls was closed (see Detail 4.,2). Two open inspection items concerning maintenance and corrective action program implementation (see Detail 4.1) and reactor coolant system temperature recorder discrepancies (see Detail 4.3)

were reviewed but require further licensee action before closure. One unresolved (

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P item was opened regarding the effect on the safety limits; as defined in Technical Specifications, of design changes which'use up the marginibetween established con-ditions and safety limits (see Detail 3.2). m

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I TABLE OF CONTENTS P_a.g.e, S umma ry o f Fa c i l i ty Ac ti v i ti e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 ,

i 1 Review of Plant Operations........................................... 1 Plant Operations Review Committee............................. ....... 2 i Followup on Previous Inspection Findings............................. 3

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4.1 Maintenance and Corrective Action Programs...................... 3 4.2 Inadequate Procedure Contro1.................................... 4 4. 3 Reactor Coolant System Temperature Recorder Discrepancies. . . . . . . 4 Followup on Information Notice 86-96, Heat Exchanger Fouling......... 5 1 Followup on Events Occurring During tha Inspection................... 7 !

6.1 Licensee Event Reports.......................................... 7 i 6. 2 Containment Isolation Valve Leakage............................. 7 i 6.3 Inoperable Phones............................................... 8 6.4 Reactor Coolant Pump Oil Leaks.................................. 8 General Electric HGA Relay Seismic Qualification..................... 9 Review of Periodic and Special Reports............................... 9 Switchgear Building Construction..................................... 9 9.1 Ground Subsidence Near Diesel Generator Fuel Storage Tanks...... 10 9. 2 Unauthorized Anchor Bolt Substitution........................... 10 9.3 Uncalibrated Vibration Monitor.................................. 11 10. Emergency Preparedness Dri11......................................... 11 11. Unresolved Items...................................................... 11 12. Exit Interview....................................................... 11 i

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1 DETAILS 1. Summary of Facility Activities

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On May 5 plant power was increased to 100% from 95%. (The 95% power level had {

been maintained for several weeks for turbine control system performance I evaluation.) Short duration load reductions to 50% were conducted on May 9 and June 6 for the addition of oil to the reactor coolant pumps (RCPs) and g for turbine control valve testing. On June 8, increasing bearing temperatures -

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on the No. 2 RCP forced another power reduction to 50%. A leak in the RCP lubricating oil system was identified and repaired. Oil was added. RCP bearing temperatures returned to norma The plant was returned to full power i and remained ther l j

2. Review of Plant Operations The inspector observed p? ant operation during regular tours of the following plant areas:

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Control Room --

Security Building  ;

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Primary Auxiliary Building --

Fence Line (Protected Area) 1

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Vital Switchgear Room --

Yard Areas

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Diesel Generator Rooms --

Turbine Building

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Access Control Point --

Intake Structure and Pump Building Control room instruments were observed for correlation between channels and for conformance with Technical Specifications. The inspector observed various alarm conditions which had been received and acknowledged. Operator awareness and response to these conditions were reviewed. Control room and shift man-ning were compared to regulatory requirements. Posting and control of radi-ation and high radiation areas was inspected. Compliance with Radiation Work Permits and use of appropriate personnel monitoring devices were checke Plant housekeeping controls were observed, including control and storage of flaramable material and other potential safety hazards. The. inspector also

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examined the condition of various fire protection systems. During plant tours, logs and records were reviewed to determine if entries were properly made and communicated equipment status / deficiencies. These records included operating logs, turnover sheets, tagout and jumper logs, process computer printouts, and Plant Information Reports. The inspector observed selected aspects of plant security including access control, physical barriers, and personnel monitoring. In addition to normal working hours, the review of-plant operations was conducted during midnight shifts, weekends, and holidays on the following day May 19, 1987 5:00 AM to 6:00 AM May 25, 1987 4:00 PM to 8:00 PM

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June 2, 1987 12:00 AM to 1:00 AM i June 8, 1987 4:00 AM to 6:00 AM 4 No unacceptable conditions were identifie Operators were alert and dis-played no signs of inattention to duty or fatigu !

