ML20248B884

Insp Rept 50-443-89/01 on 890110-0227.No Violations Noted. Major Areas Inspected:Operational Safety,Design Changes & Mods,Licensee Reportable Events & Station Info Repts,Maint, Training & Licensee Actions on Previous Insp Findings

ML20248B884
Person / Time
Site:	Seabrook
Issue date:	03/30/1989
From:	Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20248B856	List: IR 05000443/1989001 ML20248B884
References
50-443-89-01, 50-443-89-1, NUDOCS 8904110082
Download: ML20248B884 (18)

Text

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i U. S. NUCLEAR REGULATORY COMMISSION f 4

Region I Report No.: 50-443/89-01 Docket No.: 50-443- License No.: NPF-56 )

Licensee: Public Service _ Company of New Hampshire' 1000 Elm Street Manchester, New Hampshire 03105 Facility: Seabrook Station, Unit No. 1-Location: Seabrook, New Hampshire Dates: January 10 - February 27, 1989 Inspector: D. G. Ruscitto,_ Senior Resident Inspector Approved By: I4 k / Na~_M 3/3c/F'9 Donald R. Haverkamp, Chiqf Dite~

' Reactor Projects. Section 'No. 3C l l

Inspection Summary: j

a. Areas Inspected Routine -inspection by the Senior Resident Inspector. Areas of inspection-included operational safety, design changes and modifications, licensee '

reportable events and station information reports, maintenance, training and licensee' actions on previous inspection findings.

b. Inspection Results An unresolved item regarding a deficiency in storing the ' diesel engine-driven cooling tower makeup - pump was identified by. the inspector. The pump was not being stored in a seismically designed building and therefore did not meet the technical specification requirements when previously entering Mode 4 (paragraph 3.a).

A second incidence occurred of damage to a primary auxiliary building door due to. excessive differential pressure. Additional licensee attention to preventing recurrence of this or similar incidents is warranted (paragraph l 3.b).

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' Inspection S'ummary'(Continued) 2 The . thermal / hydraulic' response of " bottled up" main steam lines continues-to present operational problems. During this inspection a feedwater iso--

lation occurred, the third ESF. actuation related -to steam .line pressure tran si en ts'. It ~ appears that corrective actions. for .such incidents are viewed on an isolated case basis. Future similar occurrences as well as l incidents such ~as the previously reported . reactor' coolant system / refueling water storage tank sluicing warrant a more comprehensive root, cause analysis than these have received (paragraph 4.a).

The report of the NHY Independent Review Team concerning the meteorolog-L ical tower Unusual Event was thorough and. insightful. The scope of the-l initial event was expanded and several useful lessons'were learned that might 'otherwise have been lost. The performance of the IRT continues to be a strength at Seabrook (paragraph 4.b).

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a TABLE OF CONTENTS

'Page

1. Persons Contacted........................................... 1

2. Summa ry o f Faci l i ty and NRC Activi ti e s . . . . . . . . . . . . . . . . . . . . . . 1

a. Resident Inspector Activities.......................... 1

b. Visiting Inspector Activities.......................... I

c. Plant Status.......... ............................ ... 1

3. Operational Safety.......................................... I

a. Plant Inspection Tours (IP 71707)*..................... 1

b. Operati onal Events (1P 93702) . . . . . . . . . . . . . . . . . . . . . . . . . 3

4. Licensee Reports (IP 92700)................................. 4

a. Licensee Event Report 88-008: Feedwater Isolation...... 4

b. Licensee Event Report 88-010: Meteorological Monitoring Tower Instrument Power Fa11ure............ 6

c. 10 CFR 21 Report 88-00-03: Qualification of Agastat 7000 Series Time Delay Relays........................ 7

