IR 05000293/1987037

Insp Rept 50-293/87-37 on 870817-21 & 0831-0904.No Violations Noted.Major Areas Inspected:Review of License Followup & Corrective Actions Re Several NRC Previously Identified Items

ML20236K854
Person / Time
Site:	Pilgrim
Issue date:	10/28/1987
From:	Cheung L, Gregg H, Strosnider J, Woodard C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20236K844	List: IR 05000293/1987037 ML20236K842
References
50-293-87-37, NUDOCS 8711090387
Download: ML20236K854 (21)

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O.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report No.

50-293/87-37 Docket No.

50-293

i License No..OPR-35 Priority --

Category 1

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Licensee:

Boston Edison Company M/C Nuclear 800 Boylston Street l

Boston, Massachusetts 02199 i

Facility Name:

Pilgrim Nuclear Power Station l

Inspection At:

Plymouth, Massachusetts l

Inspection Conducted: August 17-21, 1987 and August 31 - September 4, 1987

Inspectors:

& M[

/d// /B7

E Wootrard, Reactor En'gineer

~ da'te Q}

mAz/v H. Gregg,'L"ead Reacta(f ngineer latfe

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&~J t Lg evte E Cheung, Reactor Engineer date Approved by:

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/0/26 8'7 J. Strosnider, ChTef, Materials and

~date Process Section Inspection Summary:

Inspection on August 17-21, 1987 and August 31 -

September 4, 1987 (Report No. 50-293/87-37).

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Areas Inspected: A routine announced inspection was conducted to review the licensee's followup and corrective actions related to several of NRC's previously identified open items.

Results: A number of unresolved items and inspector followup items and three previously identified violation items were closed. A significant number still remain to be closed.

No violations were identified during this inspection.

However, one unresolved item was opened.

8711090307 G71103 DR ADOCK 05000g93 PDR

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L DETAILS 1.0 Persons Contacted 1.' 1 B'oston Edison Company (BECo)

• P. Hamilton, Senior Compliance Engineer

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• K.. Roberts, Nuclear Operations Manager l

• B.' Lunn, Compliance Engineer

• J. Sury, Technical Manager

• J. Mattia, Quality Assurance Surveillance Group Leader

• S. Bibo, Quality Assurance Audit Group Leader

• F.~Famalari, Quality Control Group Leader

• M. Brosee, Outage Manager j

• H. Brannan, Quality Assurance Manger Operations j

• S. Hudson, Senior Operations Manager

• R. Grazio, Field Engineering Senior Manager R. Williams,. Senior Engineer P. Kahler,. Senior Licensing Engineer T. Burns,. Senior. Material Engineer N..Desmond, ISI Quality Control Engineer

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B. Perkins, ISI Senior Quality Control' Engineer

W. Riggs, Design Engineer i

G. James, Senior Mechanical Engineer S. Chugh, Senior Engineer Civil / Structural R. Sherry, Cnief Maintenance Engineer S. Wollman, Principal Operations Engineer 1.2 Nuclear Regulatory Commission (NRC)

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• T. Kim, Resident Inspector J. Lyash, Resident Inspector M. McBride, Senior Resident Inspector i

• Denotes those present at exit meeting.

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2.0 Licensee's Activities Related to Previously Identified Items (Closed) Inspector Followup Item 86-21-07 Erratic and Erroneous Flow sF h Indication for the B RHR Loop.

Beginning on June 28, 1986 RHR loop B flow signals to flow indicator F1-1040-1 and flow recorder FR-1040-7 in the control room became erratic with indications ranging from no flow to full flow with either the RHR B

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or D pumps running.

These pumps share a common discharge header into the reactor vessel and the flow instrumentation detector is in this header.

The licensee determined that the erratic flow indication was a loss of indication and not an erratic flow or loss of flow in the RHR loop B; and therefore, this was not a reportable licensee event.

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' Licensee.. investigation'of the cause'of the problem uncovered a leak' in i

none of the. differential pressure instrument transducer. input. lines. :The d

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jrepair'offthe-leak was1made by the: licensee;under MR 86-352.

The inspector.

j confirmed the. subsequent operational tests which, assure correct

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.Thifs. item 'is closed.

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J(Closed) Inspector Follow Item 85-30-12 Containment Pressure Evaluations

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.Following Nitrogen Purge of HPCI Exhaust after DBA

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iLicensee.' performed a safety evaluation,: for modification DCREG-89.which

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added the capability to inject nitrogen into the High-Pressure Coolant

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'! Injection' turbine exhaust line. However, the evaluation'did not' recognize i+

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. that the initial containment design conditions for a Design Basis Accident-

(DBA) would be altered and did not assess the effect that,the higher

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i initial 7 containment pressure might have on the accident analysis and-

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icontai. nment ' integrity;

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The inspector, confirmed that the licensee has performed the. additional

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analysis:to assess the effect of the higher initial containment-l

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t pressure; A review of the~ licensee analysis confirms that the containment

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._ pressure remains within safe design limits.

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JClosed) Inspector Follow Item 86-25-07-Undetected Failure. Mode of the

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D Neutron Intermediate Range Monitor.(IRM)

General. Electric (GE) Rapid Information Communications Services Information

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Letter'(RICSIL) Number 007 reported a BWR/3 multiple failure of 3/4 Ampere fuses which are connected to the 24 VDC power to the IRM channels.

j Replacement of the fuses in.the +24VDC. power supply provided normal J)

operating indications. The -24 VDC blown fuses-were not detected immediately.