3. Plant Operations Review Committee (PORC)

The inspector attended several Plant Operations Review Committee (PORC) meet-ings. Technical specification 6.5 requirements for required member attendance were verified. The meeting agendas included procedural changes, proposed changes to the Technical Specifications (TSs), design change packages, and !

field changes to design change packages. The meetings were characterized by ;

frank discussions and questioning of the proposed changes. In particular, consideration was given to assuring clarity and consistency among procedure Items for which adequate review time was not.available were postponed to allow committee members time to review and commen Dissenting opinions were en- ,

couraged. Except as follows, the inspector had no further comment i l

3.1 A special PORC meeting was held on April 7 to review Plant Design Change l Request (PDCR) 870, Reactor Coolant System (RCS) Component Support Modi-fications. This PDCR resulted directly from Safety Evaluation Program 1 (SEP) identification that three areas of the RCS could be overstressed i during the Safe Shutdown Earthquake (SSE). The affected areas include i the pressurizer surge line, pressurizer truss, and the RCP supports, j During the upcoming outage, the support and hanger installation in these j areas is to be strengthened. Also, it had been discovered during PDCR ')

processing that the pressurizer truss was not installed in accordance !

with the design drawings. This structure is designed to provide support to the pressurizer. The PORC members requested an additional review to l

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determine if this was reportable per 10 CFR 50.72 or 50.73. Site and !

corporate engineering personnel performed the evaluation and concluded l that this situation does not fall under the reportability requirement j Although the truss end connections'were not installed in accordance with !

the design drawings, the licensee does not consider the truss to be seriously degraded because adequate support for the pressurizer was still provide A lack of traceability to original design calculations pre-vented the licensee from determining if the as-built configuration placed the plant in an unanalyzed condition or outside the design basis. The licensee concluded that this situation is not reportable. The inspector discussed this decision with plant engineers and, based on the determina-tion of as-built adequacy, had no further question . 2 On May 8, the inspector attended a special PORC meeting for review of PDCR 841, RCS Average Temperature (Tave) Increase'to 562 Degrees This design change will modify the plant Tave program and plant operations so that nominal steady state RCS Tave is 562 degrees F at 100% rated power instead of the current value of 557 degrees F. The need for this change resulted from the design change, instituted last refuling outage, which relocated the RCS cold leg temperature indicators to the discharge

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I of the associated reactor coolant pump. This relocation improved the accuracy of the bulk RCS temperature measurement; previously the indi- ,

cated Tave had been 557 degrees F, but the actual RCS Tave was slightly !

higher. During this cycle, with the measured Tave at 557 degrees F, tur-bine efficiency has decreased and operations at full power require the turbine control valves to be in non-optimal positions to compensate for the reduced steam header pressure. The PORC concluded that the Inte-grated Safety Evaluation associated with this PDCR is adequate and that-this PDCR does not involve an unreviewed safety questio The cycle 15 reload analysis and TS change requests have been submitted to NRC with assuming that the RCS Tave will be increased per this PDCR before startup after the upcoming refueling outage. The inspector ques-tioned whether a revision to the current operating cycle (14) reload analysis was necessary and the effects of a higher RCS cold leg and average temperature on the accident analyses and plant operating safety limits. The licensee stated that, when the cycle 14 reload analysis was submitted to NRC, margin was allowed to compensate for the uncertainty involved when the actual RCS Tave was estimated. Originally, the licen- i see had expected RCS Tave to be 560 degrees F. The actual value of 562 J degrees F is within the margin provided and therefore the reload analysis would not be affected by any changes in the specified Tave. PDCR 841 documented minor nonconservative changes in power distribution and shut-down margin which resulted from fuel depletion due to plant operation at a higher than assumed Tave during previous cycles. With regard to the accident analyses, the licensee concluded that these changes would have no adverse effects on the analyse [

The licensee also stated that this PDCR would not cause the plant to f exceed the margin of safety as defined by TSs, but that these changes !

will absorb some of the operating margin to the safety limits. In re- l sponse to further inspector questions, the licensee described the prac- ;

tice of allowing PDCRs such as this to absorb operating margin if that !

does not affect the safety limits. This practice and the margin associ-ated with this PDCR remain unresolved pending further NRC inspection )

(Unresolved Item 213/87-12-01).

