IR 05000029/1986009

Insp Rept 50-029/86-09 on 860626-0702.Violations Noted:Rcs Vents & Emergency Feedwater Sys Inoperable from 851202-03 Until 860626 & Maint Support Dept Resources & Training Insufficient

ML20207D584
Person / Time
Site:	Yankee Rowe
Issue date:	07/03/1986
From:	Elsasser T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20207D554	List: IR 05000029/1986009 ML20207D550
References
50-029-86-09, 50-29-86-9, NUDOCS 8607220142
Download: ML20207D584 (13)

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U.S. NUCLEAR REGULATORY COMMISSION Region I Report N /86-09 Docket N Licensee N DPR-3 Licensee: Yankee Atomic Electric Company 1671 Worcester Road Framingham, Massachusetts 01701 Facility Name: Yankee Nuclear Power Station Inspection at: Rowe, Massachusetts Inspection Conducted: June 26 - July 2, 1986 Inspector: H. Eichenholz, Senior Resident Inspector indale, esident Inspe or-Haddam Neck Approved By: h//d/fS6 T. Els~asser, y f g eactor Pro cts Section 3C Dhte i Inspection Summary: Inspection on June 26, 1986 - July 2, 1986 (Report N /86-09) Areas Inspected: Special, unannounced inspection on day time and backshifts of the events associated with the discovery during surveillance testing on June 26, 1986 that the Reactor Coolant System Vents (RCSV) and the Emergency Feedwater (EFW) System were inoperable. The inspection involved 49 total hours by a Senior Resident Inspector and a Resident Inspector.

Results: Two apparent LCO violations were identified for the RCSV and EFWS being inoperable from December 2 and 3, 1985, respectively, until June 26, 1986. Con-cerns were also identified with an apparent insufficiency in resources and training i in the plant Maintenance Support Department.

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lABLE OF CONTENTS Page 1. Summary.............................................................. 1 2. Purpose.............................................................. 1 3. Event,-Initial Evaluation, and Notification.......................... 2 4. System Descriptions and Failure Analyses............................. 5 5. License 9 Corrective Action Program................................... 7 6. Persons Contacted and Managerrent Meetings............................ 9

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1. Summary During the 1985 Refueling Outage, the licensee implemented design modifica-tions that involved relocating power supplies for valves in the Reactor Cool-ant System Vents (RCSV) and the Emergency Feedwater System (EFWS) to address current regulatory issues and plant betterment efforts. Surveillance testing on June 22, 1986 and maintenance troubleshooting on June 26, 1986 resulted in determining that a discrepancy existed between the design and the installed equipment that provides the power supplies for valves EBF-MOV-557, PR-MOV-558, VD-MOV-559, and PR-M0V-560. The discrepancy involved undersized trip coils in the combination breaker / starters for the aforementioned valve The licensee immediately initiated an aggressive investigation program that included a task force to search out the underlying reasons for the observed discrepancy. Short and long term elements of the investigatory program were identified. The plant outage to repair equipment, which was in progress at the time of discovery of the discrepancy, was extended for approximately three additional days until the short term task force's investigation could provide the assurance that similar discrepancies did not exist in other installed safety related plant equipment or spare parts. The licensee's strong commit-ment to plant safety was evident in both their deliberations and implemented corrective actions. The task force conducted it's investigation responsibili-ties in an open and candid atmosphere in the presence of NRC personnel. A proper regard for safety was displayed at all times by the licensee's person-nel and their concern that the event occurred with it's potential ramification (that it represented a blemish of their long term safety record) was evioen The licensee has determined, on a preliminary basis, that the root cause for the occurrence involves a brcakdown in both the receipt inspection and post modification testing programs. The last NRC SALP Report 50-29/85-99 raised concerns for the need for licensee management attention to insure sufficient resources and training are provided for personnel who provide an important part in assuring quality in the design change and modification are Appar-ently, additional management attention in this area is still neede The licensee was continuing its investigation at the completion of the in-spection. A decision was made to restart the plant on July 1, 198 . Purpose The purpose of the special inspection was to review the details of the licen-see's identification that undersized instantaneous current trip coils were installed in the 480 VAC power supply circuit breakers for valves EBF-MOV-557, PR-MOV-558, VD-MOV-559, and PR-MOV-560, and to identify the potential impact that the discrepancy had on plant equipmen . __ _ .- . . _ _ _ . --