5. Followup Issues (IP 92701).................................. 9

a. Rosemount Transmitter Potential Failures............... 9

b. Limitorque Melamine Torque Switch Failures............. 9

c. Environmental Qualification of RG-58 Coaxial Cable..... 9

d. Residual Heat Removal System Valve Limit Switches. . . . . . 10 i

e. Containment Building Spray Sump Level Indicators....... 10

f. Radiation Monitoring System Database................. . 10

6. Design Changes and Modifications (IP 37828)................. 11

a. Emergency Feedwater System................... ........ 11

b. Containment Building Spray Pump Suction Piping...... .. 11

7. Maintenance (IP 62703)...................................... 11

a. MOVATS Testing of Service Water Valves................. 11

b. Main Steam Isolation Valve Actuators........ .......... 12 l

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L Table of Contents (Continued)

P_ age

8. Training (IP 41400)......................................... 13

a. Simulator Training for Startup Quality Control Inspectors............. .............. .............. 13

b. Plant Systems Training for Technical Staff............. 13

c. Startup Test Personnel Qualifications and Training..... 13

9. Management Meetings (IP 30702).............................. 14

• The NRC Inspection Manual Inspection Procedure (IP) that was used at inspection guidance is listed for each applicable report section.

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DETAILS

1. Persons Contacted

• W. A. DiProfio, Assistant Station Manager T. C. Feigenbaum, Vice President, Engineering, Licensing and Quality Programs

• D. E. Moody, Station Manager

• N. A. Pillsbury, Independent Review Team Manager

• Attended exit meeting conducted February 27, 1989.

Interviews and discussions with other members of licensee and contractor management and with their staffs were also conducted relative to the inspection of items documented in this report.

2. Summary of Facility and NRC Activities

a. Resident Inspector Activities One full time senior resident inspector was assigned to the site dur-ing the entire inspection period. The inspection activities were conducted during both normal and backshift working hours for a total i I

of 116 and 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />, respectively. In addition, deep backshift inspection was conducted 2:00 a.m. to 5:00 a.m. on February 2,1989.

b. Visiting Inspector Activities On January 31 -

February 3, 1989, an NRC:RI Senior Emergency Pre-paredness Specialist conducted a routine inspection of the NHY emerg-ency preparedness program. Results of that inspection were docu-mented in NRC:RI Inspection Report No. 50-443/89-02.

c. Plant Status i Mode 5 operations were conducted with primary temperature about 140 l degrees Fahrenheit (F), and primary pressure between 75 and 150 psig I with a bubble in the pressurizer and one train of the residual heat removal system in operation.

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3. Operational Safety I

a. Plant Inspection Tours l

While touring the plant the inspector identified that the diesel engine-driven makeup pump and hoses associated with the service water (SW) cooling tower were stored inside the circulating water (CW) pumphouse, which is not a seismically designed structure. Technical Specification 3.7.5.c (Ultimate Heat Sink) states that the portable cooling tower makeup pump system shall be stored to be operable for l

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2 30 days following a safe shutdown earthquake. This specification is applicable ' in Modes 1-4. Since the CW pumphouse is not seismically designed, operability of any equipment located within the building

, during a seismic event is not guaranteed. The inspector discussed this apparent discrepancy with the cognizant' licensee engineer and was informed that design coordination report 87-097 had already been initiated to provide permanent storage for the hoses. The inspector indicated that prior to the next heatup to Mode 4, a method of stor-age complying with the requirements of T.S. 3.7.5.c must be achieved.

Prior entries of the plant into Mode 4 with the current storage arrangement were determined by the inspector to .be potential viola-tions of NRC regulations of low severity level. Based on the non- -

irradiated status of the core which obviates the need for the ulti-mate ' heat sink, the violations are of minimal safety significance.

The licensee intends to submit a licensee event report pursuant to 10 CFR 50.73. This item is unresolved, pending subsequent review of the licensee's corrective actions to prevent recurrence (50-443/

89-01-01).

While touring the plant during routine periodic safety inspections, the inspector noted several minor- material discrepancies. Each inspector concern is detailed below with licensee corrective action.

All . items are of minimal safety significance and none indicate a breakdown of any programmatic control.

(1) In the hydrogen analyzer room, the inspector noted that tempera-ture switch EAH-TS-5763 was electrically determinate. The inspector discussed this apparent deficiency with the licensee who indicated that a design coordination report (DCR 88-195) hed been initiated in December 1988, which included modifications to the conduit installation at this switch. Field work was not yet complete when the initial observation was made. The inspector subsequently reviewed the DCR and verified in the field that the design change including terminations had been made.

