IR 05000333/1997005

Insp Rept 50-333/97-05 on 970526-0628.No Violations Noted. Major Areas Inspected:Operations,Maint,Engineering & Plant Support

ML20149J967
Person / Time
Site:	FitzPatrick
Issue date:	07/22/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20149J932	List: IR 05000333/1997005 ML20149J931
References
50-333-97-05, 50-333-97-5, NUDOCS 9707290175
Download: ML20149J967 (22)

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/b, U.S. NUCLEAR ' REGULATORY COMMISSION t

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, '[ih 7]Ocensa No.: dPR-59: W"' ' ' ' Report; No.:' 97-05L r ' ... . i i s' g .+ , .l,.O, V D'o,cket No.: '50-333 .. s. ;. m , - ,, v.

A Licensee: New York Power Authority

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C Post Office Box 41

~ , , . Scriba, New York'13093 ~ ' - > , , g , 2 James A. FitzPatrick Nuclear Power Plant T': ',' s % acility Name: .- . F % Dates: May 26,1997 through June'28,1997 l ' L l.

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hp ' ' . W ;j T_ 2:;lnspectors: ' G.- Hunegs, Senior Resident inspector " ' , R.. Fernandes, Residsnt inspector

H.'Barkley, Project Enginaor - ,

' Approved by: John F. Rogge, Chief Projects Branch 2 l ' .id Division of Reactor Projects ' ' , I t a <

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k, i s ,., & lG. p .. .b ' , d'. [ aQ .y t , ", . EXECUTIVE' SUMMARY ' , y w i ~ James A. FitzPatrick Nuclear Power Plant

..l , ' /$ NRC Inspection Report 50-333/97-05 . c - g 1l ' ;> . [, ,i . E i ' . Qggg

., . . , ~ e' On May 25, the licensee. inserted a manual reactor scram from 70 ~ percent reactor power du,e, to the number'3 main turbine control valve.(TCV) failure. ~ Operators ! ' , : '4 demonstrated conservative cecision-making in manually scramming the plant when

' faced with a potential loss of reactor pressure control. Operators adequately . ' r:ontrolled reactor vessellevel and pressure during the transient and appropriately ' irnplemented emergency operator procedures. Equipment deficiencies were

. addressed by the licensee.

' a'- .* de .Following the reactor scram described above, and entry into emergency operating ! '" c g g

procedures (EOPs) for reactor pressure vessel (RPV) control, a determination'of rod

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~ ' position is required. Typically, not all control rods indicate " full in" on the fu!I core. . l . display and operators use backup methods such as "* *" on the emergency and ~ plant information computer rod scan to determine control rod position..The backup indications described do not Indier,te a specific control rod position, but rather they ' ? indicate that the control rod poshion.js unknown. The issue concerning whether '

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~o o bacimp methods are acceptable to verify control rod insertion following a reactor i scram is an inspector followup item pending completion of the licensee's evaluation.

Maingampag ""' 'e Overall, maintenance and surveillance activiths were well conducted, with good - adherence to both administrative and maintenance procedures.

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' e Poor foreign material exclusion controls at a vendor facility caused a safety relief j valve (SRV) to fait during "as found" testing. Technical Specifications concernirig SRV operability requirements were met. The licensee conducted a thorough review of the valve failure.

i Um- . . . P _ e The cause of the turbine control valve failure was due to the failure of the push rod '

spring housing coupling bolts due to previous improper maintenance. The licensee's

, , disposition of vendor information related to turbine control. valve maintenance did- , ' not prevent the TCV failure and resulting plant transient.

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A'poorly implemented procedure change process resulted in an incorrect inservice inspection (ISI) procedure and caused ISI testing requirements for control rod drive

bolts to be incomplete.~.in discussions with the licensee, the NRC staff concluded

' (that'there was no cone'ern with the current control rod drive bolting configuration.

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. c;/,- 'x," , Executive Summary (cont'd) However, pending review of the licensee's ISI program with respect to procedural ' controls and closer review of the supplemental acceptance criteria, this issue will , remain an unresolved item.

During a demand t'o open, a residual heat removal system valve failed to open due

to a torque switch roll pin failure. The licensee's response to the most recent failure was good and prior incorporation of torque switch replacement during valve maintenance was a good response to industry information. The decision to accelerate the preventive maintenance program for the remaining MOVs was a " conservative decision by the licensee.

' Elant Succort

An action plan which was developed to address an increase in the amount of radioactive noble gases in the turbine building had good management support and . attention. The action plan was comprehensive and identified other minor t discrepancies in the turbine building. The performance engineering statt aggressively pursued the issue, and the operations and radiological protection staff support was very good.

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, ' e f TABLE OF CONTENTS . E X EC UTIV E S UM M ARY............................................. ii . TABLE OF CONTENTS.............. iv ............................ < i L Summ ary of Pla'nt Status............................................. 1

1. O PE R AT I O N S................................................... 1 01.

Conduct of Operations.................................... 1

01.1 General Comments ~................................. 1 01;2 Manual Reactor Scram Due to Number 3 Turbine Control Valve Failure..........................................