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3. Event, Initial Evaluation, a1d Notification A plant outage was initiated on June 18, 1986. This outage resulted from a small secondary side leak on the 2" blowdown line of the No. 3 Steam Generato On June 22, 1986, with the plant in a cold shutdown condition (Mode 5), plant personnel were performing procedure OP-4517, Rev. 7, Surveillance of Motor Operated Valves. This surveillance was being performed on the Reactor Coolant System Vent (RCSV) valves PR-MOV-558, VD-MOV-559, PR-MOV-560, and VD-MOV-56 This surveillance activity resulted from the ISI coordinator's request that the Maintenance Department perform procedure OP-4517 to collect travel times on these and other valves required to be tested when the plant is in a cold shutdown condition as specified in the licensee's ISI Progra The surveillance test consisted of recording operating amperage readings and obtaining opening and closing travel times for the valves. The first valve tested was PR-MOV-558, the inboard pressurizer head vent valve. The valve performed satisfactorily, and maintenance and operations personnel initiated the surveillance testing on valve VD-MOV-559, the inboard reactor head vent valve. According to the plant electrician performing the surveillance test, this valve went open and the Main Control Board (MCB) open position indication light extinguished. The electrician thought that the control fuse had blown, so he replaced the fuse. Once the breaker in the motor starter cubicle was placed back on, the valve was stroked closed with no abnormalities note Since it was thought that replacing the fuse had corrected the problem, no further attention was given to the matter. The next valve tested was PR-MOV-560, the outboard pressurizer head vent valve, which immediately lost the MCB closed position light when the operator placed the valve's control switch to the "0 PEN" position. Thinking it was a control fuse problem, fuses were changed by the electrician. After replacing the fuses a second time, due to a subsequent attempt to stroke the valve, the problem was not corrected and Maintenance Request (MR) 86-952 was issued on June 22, 1986 which specified that the control fuses were blowing. When tested, valve VD-MOV-561 performed satisfactorily as had PR-MOV-558, the first valve teste A plant electrician was assigned to investigate the reported trouble with the valve on June 26, 1986. As part of troubleshooting, the breaker's portion of the 480VAC combination starter motor circuit protector (ITE Gould Catalog N A821) for valve PR-MOV-560 was multi-amp tested. This resulted in determining, that the circuit breaker would trip open instantaneously at a current of ap-proximately 3 amperes. Similar multi-amp testing of the circuit breakers for i valves V0-M0V-559 and PR-MOV-558 resulted in determining that they would also trip open at a value of approximately 3 amperes.

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The performance of the tested trip coils were compared against their nameplate

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data and the design requirements specified in Engineering Design Change Re-i quest (EDCR) 84-312, Relocation of Power Supplies for Motor Operated Valves.

! The Maintenance Manager directed that the cognizant Maintenance Support De-partment (MSD) engineer for the EDCR installation contact the cognizant Yankee Nuclear Services Division (YNSD) engineer to resolve the problem. Based upon

the YNSD engineer's review, it was determined that the installed trip coil

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(an ITE Gould Catalog No. A80A10) was the incorrect device. Since the 480 VAC power supply for EBF-MOV-557, the Emergency Feedwater Alternate Feed Iso-lation Valve, was also installed as part of EDCR 84-312, it was checked and found to have the same incorrect trip coil installed in the breake Its trip value from multi-amp testing was determined to be approximately 4.6 ampere The A80A10 trip coil is sized for a maximum continuous current of 0.57 amperes, with an adjustable instantaneous current trip range of 2.41 to 4.34 ampere The correct trip coil would have provided a maximum continuous current rating of 2.75 amperes and provided an adjustable trip setting in the range of 15-28 amperes. For the three RCSV valves with the incorrect as-found trip coil in-stalled in the breakers, the full load current (FLC) and locked rotor current
[ heavy load] (LRC) nameplate ratings are 0.75 and 4.8 amperes, respectivel The E8F-MOV-557 valve motor's FLC and LRC nameplate ratings are 0.75 and amperes, respectivel By 3:30 pm on June 26, 1986, the plant mar,agement was notified that the dis-crepancy existed between installed and t'esign specified trip coils. The resident inspector was notified by the licensee of the aforementioned discre-pancy at 3:45 pm the same day. An EN', call was made at 4:51 pm on June 26, 1986 as a 4-hour non emergency reporc per 50.72(b)(2)(i). No unacceptable conditions were noted by the inspector with regard to the licensee's notifi-cation practice As a result of reviewing design and implementation documentation, and con-ducting interviews with licensee persorinel, the inspector noted the following:
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The Bill of Materials (80M) for the EDCR specified that the four 480 VAC combination starters for the valves in question are type 50AF, 2.75 AT, Size 1, FVR, Ambient Compensated. T8 Overload Relay, etc. According to the Maintenance Manager, the "2.75 AT" portion of the 80M means " amp-trip" and relates to the maximum continuous current rating of the trip coil that is installed in the circuit breaker portion of the combination breaker / starte A catalog number was not used in specifying the trip coi The purchasing process for the installed equipment reflected the above enumerated nomenclature (P.O. No. 104980).