Loss of the IRM negative -24VDC power supply will not cause an inoperable IRM trip card indicator light but will cause a signal output

'from the amplifier. This causes the IRM to lock-in above the downscale j

trip and below the upscale trip, regardless of actual core neutron flux.

l GE evaluated this previously undetected failure mode of the IRM, its

'l causes-and its potential consequences, and presented this information in j

Service Information Letter '(SIL) 445 along with action recommendations to licensees.

The inspector reviewed GE documentation of the IRM failure mode and the evaluations and recommendations made in SIL 445 which included analysis for other credible events in the neutron monitoring systems.

s Analysis was made for two design basis events in the startup range, the

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control rod drop accident and the continuous control rod withdrawal transient.

l GE concluded that "inoperability of the IRM scram does not lead to a safety

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concern" and other credible failures analysed provide " adequate warning to the operator that a system has malfunctioned."

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The inspector confirmed the licensee implementation of the GE SIL recommendations by a review of Plant Design Change PDC-21, Revision B, and supporting documentation. Changes made included the replacement of IRM power supply 3/4 Ampere' fuses with 1 1/2 Ampere fuses.

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p This item is. closed.

.(Cl o sed)_ _ Viol a ti on 87,04-01 Surveillance Tests Did Not Demonstrate

Proper Operation of Standby _ Gas Treatment System Fans i

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An. audit of approximately 30 separate licensee Technical Specification surveillance test procedures disclosed one instance where the test procedures appeared to be inadequate to meet Technical Specification

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requirements.

Technical specification 4.7.8.1.a(4) requires that, at least once every 18 months,.each branch of the Standby Gas Treatment y

(SBGT) system be automatically initiated and the SBGT fans operated at

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4000 CFM110% subsequent to this initiation.

The 18 month tests previously

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used by the licensee performed automatic initiation of the SBGT but did

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not operate the fans.

-The inspector confirmed licensee corrective actions as follows.

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Tet,t Procedure 8.7.2.1 Measurement of Standby Gas Treatment Filters and

Fan Capacity has been-revised (Revision 10, dated September 18, 1987) to

incorporate the Technical Specification 4.7.8.1.a(4) requirements for-the

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automatic initiation of each branch of the SBGT. system including operating-

the fans at 4000CFM 10% at least every 18 months,

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The inspector verified the conduct of these tests by a review of the test

report dated January 25, 1987.

The inspector also confirmed the

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requirements f or performing this test again prior to fuel load and

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reviewed the results of this test which was completed on September 1,1987.

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Based upon the licensee's implementation of both the procedural changes

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as well as the conduct of the required test to meet the Technical Specification requirement, this item is closed.

[ Closed) Inspector Followup Item 86-29-06 Electrical Cable Insulation Damage The licensee discovered severe insulation damage on the electrical instrumentation cable to the main steam line low pressure sensor PS-261-30C when removing the cable from its conduit during the installation of new cable for the Analog Trip System modifications.

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The' licensee found.a main steam valve stem packing leak which had caused I

steam to impinge upon the surrounding area including the electrical j

conduit which subjected it to excessive heat.

The licensee did not

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attribute any operational problems to the cable insulation damage (1.icensee test ' F&M86-248 verified cable integrity).

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The inspector reviewed licensee actions taken to correct this problem and i

'to prevent recurrence as follows.

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Failure and' Malfunction Report F&M 86-248

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Failure and Malfunction Report F&M 87-273 (walkdown inspection

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report of the extent of the problem of cables..in. proximity to high_ _

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temperature sources)-

Maintenance Request MR 86-1-53 (Valve repacked)

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Cable replaced and conduit relocated, Plant Design Change (PDC) 84-70

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Deficiency identification and correction training program for.

operations section personnel, NTO 87-12550.

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Electrical Raceway Design / Installation Specification E-347, Revision E4 (establishes cable separation criteria from hot pipes / devices).

No unacceptable conditions were' identified.

This item is closed.

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{ Closed)UnresolvedItem 86-25-06 Hydrogen Recombiner Safety Evaluetion Discrepancies.

The'NRC safety evaluation dated April.30,1986 which supports an exemption from the hydrogen recombiner requirements of 10 CFR 50.44 contained statements which did not accurately reflect the condition of the plant.

Specifically items 2.b.1 and 2.b.2 in the evaluation might not

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reflect actual practices at the plant. These items involve (1).a

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requirement to shut the plant down within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the nitrogen supply

(or an alternate nitrogen system) is not operable and (2) the isolation

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of the instrument air system from the nitrogen supply system by a locked

closed. valve.

The inspector reviewed the NRC safety evaluation and the licensee i

documentation which leads to the current as-installed system for the control of containment combustible gases.

The inspector found that the licensee has installed 20 nitrogen cylinders as a backup supply to the normal nitrogen supply for the operation of in-containment instrumentation.

i These cylinders provide a minimum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of operation.

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has also installed the capability to provide nitrogen into the system from

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a new liquid nitrogen / vaporizer trailer within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The trailer will supply sufficient nitrogen for 7 days of operation which provides time for obtaining additional truck delivery. The licensee has modified the containment instrument operating gas inlet valving to provide nitrogen l.

from the normal system until pressure declines from 120 psi to 110 psi and i

then automatic transfer to the cylinder supply is initiated with an

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h accompanying control room indication.