01.3, Operational Safety Verification.......................

. .O2 Operational Status of Facilities and Equipment................... 3 ' 02.1 ; Engineered Safety Feature System Walkdowns

. ............. .03 Operations Procedures and Documentation .....................:4 , 03.1 ' Verification of Control Rod insertion Following a Reactor Scram

(Inspector Followup Item 50-3 33/97005-01 )............... 4 . ll. M AI NTEN AN C E................................................. 5 . .M1-Conduct of Maintenance

.................................. M1.1 General' Comments

' M1.2 General Comments on Surveillance Activities............... 5 ' .................................. . M1.3 Conclusions on Conduct of Maintenance.................. 6 M2 Maintenance and Material Condition of Facilities and Equipment

...... M2.1 (Closed) Licensee Event Report (LER) 50-333/97004

......... M7 Quality Assurance in Maintenance Activities

' .................... M7.1 Failure of Turbine Control Valve Rod Spring Housing Coupling ' Bolts

........................................... Ill. EN G I N E E R I N G.................................................. 9 E1 Conduct of Engineering

.................,................. E1.1 Control Rod Drive Closure Bolt Crack Indications (URI 50- . 3 3 3 /9 7 00 5 0 2 ).................................... 9 E1.2 Motor Operated Valve (MOV) Torque Switch Roll Pin Failure... 10 E8 Miscellaneous Engineering issues........................... 11 E8.1 (Closed) Unresolved item 50-333/93014-03

.............. E8.2 - (Closed) IFl 50 333/96006-03

........................ > IV.-- Plant Support 1? ! ................................................. R8-Miscellaneous Radiation Protection and Chemistry Controls......... 13 R8.1 Turbine ~ Building Noble Gas issue

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, e, q j ; - . ', s s >. , L A . #~4, Table of Contents (cont'd)- ' V. MANNGEMENT MEETINGS ........................................ 14-LX1 Exit Mee' ting Summary.....................................14'

'X2 Review of UFSAR Commitments............................ 14 . . Attachment i . ' Attachment 1 - Partial List of Persons Contacted ~ " ._ inspection Procedures Used items Opened / Closed, and Discussed- . List of Acronyms Used ~ .. J , .. N I . -1

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!'. i - Rgp_g_rt. Details Summary of Plant St.gtuji The unit began this inspection period shutdown as a result of a main turbine control valve failure. On May 24, the number 3 turbine control valve failed open. As the valve is normally full open, all pre transient plant parameters remained normal. The licensee began a plant shut down to make repairs. Originally, the licensee intended to shut the valve manually, however, the valve would not operate. At 70 percent reactor power, the licensee inserted a manual reactor scram and tripped the main turbine. Shutdown cooling was entered on May 25 and the licensee implemented a short forced outage. The plant was started up on May 30 and retumed to full power on June 1. The plant remained at 100% power through the end of the inspection period.

1. OPERATIONS

Conduct of Operations' l 01.1 General Comments The inspectors conducted frequent reviews of ongoing plant operations. In general, operations were conducted well. Specific events and noteworthy observations are detailed in the sections below.

01.2 Manual Reactor Scram Due to Number 3 Turbine Control Valve Failure jDsp_q_ tion Scoce a.

c On May 25, the licensee inserted a manual reactor scram from 70 percent reactor power due to a number 3 main turbine control valve (TCV) failure. The inspector responded to the site and observed the operating crew response during the transient, b.

Observations and Findinas On May 24 at 1:56 p.m. a half scram on reactor protection system (RPS) channel 'A' was received and operators reported that they heard an abnormal noise. The licensee's investigation identified that the number 3 main turbine control valve indicated closed on the electrohydraulic control (EHC) system panel. All other plant parameters were unchanged and normal for 100 percent power. Visualinspections of the TCVs showed that the spring housing coupling bolts for the number 3 TCV had failed which resulted in the valve being failed in the full open position. It was also noted that the position indication linkage for number 2 TCV was damaged.

Main turbine control valves are hydraulically operated open and closed and spring loaded to close. The main turbine control valves regulate steam to the main turbine, ' Topical headings such as 01, M8, etc., are used in accordance with the NRC standardized reactor inspection report outline. Individual reports are not expected to address all outline topic _ _ - . _.. _ _ -. .. - . _ _ _. _ - _ 5: >

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4-a thereby controlling reactor vessel pressure. The licensee decided to conduct a.-

reactor shutdown to make repairs"and power was reduced to 70 percent. Further _ . . povser reduction was not possible because the number 3 TCV could not respond to - l

maintain pressure at lower power levels. The licensee made an attempt to close,the' '

' number 3 TCV with the intent to control reactor pressure'during the plant shutdown , asing the remaining three TCVs.- However, the number 3 TCV could not be closed, , . . ,ind operators determined that a manual scram, immediately followed by a manual ' turbine trip, would be the best course of action to safely shutdown the reactor and' .' maintain pressure control. Operators were briefed on their various duties, stationed - .at the~ appropriate locations and on May 25 at 4:56 a.m. a manual reactor scram t , followed by a manual turbine trip was inserted. Following the scram, reactor vessel , Lwater level momentarily decreased to less than 177 inches and caused a primary _ . ; _ j- ' containment isolation signal group 2 isolation. Reactor vessel water level was '

.immediately recovered. Emergency Operating Procedure (EOP) 2, Reactor Pressure Vessel Control, was entered and the plant entered cold shutdown on May 25 at-l 6:45 p.m.