-- The EDCR specified that installation and testing procedures for the equipment required by the EDCR would be developed by plant personne The inspector found that the EOCR contained neither instructions on where to set the trip coil adjustment, nor recommended acceptance cri-teria for testing the trip coils once se A memorandum from the YNSD cognizant engineer, YRP 571/86, issued on June 26, 1986, indicates that it was intended to set the instantaneous trip on the high adjustment, which would have resulted in tripping the valve motor on an instantaneous current of 28 amperes. The YNSD cognizant engineer informed the inspector that any setting in the trip range would have been acceptabl __

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' The installation of the EDCR was considered complete by the licensee on November 27, 1985 foll wing the performance of post installation testing and surveillance testing. However, the post installation testing did not directly test the trip coil In procedure OP 5000.181, Rev 0, MOV Power Supply Changes, the four valves in question had their heater (thermal overload) trip devices tested. No testing of breaker instan-taneous trip coils was performed. According to the cognizant MSD engi-neer, some problems were experienced with the circuit breakers tripping open during the testing of the thermal overloads, and to remedy this situation the trip coil settings on some of the circuit breaker's trip values were adjusted upwar He further indicated that this activity was not documente The electrician performing multi-amp testing on the E8F-MOV-557 valve on June 27, 1986, noted that its circuit breaker in-stantaneous trip setting was set to the high trip value.

, Additional post installation testing had consisted of verifying the travel time of the valves, per procedure OP 4517. Subsequent surveil-lance testing on November 29, 1985 of the RCSV valves performed per pro-cedure OP 4262 Rev 1, Operability Test of the MCS Valves, did not detect any discrepancies. No further testing the these valves was performed until June 22, 1986. The EBF-MOV-557 valve is travel time tested monthly, and has not resulted on detection of the deficiency subsequently

identified. The other valves are only tested at refueling outage The inspector noted that in the process involving receipt inspection, the MSD cognizant engineer for the EDCR will perform the receipt in-spection of the purchased material. A review by the inspector of the

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quality assurance documentation of Purchase Order 104980, which was utilized to procure the four ITE Gould combination motor starters used, , included a circuit breaker test report for the four trip coils installed in the shipped material. This documentation, as well as the inspection of the equipment directly utilizing nameplate data, did not result in the detection that the received equipment was unable to meet the design , objectives of the EDCR. It appears that the assigned MSG cognizant in-i dividual was not aware of the technical aspects involved in designing i and purchasing the trip coils, therefore, he was unable to identify the incorrectly supplied material.

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The inspector determined that the EDCR calculation (YRP 1252/85) used l in selecting the thermal overload for the EBF-MOV-557 utilized a LR The nameplate data reflects a 5.5 LRC value. This results in an expected overload condition of 1,020 percent versus the calculated 891 percent. This reduced the minimum trip time of the thermal overload from 3.2 to 2.7 seconds. The maximum trip time is 8.2 seconds, which was previously 10 seconds. The selection criteria used in the calculations j states that..."When carrying locked rotor current, the thermal overload relay should actuate in a time within the motor's limiting time for carrying locked rotor current." The time specified for LRA in the EBF-l i

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MOV-557'is assumed to be 10 seconds. This item was discussed with the cognizant engineer. Plant personnel were unaware of this discrepancy until informed by the inspecto A memorandum issued by the cognizant YNSD engineer on June 27, 1986, YRP 577/ 86, indicated that the selection of the trip coils for motor operated valves should meet the criteria that 1) sustain the maximum continuous current of the valve, 2) protect the power cable against short circuit, and 3) should not trip instantaneously at locked rotor current. It appears that for all four valves in question, this last criterion could not be reasonably expected to be me This means that for currents in excess of normal running values (~1 amp), including overloads, the breakers should not trip instantaneousl Such trips are reserved for ground or line to line faults onl As found, the 557, 558, 559, and 560 breakers' instantaneous trips actuated at approxi-mately 3-5 amps, well below currents representative of a fault. They were the wrong size for the associated valve moto . System Descriptions and Failure Analyses 4.1 Reactor Coolant System Vents (RCSV) RCSV's are provided to exhaust noncondensible gases and/or steam from the reactor and/or pressurizer heads to the containment atmosphere during a post accident situation to assure that core cooling during natural circulation would not be inhibited (see sketch A). The reactor head vents consists of two series motor-operated valves (VD-MOV-559 and VD-MOV-561).