(2) In the service water cooling tower, the inspector identified a grounding cable which was not properly connected to temperature element SW-TE-6184. After notification by the inspector, the licensee initiated a work request (89W000660) to reinstall the cable.

l (3) In the primary auxiliary building, the inspector noted what appeared to be a temporary test gauge attached to the instrument line for a differential pressure indicating switch (SW-PDIS- ,

8259). Following discussion with the licensee, work request i 89W000659 was issued to remove the gauge. The licensee will attempt to determine the origin of the gauge and if its instal-

'lation was programmatically controlled.

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'3 L (4) The inspector identified a missing screw on a pull box cover for main steam isolation valve MS-V-90. The conduit in. question is

identified as being environmentally qualified and the cover must therefore be installed. Work request 88WO564 was initiated to replace the screw.

b. Operational and Security Events (1) Damage to Primary Auxiliary Building Door On January 12, 1989, while conducting maintenance on the primary auxiliary building (PAB) air handling system (PAH), a differen-tial pressure was inadvertently generated across door'P601 which was blown off its hinges into.the room containing normal cleanup filter unit PAH-F-16. The pressure differential was created when a worker opened the filter unit access door to advance the roll filter. The single operating normal cleanup exhaust fan suction was then directed to the filter room (PAB elevation 81'-0") through the open access door. With a capacity exceeding 43,000 cfm, 'each of the two installed fans is capable of estab-lishing a damaging pressure differential almost immediately. In addition to the significant personnel safety hazard, this event also breaches the radiologically controlled area- (RCA) poten-tially allcwing an unmonitored release path. Since the RCA is presently established for training only and no contamination of the plant has yet occurred, there were no adverse radiological consequences as a result of this event. However, this is the second occurrence of this kind at Seabrook. This event is sub-ject to NRC followup inspection, including the licensee deter-mination of root cause and corrective actions in conjunction'  !

with their onnoing evaluation for station information report 89-002. The problem fs scheduled to be corrected prior to initial criticality which requires full implementation of the radiation protection program.

(2) QLss of Offsite 345 Kv Line 1

On February 10 1989, an insulator failed on the "A" phase of the 345 Kv Scecie line. The insulator is located in the termi-nation area where a transition is made from the SF6 insulated buses through the air bushings to the tower transmission lines.

The 345 Kv system is not <cfety related; however, Seabrook Tech-L nical Specifications (TS) require one offsite power source to be operable in Mode 5. At t.he time of the failure both the Newington and Tewksbury 345Kv lines remained operable satisfying i the TS. As a result, plant impact was minimal and no safety system actuation was required. A station information report which was initiated to followup this occurrence will include a j determination of the cause of insulator failure and corrective actions. This event is subject to subsequent NRC followup inspection.  !

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L (3) Control Room Ventilation Isolation Actuations On January 21, 1989, a control room ventilation isolation occurred. The cause of the isolation was a failure of radiation monitor RM-6506B located in the east air intake. Licensee analysis indicates that the detector failed. The root cause of the failure is under licensee review. Following detector replacement, the system was restored to normal. A four-hour emergency notification system (ENS) call was made to the NRC Operations Center pursuant to 10 CFR 50.72. A licensee event report (LER) will be submitted.

On February 14, 1989, another isolation was actuated during troubleshooting of the control building air handling system train "A" west air intake radiation monitor RM-6507A. The inspector. responded to the control room, obrerved recovery activities and reviewed the station computer data logger print-outs, main control board indications status and plant logs. The inspector discussed the event with the control room operator and witnessed the four-hour ENS call to the NP.C Operations Center. An LER will be written for this event. The cause of the actuation was a 100:9 fuse in the test equipment being con-nected to the radiation monitor.

All systems responded as required in both events and no adverse effects occurred as a result of these actuations. The LERs and licensee corrective actions regarding these events are subject to subsequent NRC review.

4. Licensee Reports

a. (Closed) Licensee Event Report (LER) 88-008: Feedwater Isolation.