- .. > Plant equipment operated properly during the transient with the exception of three-control rods which had no. full-in indication at the full core display panel in the

. - control room and the "C" intermediate range monitor (IRM) failed upscale.

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Additionally, while lining up for shutdown cooling, the residual heat removal heat'

- exchanger outlet valve,would not open from the control room and was manually

opened. The cause of the valve failure was due to the failure of the torque switch _ ' roll pin.

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Conclusions , Operators demonstrated conservative decision-making in manually scramming the

plant when faced with a potential loss of reactor pressure control. Operators

i. adequately controlled reactor vessellevel and pressure during the transient and appropriately' implemented emergency operator procedures. Equipment deficiencies were addressed by the licensee.

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01.3 Operational Safety Verification

The inspectors observed plant operation and verified that the facility was operated l safely and in accordance with procedures and regulatory requirements. Regular tours were conducted of the plant with focus on safety related structures and ' systems, operations, radiological controls and security. - Additionally, the operabuity 'of engineered safety features, other safety related systems and on-site and off site c power sources was verified. - No safety concerns were identified as a result of these tours, c , i

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^ " The inspection activities during this report period included inspection during normal, " jbackshift and weekend hours. Regular. tours were conducted of the following plant , areas: , .m

' control room i secondary containment building u l radiological control point - ' electrical switchgear' rooms , ' mergency core cooling system pump rooms l e ' security access point protected area fence i ilntake structure i - diesel generator rooms Control room instruments and plant computer. indications were observed for ' s c ' correlation between channels and for conformance with technical specification (TS)- r - requirements. Operability of engineered safety features, other safety related-

' i systems and onsite and offsite power sources.was verified. _ The inspectors observed various alarm conditions and confirmed that operator response was in

accordance with plant operating procedures. Compliance with TS and .; < implementation of appropriate action statements for equipment out of' service was ) , I inspected. Plant radiation monitoring system indications and coolant stack traces

were reviewed for unexpected changes. Logs and records were reviewed to' ~

determine if entries were accurate and identified equipment status or deficiencies.

These records included operating logs, turnover sheets, system safety tags, and temporary modifications., Control room and shift manning were compared to regulatory requirements and portions of shift turnovers were observed. The ' inspectors found that control room access was properly controlled and that a > professional atmosphere was maintained. Daily supervisor meetings were attended j ' to assess personnel focus on risk significant stems and plant priorities.

O2 Operational Status of Facilities and Equipment ! l 02.1 Engineered Safety Feature System Walkdowns _ The inspectors performed a walk down of accessible portions of the following systems and performed general area tours: I ' oemergency diesel generator eintake structure eemergency service water '

• residual heat removal service water ehigh pressure coolant injection

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Equipment operability, material condition and housekeeping conditions were good.

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03 Operations Procedures and Documentation 03.1 Verification of Controi Rod insertion Following a Reactor Scram Unspector Followup item 50-333/97005-01) Lnipection Scop,{t a.

n Following the reactor scram on May 25,1997, three control rods did not indicate full-in on the full core display. These rods were verified to be full-in using the emergency and plant information computer (EPIC) rod scan. The inspector reviewed procedure requirements for verification that all control rods are full-in and discussed these requirements with the licensee.

b.

Findinos and Observations Abnormal Operating Procedure (AOP) 1, Reactor Scram, directs the control room operator to immediately verify all control rods era fully inserted following a reactor scram using one or more of the following methods:

• green " full in" light on the full core display (preferred method)

ofull core rod scan (EPIC red button)

• four rod display if any control rod is not full in, odditional actions are performed as required by procedures.

During previous scrams, the Green fullin light has not always illuminated for some control rods and therefore the operator refers to the EPIC full core rod scan as a backup indication for control rod position. The operator accepts the symbol "* *" on the EPIC control rod position display as indication that the associated control rod is fully inserted. The operator may also refer to the 4 rod display as a backup indication of control rod position and can accept a blank indication on this display as indication that ihe associated control rod is fully inserted. in fact, the two backup indications described do not indicate a specific control rod position, but rather they indicate the control rod position to be unknown.

All methods for determining whether or not all control rods are inserted following a scram rely on the rod position indicator system (RPIS) to detect each control rod's position by means of the position indicator probe (PIP) and a magnet in the control rod drive mechanism. The PIP senses the location of the magnet by a series of reed switches spaced every three inches over the active 144 inch length of normal control rod drive travel. A single switch is present at most locations with additional switches located at the extremes of full in and full out positions.