4. Followup on Previous Inspection Findings 4.1 (0 pen) Inspector Follow Item (213/85-21-10): NRC to review licensee pro-gress in improving the maintenance and corrective action programs. This item was opened as a result of identified deficiencies in the completion j of Authorized Work Order (AWO) form Specifically, there was a lack j of details describing equipment at-found condition, actual maintenance I work performed, and identification of the root cause of the failure or d malfunction. At the tine that this item was initiated, the licensee was !

instituting improvements in the maintenance and corrective action pro- W grams. A Region I inspector reviewed this area several months later, identified a continuing need for improvement in this area, and brought this to the attention of plant management. This was documented in NRC Inspection Report 50-213/86-19. During the current inspection, several 3 l

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AW0s which were worked during the last six months were reviewed. They involved corrective and planned maintenance on an auxiliary feedwater pump, the diesel engine driven fire pump, the.feedwater regulating valves, ..

and Reactor Coolant System loop isolation valves. Several of these did l not contain adequate descriptions of the work performed or the root cause .s of the problem. The' inspector noted that all of these AW0s were either i planned or involved. minor off-normal equipment' conditions.such as valve packing leaks and erratic valve operation. This item will remain open pending further NRC review of AW0s associated with equipment malfunctions and failure .2 (Closed) Violation (213/87-01-01): The co'ntrol provided.by ACP 1.2-6.4, Temporary Procedure Change, over changes and substitutions of test equip-ment was not adequate in that it permitted test. personnel to change, modify, and/or substitute test methods without any supervisory or tech ~

nical review, and without documenting these changes, modifications, or substitutions on the test procedure. The licensee responded to this violation in a letter dated March 10, 1987, and committed to revise ACP 1.2-6.4 to prevent recurrence. 'The licensee implemented these commit-ments immediately following the NRC inspection. The procedure changes state that substitutions can only be made with equipment of equal of better accuracy and appropriate scale, that only a Level 2 technician or higher can determine instrument equivalence, and that any instrument substitution must be noted in the applicable portion of the procedur The inspector verified that these changes were properly. incorporate I This item is close j 4.3 (0 pen) Unresolved Item (213/87-08-02): ' Licensee determination and'NRC review of why the cold leg temperature (Tc) recorder paper scales did not match the recorder scale, how long this situation has existed, and 4 what corrective actions will be taken. Following discussions with the inspector, the licensee initiated measures to'make notations on the af-

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fected charts and to purchase correctly scaled paper. It was' determined that the scale on the paper had been incorrect for Tc since startup from the last refueling outage (May 1986) when a Plant Design Change Request q (PDCR) was performed which relocated the Reactor Coolant System (RCS)

temperature detectors and required rescaling of the recorders. .To.cor- J rect the misleading records, operations personnel provided the Nuclear Records department with the correction equation to be'used to convert'

the recorded teniperature to actual RCS temperature'and instructi.ons t .j make this notation on all of the affected Tc charts in records. . The :'

inspector. verified that the equation was accurate. The licensee also initiated a purchase order (PO) for paper with the correct scale During processing of the P0 it was identified that'this paper was already 1 onsite and in storage in the Project Assignment'(PA) warehouse (not th l plant warehouse). This has been the case _since September 1986. Upon: ]

location of the paper, it was installed into the two. recorders which record only Tc. The other two instruments record both RCS pressure and'

temperature. In order to change the paper for these recorders,'the,lic-ensee processed Temporary Procedure Change-(TPC)87-084.. This TPC pro- j i