The instrument air valve into this l

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' system is n' w locked shut to isolate drywell. instrumentation from air gation.

o The licensee Technical Specification Section 3.7 Paragraphs 5a and Sb

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Ammendment No; 87 provides the Limiting Conditions for Operation for oxygen concentration in the containment and now clarifies the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> shutdown requirement.

Licensee letter No. 287111 to the NRC da'ted June 8, 1987 provides a

' description of the " Backup Nitrogen Supply System" including the

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objectives of the changes made, design change evaluations, and a safety

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. functions and analysis of the changes made in the system.

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i Th'e inspector's review of the containment atmosphere control requirements, the licensee's implementation of the changes, Technical Specification limitation, and the safety evaluations did not reveal any areas of discrepancy or requirement for further actions.

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This item is closed.

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(Closed) Inspector Followup Item 85-27-47 Electrical Cable Degradation in Connections to High_ Power.Valcor Solenoid Valves IN-84 68.

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'On March 19, 1984, Bechtel-Power Corporation reported under 10 CFR 21 a design deficiency in field run cable to Valcor solenoid valves at Callaway

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and Wolf Creek nuclear p; ants.

The field-run cables were terminated inside a totally enclosed valve tody, housing a large energized solenoid.

No means were prc Fded to dissipa e the solenoid generated heat from

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extended period of energization.

Valve qualification documents revealed that the ambient temperature inside the valve body can approach a maximum of 250-230 F.

The field run. cable used to connect the valve solenoid had W

an insulation temperature rating of 194 F.

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J Pilgrim has 40 Class IE Valcor valves in the Post Accident Sampling System that are potentially affected by the deficiency.

The inspector reviewed the qualification and test reports for the Valcor solenoid valves and connecting cables and the licensee's maintenance program for these valves.

The inspector confirmed that the licensee has taken appropriate actions to reduce the solenoid valve body temperature of solenoid valves which remain energized for sustained operation.

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the valve manufacture, supplied voltage control modules which have be3n installed in the solenoid circuits where prolonged energization could

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cause elevated temperatures with resulting cable damage.

These modules provide the solenoids with full voltage for initial valve movement and then af ter a time delay drop the valve solenoid voltage down to a hold voltage which reduces the heating within the solenoid body proportionally.

Valcor Report MR 526-5683-20-2, Revision A, shows a qualified life of 40

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years for the coil and components inside the valve body housing, with tha maximum temperature not exceeding 194 F under any of the conditions of i

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.o operation at Pilgrim. :The inspector confirmed that the licensee has taken appropriate action to include periodic surveillance for.the new voltage i

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modules in the maintenance program.

The inspector confirmed that the

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licensee uses both Rockbestos Type TC and Anaconda Type AP63571 cables to

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The Wylie. Laboratories test reports 47066-CAB-5

and 47066-CAB-15 provide qualification for these cables up to 194*F.

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This item is closed.

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(Closed) Inspector Followup Item 86-21-06 " A" Recirculation Motor Generator

. Set Field Breaker Failures.

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' Pilgrim Station has three GE type AKF-2-25 generator field circuit breakers; two are in the Reactor Recirculation System motor generator sets end the

third is in the Main Generator Excitation System.

All three have failed

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in the same manner, ie the breaker receives a signal to open, energizir.g the trip coil, and fails to open.

This results in the breaker main contacts remaining closed and the trip coil burning out.

General Electric analysis of the cause of failure (Report 6-HR-6-379 dated December 8, 1986) did not disclose a root cause. The report opinionated

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that the failure was due to a combination of " mechanism misadjustment and improper lubrication".

GE included recommendations in this report to the licensee for lubrication, preventive maintenance and overhaul of these circuit breakers.

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The inspector reviewed licensee actions taken to resolve this problem as

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follovs:

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Corrective action report 87-526 dated May 29, 1987 establishes a program to procure and maintain three spare breakers on site.

It requires the overhaul / refurbishment of these breakers by GE personnel each operating cycle.

Further if a breaker malfunctions during a

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cycle, it is to be removed and sent to GE for analysis and overhaul.

Recommendation is made to exercise the recirculation field breakers each 6 months when station conditions permit.

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Plant Design Change PDC 87-30 dated May 28, 1987 describes changes made in the recirculation pump logic which includes the addition of ATWS trip of the MG drive motor circuit breakers (in addition to the present MG field breaker trip).

The change was made to increase the reliability of the trip by tripping either the MG set motor or the field breaker which will effect recirculation flow control.

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The addition of.these' circuit breakers to the licensee's preventative maintenance items list. This requires' overhaul by GE each outage and

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semi-annual exercise of the breakers.

l Based upon the licensee actions taken to improve.the operation and j

reliability of these circuit breakers and the systems in which they

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operate, this. item is closed.

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'f (Closed) Inspector Followup Item 86-14-05 Trip Failures of_GE Type

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AK-2A-25 Circuit Breakers On two occasions during May 1986, 480 Volt load circuit breakers failed to protect their motor control centers by not tripping to clear overload.

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each case, the motor control center (MCC) was de energized by its feeder breaker tripping to clear the overload.

i Analysis by the licensee and GE revealed a damaged MCC feeder-circuit breaker trip unit.

It alao disclosed a basic problem with this type of circuit breaker in their.mec:!anical trip. These units have a history of high failure rates and also settir.g and coordination difficulties.