The inspector noted that the vendor, General Electric, has issued services information letter (SIL) no. 532, " Full In" Control Rod Position Indication, dated ' March 27,1991. The SiL describes limitations associated with using the 4 rod display and the process computer in determining control rod position and -

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recommends that the rod positiori data processing logic be modified. The licensee has not implemented the recommended modification and is reevaluating the installation of it. Additionally, the licensee intends to continue using the backup methods described to dearmine control rod position pending additional review.

c.

Conelysions Following a reactor scram and entry into EOPs for reactor pressure vessel (RPV) control, a determination of rod position is required. The issue concerning whether backup methods are acceptable to verify control rod insertion following a reactor scram is an inspector followup item (IFI) (50-333/97005-01) pending completion of the licensee's evaluation.

11. M AINTEN ANCE M1 Conduct of Maintanance M 1.1 General Comments a, loangction Scoom The inspectors observed all or portions of the following work activities:

• werk request (WR) 97-4496-Failed actuator bolts on turbine control valve (TCV)

number 3

• WR 97-4495 - Failed linkage on TCV number 2
• temporary operating procedure (TOP) 258 - Lift testing of spent fuelin peripheral storage locations
• WR 97-04499 - Repair *C" intermediate range monitor b.

Observations and Findinas The inspectors found the work performed under these activities to be professional and thorough. Technicians were experienced and knowledgeable of their assigned task.

M1.2 General Comments on Surveillance Activities a, trnpection Scone The inspectors observed selected surveillance tests to determine whether approved procedures were in use, details were adequate, test instrumentation was properly calibrated and used, technical 4,Jacifications were satisfied, testing was performed by knowledgeable personnel, and test results satisfied acceptance criteria or were properly dispositione, , , .. - - , , .. . . , , ., - , +

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The inspectors observed portions of the following surveillance activities: ,

• ST 348,

. Reactor building exhaust radiation monitors instrument / logic system ' . functional and simulated automatic actuation test -

• ST 8Q Emergency service water systern test elSP 27-Control room radiation monitor calibration

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• ST 20T Post scram control rod time test

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Qt2servations and Findinas , . The licensee conducted the above surveillance activities appropriately and in i accordance with procedural and administrative requirements. Good coordination

and communication were observed during performance of the surveillance activities.

M1.3 Conclusions on Conduct of Maintenance

Overall, maintenance and survoillance activities were well conducted, with good I adherence to both administrative and maintenance procedures.

_M2 Maintenance and Material Condition of Facilities and Equipment j ' M2.1 ~ (Closed) Licensee Event Report (LER) 50-333/97004: Failure of Safety Relief Valve to Open During Set Point Verification Testing Due to Foreign Materialintrusion a.

insoection Scooq The pilot assemblies for the main steam line safety relief valves (SRVs) were . removed and transported to a Quality Assurance (QA) approved vendor facility for .as found set point verification testing. The inspector reviewed the LER, equipment failure evaluation and vendor certification test reports and discussed the issue with the licensee.

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b.

Observations and Findinas During testing of pilot assemblies for main steam safety relief valves at a vendor facility, the "J" SRV would not lift. An equipment failure evaluation was conducted and, during the valve disassembly, a small screw was found wedged in the valve operator between the lower spring retainer and the inside of the bonnet wall. It was determined that the screw prevented the valve from opening during "as found" testing. The source of the loose screw was not determined, but was most likely introduced during previously conducted maintenance at the vendor facility in 1994.

The licensee witnessed the valvo disassembly and the QA organization has scheduled a vendor surveillance. The vendor implemented test procedure changes to minimize the potential for foreign materialintrusion into the valve bonnet area during testing avolutions, i

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Technical Specifications 4.6.E currently requires that at least 9 of eleven SRVs be operable and maintain a nominal setting of 1145 psig with an allowable setpoint , error of plus or minus 3 percent. To meet this requirement, the licensee checks all SRVs every 24 months. During the as found testing, only one other SRV lifted outside the tolerance band, therefore, TS requirements were met. Based on the-location of the screw and the valve operating characteristics, the licensee believes that the valve would have operated automatically'if required. The inspector noted that the valve had been manually opened during a transient 1 month prior to the beginning of the refuel cycle.

c.

Conclusiong Poor foreign material exclusion controls at a vendor facility caused an SRV to fail during "as found" testing. Technical Specifications concerning SRV operability requirements were met. The licensee conducted a thorough review of the valve i failure and appropriately scheduled a QA audit of vendor activities.

M7 Quality Assurance in Maintenance Activities M7.1 Failure of Turbine Control Valve Rod Spring Housing Coupling Bolts a.

Inspection Scope The number 3 turbine control valve (TCV) push rod spring coupling bolts failed with the plant at 100 percent power. The bolts secure the push rod to the spring housing. The valve operating mechanism utilizes a spring assembly which acts - directly on the valve stem to hold the valve closed. The valve is then mechanically opened by the hydraulic cylinder which lifts the end of the lever. The opposite end of the lever is held by a pin to the spring housing. Bolt failure allows the actuator springs to disengage and relax, preventing the valve from operating. The inspector reviewed the equipment failure evaluation, operating experience report, maintenance practices, and conducted a walkdown of the TCVs.