NHY reported an engineered safety features (ESF) system actuation to the NRC by letter (NYN-88165) dated December 30, 1988. The ESF actuation occurred on December 2, 1988, and involved a feedwater isolation (FWI) caused by a thermal / hydraulic transiert in the main steam (MS) system while bypassing the "C" MS isolatioi. valve (MSIV).

(1) Background During post-core load hot functional testing (HFT) in February 1987, a MS safety valve (MSSV) lifted on the "B" MS line (refer-ence NRC:RI Inspection Reports 50-443/87-02, 50-443/87-16). The cause of this 1987 event was determined to be a condition in the MS header where steam condensed in the line and collected during a long period of time. This condition with the upstream steam

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generator (SG) hot and the downstream MSIV' closed is referred' to as the steam line being " bottled-up". ' Corrective . action for.

this occurrence as described.in station information report (SIR)87-023 included procedural' changes to ensure that the MS lines are manually drained whenever they have been bottled up for

~ greater than two hours.

In another related event, a FWI was generated on March! 20, 1987, while opening the "A" MSIV during plant cooldown. A . High-High SG 1evel trip was actuated when SG 1evel swelled as the MSIV was opened. The applicable procedures were subsequently changed to ensure that differential pressures across the LMSIV were . low enough to preclude a level " swell" due to rapid drop in SG pressures (refer to LER 87-010).

(2) Event Description On December 2,1988,-.with the plant in Mode 4 while attempting to bypass the "C" MSIV the control room operator jogged the control switch for the "C" MS bypass valve (MSBV). The MSBV's are four-inch motor operated globe valves. A FWI signal was received about 40 seconds later when level in "C" SG swelled to the High-High level setpoint as pressure dropped from 100 psig to 60 psig. A " Seismic Event In Progress" alarm actuated simul-taneously in the control room The FWI was reset and plant-temperature control established using the MSBV's off ' other MS lines and ensuring that the opening rate was sufficiently slow to preclude rapid pressure drops and the resulting SG swell.

(3) Event Analysis SIR 88-095 was issued on January 25, 1989, and provided an analysis of the event. The SG 1evel swell was caused by opening the "C" MSBV too rapidly. It should be noted that the operator.

was aware of the potential result of rapid MSBV operation and was periodically " bumping" the MSE/ open. He had previously cpened the MSBV on other SGs during that shift to control reactor coolant system (RCS) temperature without incident.

As the MSBV opened, pressure dropped about 40 psig rapidly and the level of water in the "C" SG, which was at saturation condi-tions swelled due to increased boiling. The hydraulic effect of this pressure / level transient resulted in movement of the "C" MS line. The loads imparted by motion of the steam line were applied to the containment through the pipe supports. This shock set off the containment seismic monitor which begins A

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recording at 0.01g and alarms at 0.079 Actual accelerations were later determined from reading the recorder to be 0.12g vertical and 0.10g horizontal, significantly below the design bases of the plant. A piping walkdown was conducted by NHY engineers and although one pipe support evidenced edge contact with another pipo support, no damage was noted to either piping, support steel or snubbers.

(4) Inspection Upon notification of the Pdl by a senior licensed operator, the inspector went to the control room and observed operator recovery activities. He discussed the transient with the oper-ators and reviewed instrument recorder traces of the event.

Operator followup actions were timely and appropriate.

The inspector subsequently met with representatives of NHY engineering, operations, technical support and operational pro-grams departments to discuss the event, corrective actions and how the event related .to previous similar occurrences. It r appears that the procedure changes made after the previous i events did not take into account a condition of extended hot shutdown (Mode 4) operations with very little heat input to the RCS. It is theorized that with only one reactor coolant pump (RCP) operating, there was insufficient thermal driving force to prevent the rapid depressurization of the SG when the MSBV was cracked open. The same sequence of operations with greater RCS heat input (i.e., more RCP's operating) would likely not have caused such a significant pressure drop. Operating procedures have been revised to provide more specific guidance to the oper-ators in how to avoid the accumulation of condensate in the MS lines. Additionally, a long term review of the MS drain system will be undertaken by NHY technical support department. The effectiveness of the procedural changes will be monitored by the NRC during the next heat up to verify corrective actions of this and previous similar events. This LER is closed.