This troublesome mechanical. type trip unit is no' longer in manufacture by GE

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and.nas been superseded by a solid state " Micro-Versa" trip device to overcome these deficiencies.

The inspector' reviewed licensee corrective actions as fnllows:

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Implementation of Plant Design Changes PDC-85-42 and PDC 87-15 programs to replace all of the motor cor. trol centers (MCC) AK-2A-25 circuit breaker trip units in accordance with GE recommendations.

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Replacement of the mechanical trip units by GE with the new solid

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state " Micro-Versa" trip unit in the MCC circuit breakers under Maintenance Request MR 86-43-533.

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Performance of trouble shooting and coordination reset under MR 86-309.

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Performance of engineering evaluations, approvals and safety evaluations for the changes to the Micro Versa trip under ESR 86-203.

The inspector confirmed that all of the AK-2A-25 GE-MCC circuit breakers have been modified by GE, tested and accepted.

Based upon the licensee /GE analysis of the problem, immediate corrective actions for the initial culprit circuit breaker and the implementation and completion of the upgrade program for the circuit breakers, this item is closed.

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i (Closed) Unresolved Item 65-26-04-Potential Adverse Affect of

.hultiple Grounds in 125VDC to Environmentally Unqualified Non-i Safety Equipments.

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During September 1985, the licensee traced'a ground in the 125VDC station power system to a valve position limit switch on RCIC steam

line drain valve A0-1301-32.

Insulation at this switch had been

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degraded by moisture from a steam line through the valve stem packing.

A review of maintenance records revealed prior-grounding problems with this same type limit switches due to-packing leaks. The limit switches affected are not environmentally qualified for operation in a steam environment and are n-ot safety related. However, they are powered from a 125 VDC bus which supplies power to safety-related loads and there was concern for the potential adverse affect on the power supply and the safety related loads.

The inspector reviewed the evaluation conducted by the licensee j

which revealed the following facts.

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The 125VDC system is.a floating (ungrounded) system. A single ground will not prevert it from supplying its loads; however, l

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two or more grounds will cause problems in the supply, i

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A single ground actuates a "125V Control Battery A Ground" alarm on Alarm Annunciator Panel C3 in the control room.

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Licensee procedure 2.3.2.14, Alarm Procedure - Panel C-3 Center Control Room, directs operator action to isolate and clear the i

ground.

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The licensee concluded that this failure is a maintenance issue for which the root cause is steam packing leaks which are unrelated to the normal environment.

Improved valve preveative maintenance and housekeeping programs were implemented and considered to be suf ficient to address this type of valve packing leak problems.

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There has been no recurrence of this type of Ground in tne 125VDC system.

Based upon the licensee's analysis and corrective actions taken to resolve this problem, this item is closed.

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(Closed) Inspector Followup Item 86-33-01 Deficiencies in PNPS i

Emergency Categories and Associated Action Levels Review of the PNPS Emergency Categories and Associated Action Levels

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had. revealed several instances where the initiating conditions did

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not conform with the guidelines contained in Appendix I of NUREG 0654.

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lacking.

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-The inspector reviewed the deficiencies cited and additional f

deficiencies that were found by the licensee and confirmed that they

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are now addressed in Licensee Prncedure No. 5.7.1.1 Emergency

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Categories and Associated Emergency Action Levels.

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The specific deficiencies identified and the current location of procedural steps addressing the deficiencies is presented below:

Deficiency Procedure 5.7.1.1, Rev. 6

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Unusual Event

• -No reference to LOP Att. A, Step 5.

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• Less conservative fire Att. A, Step 6.

e initiating condition I

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• Earthquake initiating Att. A, Step 7.a.

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condi;;on inconsistent

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with App. 1 j

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• No reference to fuel Att. B, Step 2.

damage accident

• Less conservative fire Att. B, Step 6.

initiating condition

• An earthquake initiating Att. B, Step 7.

condition inconsistent w/

App. 1

• Omissions of initiating, Att. B, Steps 7. & 8.

conditions from natural phenomena or other hazard Site Area Emergency

• Incorrect initiating conditions Att. C, Step 4.a.

for failure of reactor to shutdown a

• No reference to damage Att, C, Step 2.

of spent fuel in containment

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Omission of initiating,

.Att. C, Step 6.

a conditions for natural phenomena or other hazards

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General-Emergency

Incorrect initiating condition Att. O, Step 4.c.

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for failure of the reactor to shut down The inspector also confirmed that the associated individual procedures 5.7.1.2, Unusual Event, Revision 6; 5.7.1.3, Alert, Revision 5: 5.7.1.4,

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Site Area Emergency, Revision 6; and 5.~7.1.5, General Emergency, Revision 8 have been revised to be consistent with the above tabulated changes to Procedure 5.7.1.1 Emergency Categories and Associated

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Emergency Action Levels, Revision 6.

By inspection, the controlled

copies of the procedures manual were found to include these updated

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procedures.

In addition, an enlarged copy of the revised Emergency Action Levels charts was posted for ready access on a wall in the control room.

By a review of operators training records, the inspector confirmed that all the Senior Reactor Operators had completed the necessary training to the revised procedures by March 30, 1987, i

This item is ciosed.

(Closed) Potential Enforcement / Unresolved Item 85-35-01 Pertaining to the Qualification of BartLn 288A Flow Switches ID#DOIS-261-2A

through 2S.