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Observations and Findinas A previous operating experience report documented a problem with bolt failure at another plant. General Electric Technical Information Letter (Til) dated September 25,1995 was issued to advise licensee's of the potential for failure of the turbine control valve and combined intermediate valve push rod-spring housing coupling bolts. The TIL states that to prevent this type of failure, proper maintenance and assembly procedures of the control valve is required, If the couplings on these valves have not been disassembled since installation, there is little chance that the coupling bolts will fail and it is recommended that inspection and replacament of bolts be performed during the next scheduled outage, if the couplings have been l disassembled since installation, inspection and replacement of bolts should be scheduled for the next opportunity. The push rod-spring guide couplings should not require disassembly during routine valve maintenance. If they do require disassembly, there are specific disassembly and reassembly processes which are

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required. Failure to adhere to these procedures can result in an initial torque with higher than normal stresses which can lead to yielding of the bolts and fatigue over an extended period of operation. In summary, the TIL recommended reviewing and revising maintenance procedures, ensuring proper thread lubricant, evenly tightening and properly torque coupling bolts and if the couplings have been disassembled since installation, then disassemble, inspect and replace the bolts.

, To disposition the Til, the licensee reviewed turbine control valve maintenance history and maintenance procedures considering the recommendations in the TIL.

Revisions to maintenance procedures to include bolt torque requirements were made and, additionally, the licensee noted that the procedures specified the correct thread lubricant.

The licensee scheduled number 4 TCV for inspection during the last refueling outage as the valve was normally subjected to the largest dynamic loading. Although all tha couplings for the TCVs had been previously disassembled, the licensee intended to inspect the other three control valve coupling bolts based on the condition of the number 4 TCV. This approach was not as rigorous as the recommendations in the TIL. Number 4 TCV coupling bolts did not show any degradation and therefore no additional inspections were scheduled. The combined intercept valve push rod spring guide couplings had not been disassembled since installation and therefore would not be subject to failure.

An equipment failure evaluation for the failed bolts was performed. Seven bolts were broken in the threaded area and five at the head shank interface. The bolts are required to be 5 inch,0.625 5ch diameter bolts. Four of the 12 coupling bolts were 5.5 inches which may have contributed to the failure. Previous maintenance practices required wrench tight and non-uniform preload and to reassemble the mechanism, the bolts were used to pre-load the spring. The longer bolts were used to enable mechanics to engage the thread when the spring was relaxed. The

unequal botting pattern and uneven preload resulted in additional stresses and bolt failure, During the short forced outage, the licensee replaced all the applicable bolts on the control valves and combined intermediate valves.

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Conclusion 2 The cause of the turbine control valve failure was due to the failure of the push rod spring housing coupling bolts due to previous improper maintenance. The licensee's disposition of the TIL did not prevent the TCV failure and resulting plant transient.

Subsequent corrective actions to replace applicable bolts on the control valves and combined intermediate valves were appropriate.

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Ill. ENQlf_lE_ERING E1 Conduct of Engineering " E1.1 Control Rod Drive Closure Bolt Crack Indications (URI 50-333/97005-02) a.

Insnection Scone While performing a closcout review of the inservice inspection program (ISI) following the refueling outage, the licensee identified that the bolts removed from the control rod drives (CRDs) during the refueling outage had not been properly inspected. The bolts were retrieved from the waste facility and the licensee performed the required VT-1 examinations. The results were that six of the 306 bolts inspected had crack like indications. The inspectors reviewed the issue with regards to the root cause for the missed inspections. A conference call was conducted between the licensee and the NRC staff to discuss the extent of the cracking of the bolts and the condition of the remaining bolts currently in service.

b.

Observations and Findinos i Each control rod drive is joined to a housing flange by eight 1"-8UNC x 5.5" cap screws (bolts). An issue was previously identified by General Electric in service information letter (SIL) 483 that informed the licensee of the potential for crack indications in the head to shank area of the cap screw. In response to the SIL, the licensee revised their maintenance procedures to perform an ISIinspection of the CRD bolting beginning in 1987. During preparation for refueling outage 12, the licensee had contracted an individual to review the procedure and subsequently changes were made to maintenance procedure MP-004.03, CRD Removal and Replacement. The change of concern altered the wording from " perform a VT-1 ISI examination of the removed flange bolts" to " perform a VT-1 ISI examination of removed or replacement flange bolts." This change eliminated the examination requirements of the used cap screws; thus they were discarded as radioactive waste without visual examination for cracking. The licensee's root cause analysis of the event was thorough and identified several root causes including lack of training for contract procedure writers.