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I vides for recalibrating the pressure recorder to correspond with the new pape The recalibration was performed on June 7, 1987. The inspector noted that all four recorders, including the dual purpose recorders, are being replaced during the upcoming refueling outage with new recorders dedicated to RCS temperature. RCS pressure recorders will be relocated to the panels behind the main control boar The inspector had no fur-ther concerns with these corrective action The licensee has completed their investigation into the circumstances which allowed the recorder paper to remain incorrect for nearly a year although the PDCR had been closed out. It was concluded that plant per- 3 sonnel were aware of the deficient paper but an aggressive and formal follow up was not ccnducted to ensure timely delivery and installation of the correct paper. A review was also conducted of nine PDCRs which ;

have been completed in the last year to determine if this was an isolated )

instance of unaggressive and incomplete resolution of PDCR open item l Two PDCRs were identified, each with an open item which was not clearly i resolved before closeout of the PDCR. These PDCRs involved an Appendix !

R modification and the electrical portion of installation of the new I radwaste reduction building. The licensee is currently evaluating re-solution of these two item A review of the procedure governing PDCRs, ACP 1.2-3.1, Preparation, Review and Dispositioning of PDCRs, was performed in conjunction with i the PDCR review. A memorandum issued to the Engineering Supervisors I clarified the definition of open items and the manner in which these I items are to be tracked and resolved. A supervisory review of all open ,

items before turnover of a design change has been instituted as a secund verification that these items will not adversely affect system operabil-it These provisions are temporary pending institution of a new pre- 1 operational test manual for PDCRs. This program is expected to be im-plemented before the upcoming refueling outage; including the trainin of plant personnel. The licensee stated that the manual will include guidance on system turnovers, preoperational testing, and a methodology for tracking and resolution of open items and deficiencies remaining at turnover and/or identified during preoperational testing. This item will remain open pending program institution and review of PDCRs conducted under the progra . Followup on Information Notice (IN) 86-96, lleat Exchanger Fouling Licensee action on the following Information Notice was reviewed for forward-ing to appropriate management, licensee review for applicability, response timeliness, response appropriateness, response accuracy, corrective action commitments, and corrective action completio .

IN 86-96, Heat Exchanaer Fouling Can Cause Inadequate Operability of Service I Water Systems This notice alerted licensees to the potential for fouling of heat exchangers in raw water systems. This fouling can be in the form of mud and silt buildup and corrosion and could seriously degrade the ability to reject heat. Licen-see Controlled Routing (CR) 86-2169 was issued in response to this IN and a review was conducted to determine its applicability to Haddam Neck. The CR response adequately addressed the current surveillance and maintenance proce-dures which ensure that flow is present in heat exchangers cooled by river water. However, the response did not address the aspect of verifying that sufficient flow is present to provide adequate cooling in accordance with the system design basi At Haddam Neck the service water system (SW) circulates water from the the Connecticut River through heat exchangers in the residual heat removal (RHR)

system, containment air recirculation (CAR) system, and emergency diesel generators (EDGs). Due to fluctuations in river level and flow, the plant experiences' silt buildup in heat exchangers cooled by SW. Currently, main-tenance procedures associated with the CAR and EDG systems provide for clean-ing the heat exchanger tubes on a periodic basis. The RHR heat exchangers are not cleaned in this manner because they are primarily cooled by component .,

cooling water with SW as a backup. SW is introduced into the RHR coolers once ;

each refueling outage to verify the flow path. The coolers are then flushed 1 with primary wate The inspector reviewed Preventive Maintenance Procedure (PMP) 9.5-43, Emer-gency Diesel Cooling Water Heat Exchangers, performed semi-annually or as required to clean, test, and inspect these heat exchangers. If, during the monthly surveillance run of the EDGs, lube oil temperatures exceed acceptance criteria, the maintenance department evaluates the test data to determine if the coolers' require an additional cleaning. The inspector noted that the frequency of this PMP has recently been increased from annually to semi-annually in response to the heat exchanger outlet valves being blocked open in their fail-safe positio Previously, the valves had been closed and re-ceived a signal to open when the EDGs started. With the previous configura-tion, SW passed through the heat exchanger tubes only when the EDGs were run-nin After it was identified that the valves would not fail in the safe position, they were blocked open. Now, SW passes through the tubes continu-ously. This has resulted in increased tube fouling. Upon further questioning, the licensee stated that the coolers are still able to support extended diesel j generator operation when fouled to the point at which cleaning is require Also, modification of the valve solenoids to provide for fail-open operation is scheduled to be accomplished in about one mont The inspector had no further questions on this ite The CAR heat exchangers are serviced during each' refueling outage in accord-ance with Corrective Maintenance Procedure (CMP) 8.5-96, Maintenance of Con-tainment Recirculation Fan Coolers. This includes inspection for leaks, cleaning, repair as necessary, and pressure testing for leakage. Flow through I