The pressurizer head vent system also utilizes two series MOVs (PR-MOV-558 and PR-MOV-560). The MOVs are remotely operated and monitored from the control room. Cross-connect piping is provided between the two series MOVs, so that the upstream MOV on one vent line can discharge through the downstream MOV of the other vent line in the event of a single failure on one of the downstream vent valves. Both vent lines discharge to the pressurizer PORV (PR-50V-90) relief line and is piped to the Low Pressure Surge Tank (LPST), unless the rupture disc has ac-tuate CFR 50.44(c)(3)(iii) requires the licensee to have the RCSV installed and operational in order to provide " improved operational capability to maintain adequate cooling." The means for meeting the design require- . ments of the subject rule is discussed in NUREG-0737, Item II.8.1. Based ' upon an NRC:NRR review contained in the Safety Evaluation Report dated , September 14, 1983 the licensee's system as described above was deter-mined to be in compliance with the rule. Subsequently the licensee re-ceived approval to implement Technical Specifications that specified the operability and testing of the system on October 1, 1985.

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Technical Specification (TS) 3.4.11 requires,that, while in modes 1-4, at least one RCSV path consisting of at least two valves in series be operable and closed at each of the following locations: reactor vessel head (VD-M0V-559, VD-MOV-561) pressurizer steam space (PR-M0V-558, PR-MOV-560) Three of the four valves required to be operable by TS 3.4.11 (MOVs-558, 559, 560) were determined to have undersized trip coils in their re-spective trip breakers, which could cause the valve's motor operator to trip prematurely, especially under design basis conditions (differential pressures across the valves). From December 2, 1985 through June 20, 1986, both RCSV paths were incapable of performing as designed, rendering the entire RCSV inoperable. This constitutes a violation of TS 3.4.11 (VIO 86-09-01).

If the RCSV were required to operate under design basis conditions, the valves probably would have tripped earlier than desired. However, it should be noted that the emergency motor control centers (EMCC) in which the breakers are installed, are located in a post-accident shielded en-vironment (switchgear room). If and when the RCSV valves were called upon to actuate and did not, plant personnel can access the switchgear room and bypass the trip coils. A proposed TS change was submitted 10/15/85 to the NRC by the licensee to remove power from the valves, and require future use of sending an operator to the Switchgear room when the system needs to be use .2 The Emergency Feedwater (EFW) System The EFW System provides Steam Generator (SG) makeup following a loss of normal Feedwater (FW) incident and is shown in Sketch B. One steam driven and two motor driven EFW pumps are available to supply demineral-ized water to the SGs. EFW discharge is through either the normal feed-water lines (normal EFW flow path) or the SG FW inlet nozzles via a cross connection from the SG Blowdown (SGBD) lines (alternate flow path). The alternate flow path is included in the system design to insure the cap-ability to supply EFW to the SGs independent of the loss of the normal EFW flow path. The EFW pumps are capable of supplying either flow pat The EFW flow path is determined by positioning three motor-operated valves; EBF-MOV-555 (normally open), EBF-MOV-556 (normally open), EBF-MOV-557 (normally closed). EBF-MOV-557 must be opened to provide EFW flow through the alternate EFW flow pat The design that installed the EFW System's alternate flow path, which is implemented by opening the EBF-MOV-557 valve, increased the capability and the redundancy of the system. However, the alternate feed capability is necessary to provide a mechanism for supplying feed flow to the steam generators which is independent of the feedwater piping in the turbine buildin __ - _ _ _ ___ - . _ _ - . _ - _ . __ _ - _

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As indicated earlier in the report, installation of an undersized trip coil in the circuit breaker on the power supply of the EBF-MOV-557 valve precluded tne valve from operating up to it's locked rotor current (heavy load) capability in accordance with its intended design objective Technical Specification (TS) 3.7.1.2 requires that at least two indivi-dual EFW pumps and associated flow paths be operable (one must be the steam driven EFW system) during modes 1-3. However, from December 3,1985 through June 19, 1986, the operability of EBF-MOV-557 could not be as-sured of operating as designed, rendering the valve, and hence the EFW alternate flowpath of EFW inoperable. This is a violation of TS 3.7. (VIO 86-09-02).