As discussed above, six CRD cap screws exhibited crack like indications in the head to shank area. American Society of Mechanical Engineers (ASME) Section XI Code requires additional components to be inspected when examinations reveal indications that exceed allowablo indication standards of the code. As written, the licensee's examination procedure for the bolts would have required additional inspections. However, the licensee had completed the outage and additional inspections at the time of the discovery of the flawed bolts is a major effort. Upon closer review, the licensee determined that the ASME code had no acceptance criteria for CRD cap screws and, with the assistance of an engineering contractor, . developed a supplemental acceptance criteria. Applying the supplemental acceptance criteria, the licensee determined tbt the six bolts were acceptable and the additional examinations were not required. In addition the licensee destructively

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examined two'of the' cap screws for additional analysis'of the cracks. Of the 137 s CRDs,73 have the original cap screws that are of cone:ern based on the SIL. The ' , licensee intends to replace the remaining can screws during the next refueling ) outage.:

1 c.

Conclusions . The procedure change process wh!ch resulted in an incorrect ISI procedure was poor. Additionally, the licensee and the licensee's quality assuranco staff procedure . review was not thorough and did not identify the error. The licensee's efforts to i retrieve the bolts from the waste facility and perform the inspections were good in ) ' discussion with the licensee, the' NRC staff concluded that there was no concern. I with the current CRD bolting. configuration. However, pending review of the ] licensee's ISl program with respect to procedural controis and closer review of the _ . supplemental acceptance criteria this issue will remain an unresolved item, (URI 50- '333/97005-02).

, i E1.2 Motor. Operated Valve (MOV) Torque Switch Roll Pin Failure )

a.

Insoection Scope I-On May 25,.while attempting to initiate the shutdown cooling mode of the residual heat removal (RHR) system,10MOV-128, "B" RHR heat exchanger cutlet valvrs., i ' - failed to open. This delayed entry into shutdown. cooling.and the reactor coolant :.

system temperature increased such that the plant went from cold shutdown to hot shutdown with the temperature cresting at 231*F, at which time the valve was ~ - i manually opened The licensee subsequently determined that the motor actuator on .the valve had failed during the closing cycle because of an MOV torque switch roll p pin failure. This failure mode had not been previously observed at the plant but is a L known industry problem and the subject of NRC Information Notice 94-49, Failure l of Torque Switch Roll Pins, issued July 6,1904. The inspectors reviewed the ) ' licensee's corrective actions to resolve the recent issue as well as the licensce's actions essociated with the inductry information distributed in 1994.

! b.

Observations and Findinas i Under accident conditions,10MOV-128 is open and remains open to allow RHR

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' flow through the 3HR heat exchanger during the suppression pool cooling and , containment spray modes. - The valve remains in its safety position and performs no ' active safety function. Industry information had determined that the roll pins, in some applications, could fail as the result of shearing forces apphed to the pin by tne torque switch during the unseating' phase of a stuck, locked or thermally bound , valve. The resulting failure is that a valve under torque switch logic control will not open the'conuacts which de-energize C e actuator motor. In the case of 10MOV- , 128, the motor'was stalled and damaged; The licensee replaced the torque switch ' F with~the upgraded version (solid pin), as well as replaced the torque switch ior , 10MOV 12A, "A". RHR heat exchanger outlet valve.

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The licensee's initial response to the issue included ordering new upgraded torque switches with solid pins, establishing a replacement evaluation criteria, and applying that criteria to determine which actuators needed to be roplaced. The licensee dearmined the scope of replacement to include those valves thrst were flexible or solid wedge gate valves with SMB/SD 0 to 4 actuators: had non-locking stem to stem nut threads; and hac' a performance characteristic to include a high and sudden release of unwadging force relative to seating force. The licensee determined that 22 valves meet the first two criteria, but did not meet the latter two criteria and thus, did not take any corrective action at that time to replace any torque switches. However, as a method to enhance reliability, the licensee elected to put administrative controls in place to replace the torque switches in 17 of the 22 MOVs at the individual valves during the next six year maintenance period. Five of the valves were determined not to be safety related and were not put on the list for switch replacement, it was one of these valves,10MOV-12B, that failed during this event. The licensee concluded that the failure was most likely the result of fatigue. Fatigue failure from cyclic loading was not addressed by the licensee during the 1994 industry information review. Of the 17 valves, five have been replaced since the implementation of the o.iginal corrective actions. During the most recent review, the licensee estimated that 10MOV-1.2B had a greater number of cyc!es on it than the 12 remaining valves. In addition to replacing the switches on 10MOV-12A and B, the licensee plans on replacing the remaining torque switches before the end of the next refueling outage.

c.

Cpaciusions The licensea's response to the most recent failure was good and the prior incorporation of the torque switch replacement during valve maintenance was a good response to industry information. Thn decision to accelerate the preventive , maintenance on the remaining MOVs was a conservative decision by the licensee.