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these heat exchangers is also verified quarterly under a new surveillance, SUR 5.7-118, Service Water Penetration Check Valves; This procedure verifies the operability of the SW check valves upstream of the heat exchangers b verifying differential pressure over the check valves as shown by pressure indicators upstream and downstream of the' heat exchanger The acceptance criteria require at least.a 12 psi drop, which is equivalent to a flow rate .

of 400 gpm through the heat exchanger associated with.the check' valve being

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teste This is in agreement with Administrative TS 3.11, which requires-

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that four heat exchangers be operable with a minimum o.f'400 gpm flow through each. The inspector reviewed the normal operating procedures associated with SW system manipulations of which could cause flow disturbance in the CAR heat exchangers. Each procedure contained an appropriate precaution statement to ensure that flow is not interrupted. .The inspector also reviewed the auxiliary-operator logs and noted that the remarks and limits portion of the log for the SW flow to the CAR heat'exchangers contained a range of 100 to 670 gp This was brought to the attention of plant management. The-inspector.was assured that the log will be corrected to correctly represent' the flow re- .

quirements of this system. This will be re-examined during routine inspectio No further inspector concerns were identifie . Followup on Events Occurring During the Inspection 6,1 Licensee Event Reports (LERs) y

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The'following LERs were reviewed for clarity, accuracy of the description of cause, and adequacy of corrective action. The inspector determined whether further information was required and whether there were' generic implications. The inspector also verified that the reporting require-ments of 10 CFR 50.73 and Station Administrative and Operating Procedures had been met, that-appropriate corrective action had been taken, and that the continued operation of the facility was conducted within Technical

' Specification Limit Fire Barrier Declared Inoperable'Due to Engineering Evaluation That Design Was Inadequate )

• 87-05 Water Contaminated Turbine Contro'l Oil Causes Control Valve Malfunction and Reactor-Trip

• 87-06 Spurious Down Spike on Nuclear Instrument Causes Turbine' Load Runback

• Inspection documented in NRC Inspection Report 50-213/87-0 ;

No unacceptable conditions were identified, a

6.2 Containment Isolation Valve Leakage After taking an RCS hot. leg sample on May 17, a chemistry technician identified that the hot leg sample containment isolation valve (SS-TV-965).

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was leaking. The valve was declared inoperable and the. leak was.. isolated-

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by closure of three manual isolation valves, in series, downstream of the leaking valve. Leak testing of the valve was conducted on May 18 per SUR 5.7-29, Liquid Sample Lines, and quantified as 3.0 cc/ min. The valve was cycled periodically and the leak check reperformed; no change l in leakage was identified. This amount of leakage was determined to be within that allowed for containment isolation valves. The inspector had no further question . 3 Inoperable Phones On June 1, for a period of about 30 minutes, the Middletown Office of the Southern New England Telephone Company lost both normal and emergency powe This caused the site phone lines which are routed through that office to be inoperable. This occurrence also rendered the dedicated hotline to the NRC Operations Center inoperable. The licensee contacted the NRC Operations Center via a SPRINT line and communications were established through the Millstone site on the licensee's microwave phone system. The licensee expects that more telephone service interruptions could occur this summer as a result of the predicted system brown-out A description of other available phone links to be used during such a situation was provided to the control' room staff. This includes the NU microwave phone system. The inspector reviewed the alternative communi-cations instructions and had no further question The hotline to the NRC Operations Center also was inoperable between May 6 and June 2 awaiting parts. During this period, the licensee communi-cated with the Operations Center via commercial telephone line .4 Reactor Coolant Pump (RCP) Oil Leaks During the week of June 1, the licensee observed an increasing lubricat-ing oil temperature on the No. 2 RCP upper bearing. A load reduction to 50% was conducted on June 6 to allow a_ containment entry to investi-gate this problem. A low oil level due to a leaking high pressure hose fitting between the N0. 2 RCP motor and the lift pump was identified; ten gallons of oil were added and a broken seal ring was replaced on the i hose coupling. The other RCPs were also inspected for' low oil' levels !