There are several other mitigating circumstances to the postulated even First, if both the normal and alternate EFW flow paths were lost, there were additional (non-tested, non-nuclear safety grade) connections to both the normal and alternate flow paths which could have been utilized as additional EFW backups (from both charging and safety injection sys-tems). The charging pumps can supply demineralized water via a spool piece directly to the main FW header or to the alternate EFW path (into the SGBD line downstream EBF-MOV-557). The SI System can also provide a source of water to the SGBD connection downstream EBF-MOV-557, however, the supply is borated water and should be used only as a last resor Additionally, the Safe Shutdown System (SSS), can provide a separate and independent source of EFW to SGs. The licensee's Emergency Procedure OP-3203, Rev. 13, Loss of Feedwater, provides instruction on using the charging and SI pumps (but not the SSS). Licensee Corrective Action Program Directly following the notification to senior plant management that the defi-ciency was identified, an investigation task force was formed. They were in-structed to investigate the factors leading to and the resolution of the deficiency. The composition of the task force consisted of senior plant man-agers and supervisors, a Corporate QA manager, operational QA and QC managers and supervisors, and a YNSD Engineering manager. The elements of the investi-gation included:

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Determine the underlying factors leading to the even Recommend to the Plant PORC Committee and Plant and Corporation Manage-ment short term corrective actions to assure the operability of poten-tially effected safety related electrical equipment within the Yankee plan Recommend long term corrective actions to preclude recurrence of this incident, or incidents of a similar nature. These recommendations may include, but are not limited to: improvements to personnel training, receipt inspection, test procedure development, post installation test-ing, QC/QA inspection and document revie __

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Provide a final report with recommendations to the Manager of Operation The inspector noted that the charter of the task force included instructions to be open and candid with NRC inspection personnel during all phases of their investigation. The actions identified for consideration on a short term basis ' by the task force consisted of:

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Review the events associated with design, procurement, receipt inspec-tion, installation and testing of EDCR 84-312, and review input from cognizant plant personne Review the actions completed to date to correct the discovered deficienc Review the inspection / verification of all ITE Gould switchgear equipment within the plan Review all EDCR's and Plant Design Change Requests PDCRs involving elec- ! trical equipment since the 1984 refueling outage to determine if a similar event could have occurred in other installed safety related equipment changes.

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Propose any other short term corrective actions deemed necessary to as-sure the operability of safety related electrical equipment at the plan Report the results of the short term corrective actions to Plant Opera-tions Review Committee (PORC) and Plant and Corporate managemen Based upon the identification of additional discrepancies, none of which re- i sulted in a determination of inoperability, the task force increased on June 29, 1986 the scope of the design changes to be reviewed. Additionally, the licensee instituted a review of purchased material in the stockroom and test- , ' ing of ITE Gould installed equipmen t On July 1, 1986, the efforts implemented by the task force's finding from the short term program were discussed at the plant with the Manager of Operations, the Plant Superintendent, and the PORC. A decision to restart was made on this I date based upon a determination that there was reasonable assurance that the - installed plant equipment represented the intended design and that the deft-ciencies identified and resulting inoperabilities were an isolated incident.

! Additionally, the licensee will be reviewing the actions of their vendor, ITE , Gould (now known as Telemecanique, Inc.) for possible 10 CFR 21 reporting.

, The inspector reviewed the licensee's corrective action to resolve the under-j sized trip coils associated with the initiating deficiency, determined that their actions were appropriate, and identified no discrepancie : > ! t

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l i 9 i Persons Contacted and Management Meetings

Interviews and discussions were conducted with members of the licensee staff

, and management during the inspection to obtain information pertinent to the

areas inspected. Inspection findings were discussed with the Plant Superin-

'. tendent at the completion of the inspection. The contacted personnel included the following peopl l'    E. Begiebing, Maintenance Supervisor L. Bozek, QC Supervisor i
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B. Drawbridge, Assistant Plant Superintendent J. Falconieri, Maintenance Engineer

K. Jurentkuff, Assistant Manager Plant Operations

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J. Kay, Technical Service Manager

P. Laird, Maintenance Manager j S. McConnell, Plant Mechanic - Electrical R. Mitchel, Maintenance Support Supervisor

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3 J. Piela, YNSD Engineer i D. Pike, Manager of the Operations Quality Group i A. Shepard, Director of Quality Assurance Department j N. St. Laurent, Plant Superintendent i

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