b. (Closed) Licensee Event Report 88-010: Meteorological Monitoring Tower Instrument Power Failure. NHY made this report by letter (NYN-89010) for information only on January 18, 1989. NRC inspection of the declaration of an Unusual Event related to this power failure was documented in NRC:RI Inspection Report 50-443/88-17. Two inde-pendent licensee evaluations of this event were conducted. Station information report (SIR)88-103 was completed on January 30, 1988.

The NHY Self Assessment Team (SAT) completed an evaluation of the event on January 11, 1989. While the SIR dealt with the technical issue behind the failure and addressed corrective actions, the SAT report focused on lessons learned in all areas. The recommendations

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7 of the SAT report focused attention' on .possible enhancements in such I. diverse areas as primary responder rosters, root cause analysis and I

event evaluation and reduction programs, news service ' activities, -

shift superintendent emergency planning tasks, and technical specif-ications and emergency . plan consistency reviews. The SAT effo'rt involved considerable manpower resources and demonstrated the manage--

l ment philosophy that.there are lessons to be learned from every event l regardless of how well the plant staff responds. While the SIR met f the reporting program requirements, the SAT report was a more useful document which clearly demonstrated the organizational benefits which-may be . derived from in-depth, independent analysis of plant events.

It is recognized that this depth of analysis-is not ' required for each plant occurrence, however, similar evaluations may be appropriate in selected future circumstances. This LER is considered closed,

c. '(Cpe:,) 10 CFR 21 Report 88-00-03: Qualification of Agastat 7000 Series Times Delay Relays (1) Background NRC:RI Inspection Report (IR) 50-443/88-15 described licensee activities related to evaluation - and replacement of certain Agastat 7000 series time delay relays. The 88-15 IR stated that-this item remained open pending resolution of the'following four issues:

(a) Review of design change and work control documents which implemented these replacements.

(b) Review of deportability and operability determinations made by the licensee with respect to those circuits requiring -

modifications.

(cl Technical review of engineering evaluation 88-033 to ensure that specific judgements made with respect to categoriza-tion of relays as to corrective action required was appropriate.

(d) Review of NHY actions taken in response to NRC Information Notice 87-66.

(2) Inspection During this inspection a review was conducted of of the first t b m- of the four above issues. Each issue is individually acJn sed below.

(a) The five relays in question were replaced under the follow-ing work requests (WR).

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1 System Relay Location WR Number l MS TDR-1 CP-108A 88W005730 MS TOR-1 CP-1088 88W005731 EAH 62X-180A MCC-521 88W005771 EAH- 62 MCC-512- 88W005770

-EAH 62 MCC-621 88 WOO 5769 The. enclosure air handling (EAH) relays involved a specific design . system change' since " Class 1E" relays were initially specified and the MCC qualification summary report specified 1

"E7000"' series relays. A design change was therefore initiated to justify relay replacement of 7012AE relays with E7012AE relays (DCR 88-187). The main steam (MS) system relays which are used in emergency feedwater (EFW) system valve - circuitry were originally specified to be "E7022PD" type relays so replacement with the ' correct relay did not require a~ design change. Other documentation discrepancies identified in Engi-neering Evaluation 88-033 were corrected in document revision report (DRR)88-033.

b. Since none of the relays which were replaced were known to be inoperable during any operating mode in which operability was required, no violation of NRC requirements existed. Since this issue was previously the subject of an NRC Information Notice and had been reported to. the NRC pursuant to 10 CFR 21 by another utility, no other reporting was necessary. It should be noted that the relays in question, while not fully qualified for the application in which they were used, had been previously qualified for 1E use. As such, their status was indeterminate.

There is no reason to believe that these relays could not have been qualified and left in place; however, the licensee chose the conservative and expedient solution to replace the relays.