and oil leaks. Five gallons each were added to the No. 1 and 4 RCPs; i no additional leaks were identified. During this entry the. licensee identified that the ventilation dampers around the No. 2 RCP were mis-positioned. They were repositioned to provide better cooling-to the RCP motor and seal packag Power was increased to 100% on June The l control room staff continued to monitor the RCP bearing temperature The No. 2 RCP oil temperature again began increasing and, en June 8, it reached 200 degrees F (the maximum allowable temperature per the vendor manual is 205 degrees F). Another~ load reduction to 50% was conducte The licensee discovered that the seal ring replaced during the previous load reduction was defective. The seal ring was replaced again and 13 gallons of oil were added. In response to inspector questions, the lic-ensee stated that the oil which leaked out had been collected in the oil

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collection tanks and residue on the pump motor and in the surrounding area was cleaned up. At the close of the inspection period, the bearing oil temperature on the No. 2 RCP was steady at about 160 degrees F. The inspector had no further question . General Electric HGA Relay Seismic Qualification An inspection was conducted to determine if GE type HGA relays are used by the licensee in safety-related applications. It had been identified at another plant that these relays experienced contact chatter during seismic I testing. In an actual seismic event, the chatter'could render the associated equipment inoperabl The licensee has identified several safety-related uses of these relays.at Haddam Neck. A review of the seismic qualification information provided by GE for these relays shows them to be in compliance with original plant design criteri However, explicit qualification information regarding the original design criteria for this pisnt is not available. This issue was previously raised during the Systematic Evaluation Program (SEP). Unresolved Safety Issue (USI) A-46 was created to resolve generic questions regarding seismic qualification of electrical and mechanical equipment. This USI was described in Generic Letter 87-02 which requires licensee response by December 198 The inspector had no further question . Review of Periodic and Special Reports Upon receipt, periodic and special reports submitted pursuant to Technical Specification 6.9 were reviewed. This review checked on whether the reported information is valid and includes the NRC required data;.that test results and supporting information are consistent with design predictions and per-formance specifications; and that planned corrective actions are adequate !

for resolution of the problem. The inspector also assessed whether any re- '

ported information should be classified as an abnormal occurrence. The fol-lowing periodic report was reviewed:

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Monthly Operating Report 87-04, Plant Operations from April 1-30, 198 )

i No unacceptable conditions were identifie '

9. Switchgear Building Construction i In March of 1987 the licensee began construction of a new switchgear building which, when completed, will fulfill licensing commitments related to compli-ance with 10 CFR 50 Appendix R. Construction activities are being performed by Bechtel Eastern Power Corporation and have consisted of driving sheetpiling and excavation of the hole for the new building. During this inspection period, construction work was stopped on two separate instances relating to the effects of the integrity of the Emergency Diesel Generator (EDG) fuel storage tanks and the acceptability of. anchor bolt substitution. During the l

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last work stoppage, it was discovered that an uncalibrated vibration monitor !

had been used during sheetpiling driving. Currently, the licensee is not permitting any construction activities to continue until Bechtel satisfies plant management that there will be an improvement in the assurance of quality i of construction of the switchgear buildin !