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These flow switches were used to detect the Main Steam Line Break

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(MSLB) outside the primary containment.

The inspector reviewed the

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EQ file for the Barton switch which contains a similarity analysis between the installed model (Modified Model 278) and the tested model (Model 288A).

The inspector reviewed this similarity analysis entitled i

" Similarity of Barton 278 Modified by Boston Edison to the Barton l

288A" dated December 12, 1985 and had no further concerns.

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The licensee performed an evaluation on the post accident environment

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conditions in the balcony of the RCIC compartment where these flow i

switches were located.

The evaluation indicated that these switches

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would complete the required safety function in 0.5 second following a

MSLB outside the primary containment (the only accident these switches

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are required to mitigate), and the total temperature rise in this

duration is about 2 F.

The replaced Barton switches should have no j

hccuracy problems for such conditions.

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.The EQ file contains a' document indicating that the installed Barton

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switches used Viton 0 rings which were calculated to have a 40 year i

qualified life.

Revision 3 of the Equipment Qtialification Evaluation Sheet (EQES) does show that the qualified life had been corrected'

from 15 years to 40 years.

This' item is closed.

j (Closed)'Open Item 85-35-03 Pertaining to procedural Control of Changes to EQ Documentation NRC Inspection 85-35 identified an inconsistency between the BEC0 QA

Manual and Procedure NEDWI 310 in the procedural control of changes

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to EQ documentation.~

Licensee procedure NEDWI 310 Section 6.1.2 states

that the master EQ Evaluation Sheet is considered a " working Copy" and will contain " whiteout," correction tape, erasures, etc; the record copy used for documenting qualification is a copy of the master

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and is considered the original.

The BEC0 QA manual requires that QA

records corrections or supplements be performed in accordance with procedures which provide for appropriate review, approval, the date and the identifications of the person authori7.ed to issue such a

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correction.

Procedure NEDWI 310 required modification to be consistent with the.QA manual.

The inspector' reviewed'the updated procedure NEDWI 310 (Revision 1 dated April 3,1986) which now includes.a requirement in Section

6.3.2 for preparation of an EQ documentation cover-sheet which

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includes the descriptions, data, dates, and approvals required to be

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corsistent with the BEC0 QA manual.

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This item.is closed.

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{CJosed) Open Item 50-293/85-35-04 Perta ning to the EQ Maintenance Prg r_am r

"his open item concerns that 1) N00 procedure No.1.5.3 did not contain explicitly special EQ handling requirement and 2) some EQ maintenance y

procedures were not yet issued at the tir of the EQ inspection in December 1985.

The inspector reviewed NL 3 we 1.5.3 Revision 17

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dated July 30, 1986 and verified that sect a/.

A.6.C contains a statement requiring the staff engineer to es'lt".1 the maintenance request for special handling requirements, such as EQ.

At this time, all EQ maintenance procedures had been completed and issued.

Twenty-four are dedicated EQ maintenance procedures, while others are included in the component specific preventive maintenance procedures. The inspector concluded that Pilgrim has a complete EQ

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maintenance program and this item is closed.

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'(Closed) 0 pen item 50-293/85-35-05' pertaining to discrepancies

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, identified in 3 EQ files _

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For'the Yarway level. switch terminal block, the old EQ file did not contain a si'ailarity analysis between the installed terminal-block and the test terminal block.

The inspector: reviewed the EQES for.this component which references a similarity analysis

~ dated December 19,'1985, which identified both terminal blocks

'to be TRW Cinch-Jones terminal blocks.

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EQ file for Fenwal Temperature switch TSW-1A and IB referenced a

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terminal block instead of snap acting switch, and the Isomedix certificate of testing was missing.

The inspector reviewed the

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EQES for these components and noted that Note No. 10 has been

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revised to read " Field wire is connected to internal switch of

.the controller......." and that the Isomedix certificate of testing is in the EQ file.

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EQ file for Static-0-Ring Pressure Switch PS-1360-98 was identified'to contain an error (105 C) for aging calculation.

However, the actual aging calculation did use the proper

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temperature 150 F.

The inspector verified that this error has been corrected in the EQES for this component.

Based on.the above verifications, this item is closed.

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{ Closed) Inspection Followup Item 50-293/85-30-07 Pertaining to Lack of Maintenance program for Limitorque Valve Actuators and Limit Switches

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It was identified by the NRC inspector at the time when this j

item was opened that no maintenance procedures existed for i

Limitorque valve actuators except the vendor technical manual which were used by the plant personnel for maintenance activities.

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item was updated in Inspection Report No. 50-293/86-24.

In response to this findir,g, the licensee has issued the

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following procedures for Limitorque valve actuators and the associated Limit Switches.

Procedure No.

Title, Revision, Date j

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3.M.3-24.1 Limitorque type SMB Valve operator removal,

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Revision 4, dated June 2, 1987.

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3.M.3-24.2 Limitorque type SMB-000/00 motcr operator i

overhaul,. Revision 2, dated January 15, 1987.

3.M.3-24.3 Limitorque type SMB-0 through-3 motor operator overhaul, Revision 2, dated January 7, 1987.

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3 M.3-24.4'

Limitorque type SMB-5 motor operator

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overhaul, Revision 1, dated March 28, 1987.

3..M.3-24-5 Limitorque type S8/SMB electrical check out

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and adjustment, Revisien 5, dated May 7, 1987.