(c) The. inspector conducted a detailed review of Engineering Evalua-tion 88-033. He discussed the evaluation with the Lead Elec-trical Engineer and conducted independent spot checks of elec-trical schematic drawings .and control panels. No undocumented 7000 series relays were found. The engineering evaluation was noted to be thorough and detailed. The dispositioning of each discrepancy found, whether it be hardware or documentation related, was appropriate.

Sub-items (1)(a), (b) and (c) are closed. This item will remain open pending NRC review of sub-item (1)(d) above.

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5. Followup Issues

a. Rosemount Transmitter Potential Failures.

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In a letter dated February 7,1989, Rosemount, Inc. infor.ned NHY of a l potential defect pursuant to 10 CFR 21. The defect involves a loss of fill fluid from the transmitter sensing module due to internal leakage. This condition may cause a slowly responding or drifting transmitter prior to detectable failure. Reported failures to date; have occurred within the first 30 months of service and Rosemo' nt believes that the existing data indicate but do not confirm tnat transmitter failures af ter _36 months of operation are unlikely. A records check by NHY personnel revealed two affected. transmitters onsite. One was a spare located in storage and the other is installed in the emergency feedwater. (EFW) system. The transmitter in stores will be returned to Rosemount. The EFW transmitter will be recalibrates. Since it has been in service greater than 36 months with' no detectable problem, it will be used if the calibration is satisfactory. Once Rosemount completes their more detailed analysis on the acceptability of the 36-month time restriction, NHY will re-evaluate the acceptability of the EFW transmitter. If at any time the instrument cannot be calibrated or exhibits sluggish or otherwise abnormal response, it will be replaced. While licensee activities in this area appear well directed, this item is subject to subsequent NRC inspection pending Rosemount evaluation of the 36-month time restriction.

b. Limitorque Melamine Torque Switch Failures.

On November 3, 1988, Limitorque Corporation notified Yankee Atomic Electric Company (YAEC) of a potential defect in melamine torque switches ' in certain Limitorque valves. The YAEC quality assurance department advised NHY of the relevant information and a review was conducted at Seabrook. Engineering evaluation 89-004 issued on February 14, 1989, concluded that certain Limitorque valve motor operators at Seabrook could possibly contain these torque switches.

The environmental qualification files for 110 operators were checked by NHY technical personnel, who verified that no- operators had melamine torque switches. Spare parts from storage were also checked with negative results. Additionally, the subject switches are being added to the restricted material listing to preclude future procure-ment of similar equipment containing melamine switches. Licensee action'on this issue was thorough and timely.

c. Environmental W alification of RG-58 Coaxial Cable On January 17, 1989, the Atomic Safety and Licensing Appeal Board (ASLAB) issued a Memorandum and Order (ALAB-909) which af firmed a prior Memorandum and Order (LBP-88-31) issued by the Atomic Safety and Licensing Board. The ASLAB noted an uncertainty with respect to observance of the color-coding scheme of the twelve RG-59 cables which were installed by the licensee to replace the RG-58 cables

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10 which were in. question. The inspector discussed the cable replace-ment with cognizant licensee technical support department and engi-neering department engineers. He conducted field inspections of all twelve cable installations, verifying that (1) the old RG-58 cables were determinate, capped and spared, -(2) RG-59 cables had been' installed in place of the RG-58 cables and (3) the RG-59 cables were black with a red stripe indicating that they are not safety grade but are associated with the "A" train electrical system. This color coding scheme is fully in accordance with the existing Seabrook. cable identification scheme and was not developed specifically for this installation. The inspector noted that the original RG-58 cables, including. installed spares, were also properly. color coded black with a red stripe. In summary, no deficiencies were noted in electrical  ;

termination, cable installation, cable sparing, cable marking or cable material condition.

d. Residual Heat Removal (RHR) System Valve Limit Switches The inspector conducted a design review of the valve position limit switches on the RHR heat exchanger outlet valves. A construction deficiency report (CDR) pursuant to 10 CFR 50.55(e) k.d been submit-ted to the NRC by another Westinghouse pressurized water reactor in November 1988. This CDR described the improper wiring of limit switch contacts to the valve position status lights in the control room. The inspector reviewed the field configuration of the Seabrook valves, switches and lights and reviewed the associated wiring dia-grams. This review confirmed that the . NHY design differs from the reported design and the problem does not exist at Seabrook.