E8 Miscellaneous Engineering issues E8.1 (Closed) Unresolved item 50-333/93014-03: Control room ventilation single failure vulnerability. Thia item was opened in NRC inspection report 93-14 and epdated in NRC inspection reports 93-20, 93-24, 93-82, 94-06 and 94-09. The original prob!em was identified by tne licensee during an operational experience review and involved identification of an unquantified arnount of unfiltered airflow into the control room through a normally open manually operated bypass damper with the system in the iscletion mode concurrent with a single failure, specifically the failure of an upstrearn motor operated valve (MOV) isolation valve to close. Later, the licm.see identified other ventilation system discrepancies or degraded conditions involving possible electrical separatico concerns, aegraded equipment conditions (i.e damper / valve resiliert seald allowing outside air infiltration and inadequately balanced air flows. As a result, the system was operated in the isolate mode frem July 1933 until May 1994 while it was thoroughly evaluated and system design concerns and materini deficiencies were corrected at the first available outage. The licensee's fMdings and correcthia actions in this matter were documented in LER .93019-02 and in their control room ventilation systism action olan.

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The licenfee's review of this rnatter determined the root causes to be inadequate design control and design and licensing basis documamation combined with inadequate preventive maintenance on the MOVs and dampers and inadequate surveillance testing for system air in-leakage. The licensee developed and i , implemented a comprehensive corrective action plan that was completed during a maintenance nutage in April 1994 and resolved the open design and surveillance concerns with the system. In addition, the licensee reperformed and resubmitted to the NRC in' March 1995 a revised control rnom habitability stu'dy per NUREG-0737, which reflected the numerous minor changes in the design and operation of the. j system as well as documented the safety significance of the findings maue by the licensee during their investigation of the system. The licensee's evaluation determined that ' tith the design disciepancies and systeen degradations noted, the , control room would still have remained habitable following a design basis accident, i The inspectors conducted a_walkdown of the control room ventilation system with - the system engineer, which included a review of the system flowpath, the valves and dampers previously in question, and inspection of the roof mounted primary and secondary air intakes. The inspector noted evidence of system flow bt. lancing, modifications to reduce air in leakage and system equipment imp;ovements (e.g.

] damper and fan motor replacements and preventive maintenance). Selected systern modifications were also reviewed as was the most recent system charcoal filter test results. The Nuclear Safety Evaluations which changed the final safety unalysis report (FSAR) description of the operation of the system as well as returned the i system to the normal (versus isolate) mode of operation were reviewed as were elements of the recent power uprate amendment which impacted on the control room habitability analysis. The power uprate amendment greatly reduced the maximum permissible reactor coolant system activity level allowed, thereby significantly reducing the post-accident estimated radiation dose in the control , room. Based on this review and discussions with the system engincer, the multipia concerns tracked by URI 50-333/93014-03 have been resolved.

. , The failure to provided adequate design controls and design documentation for this safety-related system, as evidenced by the design discrepancies and multiple system degradations identifad in 1993 1994, along with inadequate surveillance testing for in-leakage to the v7em and the control room envelopc, constitute violations of 10 CFR 50, Appendix 8, Criterion til and IX. However, these violatiens were identified by the licensee and corrected by immediate and long-term comprehensive corrective actions. Similar previous violations had not been - identified in this area. Furthermore, while the existing system deficiencieu arrd - degradations required correction, subsequent reanalysis indicated that the control room ventilation system would still have provided a habitable enviror, ment after a design basis accident. Therefore, these violations will not be cited as t' e criteria of h Section Vll.B.1 of the NRC Enforcement Policy warc satisfied. (Non cited violation - (NCV) 504333/97005-03) The NRC's review of the licensee's revised NUREG 0737 control room habitability study is currently pending.

E85 (Closed) IFl 50-333/96006 03: Seismic qualification process for commercial grade equipment. NRC inspection report 50 333/30006 documented a case in which the , s !

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seismic qualification process for n cornnercial grade component, in this case battery . , cells, did not ensere that the ccmponent was an exact ons for one replacement of a previously qualifted component. In this particular case, historic product design changea made by the vendor did not degrade the seismic qualification of the _compenent. This issue was left as ar inspector followup item pending review of ' . ' the licenssa's root cause evaluation of the inatter and extent of condition ' ' evaluation.

The licensee determincd tha root causos of the poblem included inadequate communication with the vendor regarding component dasign changes that cauld potentially impact on seismic qualification, a lack of avaifable dosign infccmation in the original battery cell seistnic qualifi,ation report and a poor questioning cttituc'e r by those involved in the design change approving tha use of thesa batteries (which - were slightly taller than the original and had different inter-cell connectors, but the , same model number). The licensee's corrective actions in this matter were prompt and comprehensive and !ncluoed changes to procurement' documents to require more explicit inf;rmution from suppliers regaiding changes madc to their products beirg procured through the commercisi dedication process, procuremant and j ~ ergineering personnel training and a selected sampling of previcus commercial , ' grade dedication packages for similar situations. That review, which encompassed i 36 of the 203 commorcial grade ded!cotion packages since 1992, formd that each package satisfactorily addresaed seismic qualificat on. Based on the documentation i . revieweo, which implementcd the above corrective actions, and an extent of j condition review which found no other prob!ams, this inspnctor followup item is considered closed, ' jV. Plant,Juppt R8 Miscellaneous Radiation Protection and Chemistry Controls - R8.1 Turbine Building fJoble Gas issue a.

insoeqtion Scoqft

l Fo!!owing the startup from the last refueling outage and during the current opereting i-cycle, the licensee noted an incieasing trend in tha amount of noble gases in the ' turbine builoing. The decay products of the noble gases had the adverse effect of

causing whola body caunt monitor (WBCM) alarms and delays when personnel were attempting to exit the radiologically controlled ersa (RCA). The inspector discussed the issue with radiation protection personnel and reviewsd the licenseo's corrective . actions.- ' b.