3.M.3-24.6 Limitorque type SMB valve operator installation, Revision 3, dated June 1, 1987.

3.M.3-24.7 Limitorque type S8-00 through SB-3 valve operator disassembly / removal Revisien 0, dated February 20, 1986._

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d 3.M.3-24.8 Limitorque type S8-00 through SB-3 valve l

operator reass6mbly/ installation, Revision 1 g

dated March 2.,

1987.

J.Q.3-8 Limitorque type SB/SMB valve operator EQ

maintenance, Revision 3, dated July 7, 1987.

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The' licensee has also developed a computer prugram entitled

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" Preventive Maintenance Item List" which identifies all items requiring preventive maintenance (Lfmitorque valve motors were

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included). The inspector reviewed a copy of the computer prirtout i

of this program and noticed that it identifies the Limitorque j

valve by component number, required procedure number, maintenance s

frequency, date last maintenance performed and the next due date.

In addition the licensee has developed Procedure 1,8.2, PM Track.ing Program, to ensure each preventive maintenance activity is performed l

on time.

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Based on these corrective actions taken by the licensee, this item is closed.

(Closed) Inspector Followup Item 85-28-06 Pertaining to High Drift l

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Problem with Fenwal Temperature Switches.

On October 24, 1985, the licensee had found that the set poir.t for a HPCI area temperature switch TS-2373B drifted aoove the Technical Specification limit of 170 F during a routine surveillance test.

In response to NRC's concern, the licensee committed to take the following actions to strengthen the calibration process:

Switches will be stored in individual padded compartments in

the IGC lab drawers.

The switch storage drawers will be locked.

• The switch calibration will be performed just prior to

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installing the switches, rather than just after removal from the plant.

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i Switch calibration data sheets will be incorporated into the e'

station procedures. - Currently, this data is kept separately on forms that are not controlled by procedure.

The inspector physically observed the storage crea for these instruments located in the Reactor Building at elevation 51' and verified that the temperature switches were stored in individual padded compartments in the I&C laboratory drawers and the drawers were locked. The licensee also revised the following procedures to require the temperature switches'to be calibrated within 22 days of installation and that the temperature switch serial numbers as well as the as-found and as-left data be recorded in these procedures:

Procedure J.M.2-1.2.2, Reactor water cleanup area high

temperature Procedure 8.M.2-1.4.1, Main steam high temperature.

• Procedure 8.M.2-2.5.3, HPCI steam line high temperature.

• q Procedure 8.M.2-2.9, Safeguard area high temperature functional

& calibration.

The inspector verified these procedural changes by reviewing Attachments B and C of these procedures and had no more concerns.

i This item is closed.

{0 pen) Potential Enforcement / Unresolved Item 85-35-02 Pertaining to the Environnerital Qualification of GE E8-25 Terminal Blocks (File No.

.C158).

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The old EQ file based the qualification on the test results of GE EB-5 terminal blocks (tested by Lim torque for use in Limitorque d

valve operators) and supplemented by sitailarity analysis.

Insulation j

resistance (IR) as low as 500 ohms was measured during the test of I

EB-5 terminal block and the effect of this low resistance was not f

addressed in the old EQ file.

In response to this finding, the

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licensee had updated their EQ file. The Limitorque test report for j

EB-5 terminal block was deleted from the reference list.

The licensee

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uses a new test repcet by Wyle Laboratories, Report No. 17775-.1, entitled " Test report on GE EB-5 and GE EB-25 terminal blocks for use in Niagara Mohawk Power Company's hine Mile 2 Nuclear Power Plant",

dated December 31, 1985, to support the qualification of EB-25

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terminal blocks.

The Equipment Qualification Evaluation Sheet in I

this EQ file was revised accordingly.

The inspector reviewed the new test report and noted that the test profile does envelop the Pilgrim

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required profile.

However, the effect of low IR (e.g., the IR measured on October 0, 1985 at 16:44 was 52000 ohms) was not addressed nor evaluated to determine whether it is acceptable in the actual instrumentation circuits.

Instead, the licensee used the leakage j

current of 0.02 mA f.or their evaluation.

According to the test setup as shown in Figure V-12 and other revised diagrams in the test report, it could not be determined that this represented the total leakage current through the terminal blocks being tested.

The licensee further i

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stated that they had. contracted Bechtel to perform an evaluation of the effect of low IR~on the actual instrument circuits (total loop

circuit, including cable, terminal blocks etc.).

The Bechtel report

on this. evaluation has been issued and was being reviewed by the licensee (not ready' for NRCs review).

This item remains open pending

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licensee's inclusion into their EQ file of the evaluation of the lowest IR measured during the test.

(Closed) Inspector Follow Item (86-34-05) Piping Surface Examination

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Indications

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The licensee's second 10 year ISI inspection program imposed a

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requirement for surface examinations of welds from the newer 1980 Edition winter addenda of ASME Section XI (previous code requirements were only. volumetric examination).

This resulted in a number of linear indication findings and a need for licensee's evaluations and j

. corrective actions.

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The licensee surface inspections resulted in approximately 35 NCR's l

which described unacceptable linear indications that were identified by MT or PT examination. The inspector determined that all the NCRs of linear indications were closed out and that in each case rework

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by surface dressing removed the indication.

The inspector also noted

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that wall thickness was above the minimum requirement in all cases.