e. Containment Building Spray (CBS) Sump Level Indicators i

NRC:RI Inspection Report 50-443/88-10 indicated that the licensee would provide labels to identify the CBS sump level indicators in the essential switchgear rooms. On February 3,1989, the inspector ver-ified that both indicators had been labeled.

f. Radiation Monitoring (RM) System Database Station information report (SIR)88-072 was initiated in August, 1988 to evaluate inadvertent modifications to the RM system computer data-base. The SIR recommended purchase of a sof tware package from the vendor. This package would automatically check certain database i items. The inspector followed up this recommendation by verifying l that a design coordination report DCR 88-200 had been initiated to implement the requisite changes.

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'6. Design Changes and Post Modification Testing

a. Emergency Feedwater (EFW) System NRC:RI Inspection Report (IR) 50-443/88-17 described Seabrook design coordination report (DCR)88-189 which involved changes to EFW system valve solenoids. The 88-17 IR indicated that special test procedure (STP) number 116 had been developed. The inspector observed portions of the testing on January 12, 1989. Testing was properly- suspended when a test procedure ~ change was required because: the step, which verified proper operation of MS-V-395 from the main control board (MCB), could not be completed. When valve operation was attempted from the' MCB, the valve did open because the control sw'. tc h for MS-V-395 (MS-CS-3062) is a dual train switch and although valve con-trol via the train "A" circuit was disabled, the train "B" circuit was still operable. The procedure change ensured that the ' local /

remote selector switches in both trains were in the local position prior to verifying that MS-V-395 would not actuate. The test proced--

ure was satisfactorily completed on January 20, 1989. Final startup testing of the EFW system will be conducted during the heatup prior to initial criticality in accordance with STP-101.

b. Containment Building Spray (CBS) Pump Suction Piping An unresolved item (86-54-02) which was closed-in NRC:RI Inspection Report (IR) 50-443/88-10 described licensee plans to resolve . the issue of overpressure protection for a portion of the CBS system.

The 88-10 IR indicated that the NRC Office of Nuclear Reactor Regula-tion presently considers this a post low power licensing issue. Dur-ing this inspection NHY conducted a special test to quantify current CBS check valve leakage. Procedure ES-89-1-4, CBS Suction Check Valve Leak Verification was c6nducted and observed leakage on each train was determined to be less than 0.5 gpm. NHY intends to use i this data in evaluating options available for ultimate resolution of this issue. The inspector reviewed the procedure and test results and walked down the system in the field using current system draw-ings. Design coordination report 87-311 has yet to be issued and additional NRC review and inspection of this design change is a planned.

7. Maintenance

a. M0 VATS Testing of Service Water Valves On February 6 and 8,1989, the inspector observed MOVATS testing of service water (SW) valves SW-V-25 and SW-V-29. Both valves are 24" diameter motor operated butterfly valves. The inspector discussed the procedure with the cognizant technical support and maintenance department personnel at the job sites. All individuals were found to be knowledgeable about the test equipment and procedures as well as the technical details of these Limito: ,ue motor operators.

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i 12 While testing SW-V-25 in' the SW cooling tower (CT), a laborer rein- )

stalling . the butterfly valve body with the.' operator still removed

-received an electric shock. The individual was not. injured. Testing activities were immediately terminated and a voltage check conducted of all the wires which had been de-terminated when the motor operator. i for SW-V-25 had been removed. Approximately 60 volts a.c. was dis-covered on a single lead which most likely contacted adjacent conduit.

and/or .the chainfall being handled by the .1aborer. The wire ' was immediately- insulated with electrical' tape and a re-verification of the isolation tagging order was conducted. All tags were found in place; however, a sliding link (located inside the' train "B'! terminal box) which had been opened to isolate the circuit which was found to 1 be energized appeared to be improperly wired. Both incoming and out-going wires were landed on the same terminal, effectively bypassing the sliding: link which had been opened. The lower terminal of the j link had no terminating wires however the terminal nut was partially screwed on. A similar configuration was found on one other sliding link in that' terminal box. The corresponding train "A" terminal box

' had no similar link configurations. A station information report (SIR) was initiated to investigate this occurrence. This SIR will be the subject of ongoing NRC inspection.