9.1 Ground Subsidence Near EDG Fuel Storage Tanks On May 4, it was identified that the ground in an area south of the "B" EDG fuel storage tank and north of the Primary Auxiliary Building (PAB)

had subsided about six feet as a direct' result of the excavation for the building. Excavation work was-halted and fill was placed on both sides of the sheetpiling to prevent further subsidence. The licensee initiated '

a review to determine if this subsidence would affect the EDG fuel oil storage tanks and therefore the EDG operability. A special Plant Opera-tions Review Committee (PORC) meeting was held on May 8 to review the engineering evaluation results, determine EDG operability, and initiate any necessary corrective action The engineering evaluation concluded that the tank did not' move based on field inspection of the area that settled and the as-built drawing for the tanks and the associated piping and supports. From this evaluation, the PORC concluded that the subsi-dence had not affected the tanks' integrity and that the EDGs remained operable. The PORC discussed this situation with'the Bechtel personnel present and emphasized the need for plant management to be made aware of any anomalies in the construction process. Several methods to prevent further subsidence were discussed. The licensee chose to extend the sheetpiling to the PAB wall in order for the area where settlement had i occurred to be fully supported. Excavation work was permitted to con-tinu i The inspectors had further questions concerning the basis for the con-clusion reached by the engineering evaluation. Specifically, the in-spectors questioned if the licensee planned to reevaluate the seismic qualification of the storage tanks, including the associated piping and supports, now that the ground composition in the area has been identifie The licensee decided to include reevaluation of these seismic qualifica-tions in the ongoing efforts to resolve Systematic Evaluation Program j (SEP). Topic III-6, Seismic Design Consideration j 9.2 Unauthorized Anchor Bolt Substitution

On May 22, Stop Work Order CY-SWO-87-01 was issued as a result of the ,

identification that unauthorized substitution of anchor' bolts had oc- !

curre This unapproved substitution involved cinch anchors used t !

attach lagging supports to the Primary Auxiliary Building (PAB) wall.-

Thunderbolts were installed instead of the Hilti bolts specified by the applicable drawings and specifications. It appears.that contractor per-sonnel were aware of these substitutions during the work stoppage for

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the EDG operability evaluation and did not inform the licensee about i The licensee performed an engineering evaluation of the Thunderbolts and concluded that they are an adequate substitute for Hilti bolt Tne stop work order halted all Bechtel work on the new building. :The gate to the construction area was locke .

No contractor personnel are permitted to enter without accompaniment by licensee personnel. The order.i.s not to be lifted until Bechtel resolves technical discrepancies and corrects the root cause of this quality assurance proble . 3 Uncalibrated Vibration Monitor  :

As a result of the discrepancies associated with the bolt substitution, licensee QA audited of the records associated with the constructio This audit identified the use of an uncalibrated. vibration monitor during the driving of sheetpiling. These vibration monitors are used to measure potential effects on the surrounding buildir,gs. When questioned by the ,

licensea, contractor personnel statod that time constraints associated with the project affected their decision to continue work without a calibrated instrument. The licensee performed a calibration check on '

the vibration monitor and determined that it was within toleranc Construction of the new switchgear building will be inspected incident to ,

routine resident and region-based inspection . Emergency Preparedness Drill The licensee conducted their annual Emergency Preparedness drill on May 26, 198 The objectives of this exercise included demonstration of the activa '

tion of the emergency response facilities, notification of key officials.in the emergency organizations, utilization of communications systems, determina-tion of emergency action levels, and technical evaluation of plant conditions i to develop strategies for accident mitigation and recovery. The inspector observed Control Room activities to ensure that exercise objectives were being adequately tested and that previous drill performance weaknesses were cor-rected. An NRC inspection team was also onsite for drill' observation, the results of their inspection are detailed in NRC Inspection Report 50-213/87-1 . Unresolved Items

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Unresolved items are matters about which more information is required in order to determine whether they are acceptable items or violations. An unresolved items identified during this inspection is discussed in Paragraph . Exit Interview During this inspection, meetings were held with plant management to discuss the findings. No proprietary information related to this . inspection was identified.