The inspector reviewed specific NCRs (NCR 86-57, 61, 62, 65,

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70.and 77) and verified that each was acceptably dispositi.oned by

'l minimal dressing of the indication.

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This item is closed.

(Closed) Violation (87-15-01) Failure to Identify Relevant Magnetic j

P_ article (MT) Indication.

j An independent NRC MT examination of weld DP-23-2-1C foend a linear l

surface indication (3/4" in length) that exceeded ASME Section XI, l

Table IWB 3514.2 limits.

The licensee hypothesized that the NRC inspection that identified the j

indication was done after an additional UT surface preparation, j

whereas the licensee's examination was done prior to the additional (

surface preparation.

In order to assure that a similar problem

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would not exist the licensee's QC group reviewed records and identified three other welds that had additional surf ace preparation

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after MT inspection.

These three welds were reinspected and no i

unacceptable indications were identified.

I The licensee's corrective actions were:

1) to revise BECO Quality Control Instruction QCI 20.41 to require that all weld surface t

preparations be completed prior to being released for inspection, 2)

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to further revise this QCI to document that surface preparation is completed. prior to starting examinations and to require reinspection if subsequent surface preparation is performed, and 3) to require that all BEC0 NOE personnel sign-off that they have reviewed and understand the contents of the QCI.

The inspector verified that the licensee's procedures have been revised to include the above changes.

This item is closed.

(Closed) Violation (87-15-02) Inappropriate Experience Requirement for Level 11 Visual Examiners.

The ISI contractors procedure QC-7, Rev 3 required only 4 months experience for Level II visual examiners which is less than the AShiE Section XI and the ANSI N45.2.6 1978 Edition requirements committed to by the licensee.

The licensee's corrective actions were:

1) to revise BEC0 Quality Control' Instruction QCI 20.40 to require that experience certification of all NDE examiners be in accordance with SNT-TC-1A and' ANSI N45.2.6, 2) to further revise QCI 20.40 to require BECO j

review Contractor NDE procedures and personnel certification, and 3)

j to revise contractors procedure QC-7 to conform to the ANSI N45.2.6 l

personnel experience requirement.

j The inspector verified that the qualifications of contractor personnel performing NDE during RF0#7 had a minimum of 4 years experience and that the procedure revisions were in place.

I This item is closed.

[ Closed)UnresolvedItem(83-26-02) Fuel Pool Heat Exchanger Support Adequacy The NRC concern related to the licensee's use of an assumed and non f

conservative 500 lb nozzle loading as a bounding limit when actual loads were already available.

The 1-icensee re-evaluated the loadings and determined the weight contributions initially used in the evaluation were non conservative.

The re-evaluation increased the nozzle loading from 500 lb. to 1110 lb.

The licensee also performed a walkdown verification to verify that the as-built structure reflected the re-evaluation calculation.

Additionally the re-evaluation utilized the computer code "STRUDL" which resulted in a more acceptable form of analysis than the original calculations.

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L 10250-083-001)land. verified that.theLstresses on structural' members;

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that the; forces on-the' welds joints was less than the-allowable

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' forceh in each;: instance the.' allowable stress 'and allowable, force.

l-was not; exceeded.

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<This item isLclosed, y

JClosed)'Unresol'edItem(84-21-02) Liquid Penetrant Acceptance v

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Standards for Control-Rod Drive-(CRD) Collet Housings >

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l0 The NRC~had1 identified the need-for the licensee to have a detailed

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' procedure to specif' ally address the. penetrant testing.and acceptance.

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criteria for.,the. inspection of CRD, housings. The licensee'had been usingLa more restrictive interpretation of-GE SIL 139, however, there c

was no formali procedure and.this caused some confusion to the contractor Additionally,:there have been many revisions to SIL'139'and three different CRD. housing. designs since the-first SIL identified CR0 housing cracking.

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i The licensee communicated their concerns to GE based on their-r

' findings of crack indication _ in the' heat affected zone of the

- upper to lower tube weld of the intermediate CR0 housing design

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'(9190258G2)'. RGE by letter dated February 11, 1987 responded to

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'the; licensee and agreed with the licensee that no crack indications arefallowed in the weld area.

The licensee's' corrective action was

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formalized in'their QC.-Instruction 50.12, Liquid Penetrant Examination of CRD. Collet' Housing, a new procedure that was issued for. Refueling Outage 7.

-TheLinspector reviewed the licensee's procedure 50.12.and verified thattit: f1) was specific for. CR0 Collet Housings, 2) covered each of ~

the dif ferent designs.of CRD housings, and 3)' contained appropriate acceptance criteria.

The licensees procedure also included the l

inspection surface area to be examined for the latest CRD housing (919D258G3) that has.two 0.0. welds.

This item is closed.

[ Closed)UnresolvedItem(84-17-02) Review Evaluation of CRD Collet Housing Cracking During the 1984 outage (RFO-6), the licansee's inspection plan was

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to inspect and rebuild 27 of the 51 older model CRD Housings. The

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remaining 24 older design housings were to be left in place without inspection until a future outage and this was the NRC concern that opened this item. This older design (model 9190258G1) was the

original design supplied by the NSSS vendor and was susceptible to cracking as described in GE SIL 139.

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The ~ inspector determined that during RFO-6 the licensee removed and'

inspected 75 CRD housing of all types (older or original design No. 9190258G1, intermediate or 2nd design No. 919D258G2, ar.d j

'last or 3rd design No. 9190258G3).