Obsgativas andJJndirm

' The licensee developed an action pla, requiring support from ocarations, radiation prNection, and engineering personnel.. Corrective actions included: steam area " ' iwalkdownt; ar.d minor steam leak repairs; turbine building ventilation sy7te.n ' ~, . waikdewr's and adjustments: and atmospherin monitoring during off cas system

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, adjustments. During the most recent forced outage, the licensee determined, through the performance of a leak test, that the offgas dryer purge flow element j.

had a leaking flange. Following the repair of the flange and subsequent startup of the plant, the WBCM alarms due to noble gases were essentially eliminated.

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. ' The issue 'had good management support and attention. The action plan was

. comprehensive and identified other minor discrepancies in the turbine building. The . performance engineering staff aggressively pursued the issue, and the operations and radiological protection staff support was very good.

V. MANAGEMENT MEETINGS ' eX1 Exit Meeting Summary j , @ tThe insper: tors presented the inspections results to members of the licensee management at the conclusion of the inspection on July 15,1997. The licensee acknowledged the findings presented.

The inspectors asked the licensee whether any materials examined during the inspection

should be considered propsietary. No proprietary information was identified.

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h X2 Review of UFSAR Commitments-A recent discovery of a licensae operating their facility in a manner contrary to the Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused

review that compares plant practices, procedures and/or parameters to the UFSAR dauedption. While performing the inspections discussed in this report, the inspector reviewed the applicable portions of the UFSAR that re!ated to the areas inspected. The inspector verified that the UFSAR wording was consistent with the observed plant W ' practices, procedure and/or parameters.

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Attachment 1 , PARTIAL LIST OF PERSONS CONTACTED - Licensee , ~ M. Colomb, Site Executive Officer

P.' Brozenich, Operations Manager D. Lindsey, General Manager, Operations. J. Maurer, General Manager, Support A. McKeen, Manager, Radiological and Environmental Services - ., H. Salmon, Vice President, Nuclear Operations.

' D. Ruddy, Director, Design Engineering ' D. Topley, General Manager, Maintenance - . NRC W. Hehl, Director, Division of Reactor Projects L. Nicholson, Deputy Director, Division of Reactor Projects (Acting): i i .i - i , . l i ~J U ,., ,, , _ _ _ _ .. _. 'C,_,' ll: ' .y , ., ) W ,r . t @. Attachment 1~

, !!NSPECTION PflOCEDURES USED ,. ,-37'551 Onsite Engineering ' '

~ 62707-Maintenance Observations , ' ' iB1726 Surveillance Observations . . .

71707 Plant Operations -

. 71760-Piant Support . 92903 ' Engineering Followup

' ITEMS OPENED, CLOSED, AND DISCUSSED . 3' (Qpened 50-333/97005 O't. IFl Use of alternate methods to verify full in control rod position indication

, 1i02333/97005-02 URI ISI program with respect to procedural controls end suppleinental ecceptance criteria , 50 333/97005-03 NCV ' Failure to provide adequate' design controls and design

documentation for the control room ventilation system.

E!911d 50-333/93014-03 URI Control rocm ventilation sin 0 o. failure vulnerability i 50 033/96006 03 IFl Seismic qualification process for commercial grade equipment 50 333/97005 03 NCV Failure to provido adequate design controls and design k documentation for the control room ventilation system.

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I LIST OF ACRONYMS USED a: 7. l i ! AOP'.. . Administrative Orsrating Procedure

ASME-i American Society of Mechanical Engineers ' CRD.

Control Rod Drive I - EHC.

Electrobydraulic Protection Control System ) ', - =EOP; Emergency Operating Procedure

EPIC.

Emergency and Plant information Computer. '! l

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'. Inspection Followup Item n IRM i sintermediate Range Monitor j ISI inservice inspection H ' i-

LER.

Licensee Event Report MOV . Motor Operated Valve f . NCV Non-Cited Violation ' ' ?NRC.

Nuclear Regulatory Commission j

- PIP - Position' indicator Probe-I l T OA Guality -Assurance 'RCA Radiological Controlled Area - ' RHR' Residual Heat Removal q , , RPIS Rod Position Indication System.

RPS-- - Reactor Protection System - j _ RPV Reactor Pressure Vessel .l '

Sit Services information Letter i

SRV Safety Relief Valve - ) . TCV 2 Turbino Control Failure j TIL

Tuihine Information Letter

! ' - TOP- ' Temporary Operating Procedure H TS . Technical Specification

.-l FSAR Final Safety Analysis Fieport

- - URi Unreasolved Itarn. <; !- UFSAR Updated Finnt Safety Analysis Report I WBCM Whole' Body Count Monhor-

WR Work Request

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