The inspector also inspected the condition of the Belzona seating surface of SW-V-25. It was found to be undamaged. Discussion with the cognizant system engineer revealed similar findings regarding the SW-V-29 seating surface.

b. Main Steam Isolation Valve Actuators 1 New Hampshire Yankee (NHY) was informed by letter dated ,

l January 17, 1989, that Rockwell International Corporation recommended '

certain modifications to the Seabrook main steam isolation valve (MSIV) actuators. Rockwell is the manufacturer of the Type A MSIV actuators. The actuator is opened by applying hydraulic pressure below a piston which compresses nitrogen gas in a dome-shaped accumu-lator. On loss of hydraulic pressure the compressed nitrogen forces the piston downward, closing the valve. The reported problem in- i

, volves possible debonding of the aluminum bronze piston coating.

While this potential problem was evaluated by Rockwell not to be a substantial safety hazard pursuant to 10 CFR 21, it has caused prob-lems at other sites and is of concern. The inspector. discussed the technical issue with the NHY System Support Manager with respect to the availability of replacement parts and the potential safety impact during low power testing. NHY contacted Rockwell to evaluate the

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c13 impact ' of low power testing on the pre'sently installed pistons. By letter dated . January 27, 1989, Rockwell replied to NHY reiterating that this potential debonding was not a significant safety hazard and indicating that' it was unlikely. that any malfunction would occur that .

could d.elay testing or' otherwise cause an operating hardship. Based upon this evaluation, the replacement of the pistons is not required

~ for plant operation below 5's of rated ' power, and will be subject to future NRC inspection.

8. Training

a. Simulator Training for Startup Quality Control Inspectors NRC:RI Inspection Report (IR). 50-443/88-12 documented a NHY commit-ment to provide additional training to quality control-' inspectors.

On November ~ 18, 1988,- the inspector witnessed simulator training for

selected QC inspectors.

While the - instructor to student ratio appeared low, .-the simulator familiarization was beneficial in preparing the inspectors for wit-nessing startup testing.

b. Plant Systems Training for Technical Staff The inspector -held discussions with the Operations Training Manager about the progrcm which was established in late 1988 to provide plant systems training for non-operations department technical staff. Three levels of training are provided. Level I systems training is a generic introductory treatment of the operating principles and char-acteristics of commercial nuclear power plant . systems. This course consists of ten hours of classroom instruction. Level II systems training is a basic systems course which is conducted for one-half day, each day, for about six weeks. This course provides an overview

.of plant systems. The Level III training is a more ' comprehensive course, similar to the systems portion of the licensed operator training program and runs full time, one day per week, for about six 1 months. Openings for each course are made available to technical and.

nuclear quality personnel. This training program initiative is a expected to strengthen the level of knowledge of non-operations department personnel.

c. Startup Test Personnel Qualifications and Training NRC:RI IR 50-443/88-13 indicated that one of the proposed startup test directors did not meet the formal education requirements and that a waiver was being evaluated by the licensee. The inspector reviewed the training and qualification matrix for startup test per-sonnel and evaluated the areas of specific deviation from ANS 3.1 1 I

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14 requirements. In all cases, the equivalent experience of each indi-vidual met or exceeded the intent of the training requirement. The Seabrook startup staff was found to be qualified to perform and direct upcoming low power testing. The qualification of these indi-viduals to conduct testing above 5"s power may be the subject of future routine NRC review.

9. Management Meetings At periodic intervals during the course of this inspection, meetings were held with plant management to discuss the scope and findings of this inspection. At exit meeting was conducted on February 27, 1989, to dis-cuss the inspection findings during the period. During this inspection,  ;

the NRC inspector received no comments from the licensee that any of his )

inspection items or. issues contained proprietary information. No written material was provided to the licensee during this inspection.

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