Out of the 27 of the older drives inspected about 30% were rejects.

Due to the high reject.

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rate.the remaining 24 older drives were removed and inspected and

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the reject rate was similar to the first reject group. Those j

housings that could be rebuilt were returned to the core or placed in storage and those that could not be rebuilt were scrapped. Of more significance, the licensee identified 4 of the intermediate design of housing to have crack indications. On 2 of the housings the indications were-removed by dressing the surface, however on the_

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other 2 the indications were through wall. These housings were returned to GE for further evaluation because GE, in a subsequent SIL 139 revision, had informed licensee's that due to the improved design the new drives did not require routine inspection.

The inspector reviewed the GE evaluation dated October 15, 1986 of the licensee's findings of through wall indications in the intermediate CRD housing design.

GE confirmed IGSCC cracking at the cold worked and weld sensitized area of the lower tube (this design

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had a swaged or flared lower tube that is welded to the upper tube

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in the swaged location). Due to the crack findings GE revised SIL

139 to rescind their previous SIL. inspection exemption.

The inspector also verified that 21 CRD housing inspected during this refueling outage (RFO-7) included all three design.

Indications found on 6 housings were' surface dressed and the indications were removed.

One housing of the 3rd design model was scrapped due to an

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out of round condition.

Through cracks were not identified on any of these 21 inspected housings.

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The licensee has continued their examination of all type CRD housings, even the latest model (the third design that is not swaged).

The industry has also been advised of the cracking in the intermediate model CRD housing. Additional verification was made by the inspector that to date, there have been no rejects due to crack indication in the 3rd design.

This item is closed.

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(Closed) Unresolved Item (86-31-01) Evaluation of Concrete Wall Rebar Reinforcing in Safety Related Concrete Walls Several discrepancies identified in a non safety related block wall installation led to the concern of potential discrepancies in rebar reinforcing of safety related concrete walls.

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In-response to the NRC: concern the licensee provided the as-built rebar pattern of several concrete wall areas where through wall:

modifications were made.

The modifications involved the mapping or j

determination of the rebar.

r One modification (PDC #85-82) involved cutting a door opening through the:Radwaste building concrete wall at column line K.6.

The rebar encountered (both vertical and horizontal) was sized and i

spaced to _the drawing requirements. Another modification (PDC

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1. 84-03A) involved mechanical and electrical penetrations of the

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Radwaste building concrete wall at column line 20 and P.

For this modification the rebar configuration in these wall areas was determined by chipping away concrete.

Additional chipping of concrete also determined the vertical dowel spacing so that dowels would be cut so as to minimize the cutting of the main vertical rebar.

The as-built determined rebar spacing agreed with the drawing requirements.

The final modification (PDC #85-07) was the penetration of the 6 foot' thick drywell concrete wall for installation of liquid

. level instrumentation. To avoid cutting the outside face rebar,

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f minimal opening of the concrete was made by chipping away of the outside face at Azimuth 58* elevation 82' - 6" and by overlapping two 10" dia core drill holes at Azimuth 198 elevation 82' - 6".

The exposed rebar size agreed with the drawing requirement.

'The inspector reviewed the drawing rebar requirements (from Orawings C-77,~C-112, C-113, C-114, C-115, C-193, C-194, C-195 and C-197)

against the as found modification sketches and verified the agreement of rebar spacing and size.

This item is closed.

i 3.0 RHR piping Corrosion j

l During the inspector's evaluation of open item 86-34-05, NCR 87-318 which i

dealt with a non conformance of pitting corrosion was reviewed.

The ISI l

inspection of weld GB-10-F129 identified pitting corrosion in the pipe l

base metal near the welds.

The pitting corrosion was thought to be the a

result of a prior heat exchanger flange leak and the entrapment of water

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in the piping insulation.

Disposition of the NCR required removal of the

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base metal pitting by flapper wheel or careful grinding.

This resulted l

in a number of the reworked areas being under minimal wall thickness J

requirements.

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The inspector expressed concerns that the licensee's rework procedure may

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not be in accordance with ASME code or licensee's construction and FSAR l

requirements or commitments.

The rework procedure specified weld j

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over. lays of the under min wall areas and over the existing weld with subsequent inspections-of only MT or Pl.

There were no limitations

.regarding depth and area of defects relative to the welding rework or the.

i area of weld overlay over the piping base metal or weld joints.

The inspector did not consider the ' limited coverage of ASME B31.1, the construction Code for this site, or the original A-106 GrB piping material specification, or the use of only one specific test criterion of the later SA-655, piping material specification as being adequate to permit the proposed repair and inspection.

This item is unresolved pending NRC review of the licensee's evaluation of problem root cause, verification that the same pitting problem isn't widespread, and formal evaluation that code requirements and commitment adherence (repair-inspection-testing) is satisfactory (50-293/87-37-01).

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4.0 Unresolved Items l

Unresclved items are matter about which more information is needed to

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determine whether it is acceptable or a violation, unresolved items are j

discussed in paragraph 3.

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5.0 Exit Meeting j

l The inspector met with the licensee's representative (identified

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in paragraph 1.0) at the conclusion of the inspection of September 4, 1987, f

to summarize the findings of this inspection.

The NRC: Resident Inspector, T. Kim, was also in attendance.

t During this inspection, the inspector did not provide any written material to the licensee.

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