IR 05000309/1987001

Insp Rept 50-309/87-01 on 870126-30.Violations Noted:Test Records Did Not Identify Date of Test & Documentation Results Not Evaluated Adequately by Responsible Authority

ML20206A752
Person / Time
Site:	Maine Yankee
Issue date:	03/12/1987
From:	Eapen P, Napuda G, Runyon M, Shannon M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20206A681	List: IR 05000309/1987001 ML20206A677 ML20206A708
References
50-309-87-01, 50-309-87-1, NUDOCS 8704080157
Download: ML20206A752 (14)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report No. 50-309/87-01 Docket No. 50-309 l License No. DPR-36 Licensee: Maine Yankee Atomic Power Company

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83 Edison Drive

! Augusta, Maine 04336 Facility Name: Maine Yankee Nuclear Power Station

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Inspection At: Wiscasset, Maine Inspection Conducted: January 26-30, 1987 i

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Inspectors: >

G.p a da, Lead Reactor Engineer

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/I7 Mate p [Au Aar U SA on, leactor Inspector, NRC Region II I 'Th 7 / fate

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{M.6Shannon,'ReactorInspector,NRCRegionII Y/L D

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Approved by: b-Dr. D. K. Eapen, Chief tW 3 L/ffI

/date Quality Assurance Section, 08, DRS Inspection Summary: Unannounced inspection conducted on January 26-30, 1987 (Inspection Report No. 50-309/87-01)

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Arear Inspected: Design Changes / Modification Control, Plant Surveillances, QA/QC and the effectiveness of quality verification activities, and followup

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on a previous violatio Results: One violation (test records) and one unresolved item (ASME classi-fication cf Primary Containment Cooling Water System) were identified.

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DETAILS

Persons Contacted

, *J. Atkinson, Director of Materials

! D. Boynton, Reactor Engineering Section Head I

• S. Evans, Licensing Engineer 1 *C. Frizzle, Vice President and Manager of Operations

! *J. Frothingham, Manager Operations Department

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T. Gifford, Acting Plant Engineering Department (PED) Section Head

! *J. Grifford, PED Section Head J. Herbert, Manager PED

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M. Huber, Administrative Aide Operations Department

• B. Jameson, Assistant to Manager of Operations i R. Jordan, Reactor Engineer

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• L. Lawson, Quality Assurance Section Head

*R. Lawton, Manager Quality Assurance

• S. LeClerc, Quality Control Supervisor

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D. Lemieux, I&C Shop Supervisor P. Mehlhorn, ISI Coordinator

*W. Peterson, Manager Audit Group (Yankee Atomic Electric Co.)

*G. Pillsbury, Radiation Controls Section Head

D. Ross, PED

., E. Soule, Project Engineer j J. Speed, PED N_RC N. Holden, Senior Resident Inspector Design Changes The licensee's quality assurance effectiveness in the area of design

, changes was assessed through an in-depth review of one recent design

{ chang This approach was chosen to evaluate each element of the design

change process as well as to examine interfaces among the design control

] program and other related programs (such as special processes and

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testing).

The inspector reviewed Engineering Design Change Request (EDCR) 85-24, entitled Appendix J - Primary Component Cooling (PCC) Modifications. This EDCR, installed during the Fall 1985 refueling outage, entailed the upgrade of a section of piping and associated valves from Safety Class 3 to Safety Class 2 (as defined by ANSI N18.2-1973) and the modification of several PCC lines (including the installation of test boundary valves)

that penetrate the containment structur To establish the EDCR's conformity to design base documents, the inspector 4 reviewed the following portions of the Final Safety Analysis Report (FSAR)

l and Technical Specifications (TS):

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FSAR 5.1.1.4 Penetration Design 5.1.2.1 Containment Isolation Design Bases 5.1.6.2 Leakage Rate Test 9.4 Component Cooling System Figure 9.4-3 Primary Component Cooling Water System Table 5.1.2.1 Containment Isolation Data

.T_S 4.3 Reactor Coolant System Leak Tests 4.4 Containment Testing Of these documents, only FSAR Table 5.1'.2.1 and Figure 9.4-3 were affected by the EDCR. Table 5.1.2.1 was revised to indicate that PCC-M-219, an existing automatic trip valve, had replaced four check valves, as the outside isolation for four PCC containment penetrations. Figure 9.4-3 was accurately revised to show all new test boundary valves and the previously mentioned safety class upgrade. The review of base documents confirmed that all required changes were made and that the EDCR was consistent in intent and structure with the design bases and the as-configured status of the plan The inspector reviewed design input specifications and documentation to determine whether they were adequately reflected in design output document One element of the design input was the proposed use of autotrip valve PCC-M-219 as an outside containment boundary. During past performances of penetration leak testing, PCC-M-219 was credited as the outside containment isolation valve that closed upon receipt of a Containment Isolation Signa However, Table 5.1.2.1 of the FSAR identified four downstream check valves, one for each associated contain- l ment penetration, as the outside containment isclation boundary. These four PCC lines feed to reactor coolant pump cooling lines, high pressure drain coolers, pressurizer quench tank coolers, neutron shield tank coolers, control rod drive mechanism air coolers and others. The decision to move the boundary upstream to the auto-trip valve PCC-M-219 was based on facilitating leak testing and to establish conformance with 10 CFR 50, Appendix A, Criterion 56, which does not permit a simple check valve to be used an an automatic isolation valve outside containment. This con-tainment isolation redesignation necessitated the upgrade of piping and valves upstream of the four check valves to PCC-M-219 from Safety Class 3 to Safety Class 2 as defined by ANSI N18.2-1973. The safety class upgrade was determined by the licensee to require no hardware changes; the only changes deemed necessary were to plant drawings, the FSAR, and the upgrading of spare parts to PCC-M-219. The justification for a nonhard-ware upgrade of this portion of the PCC system was based on the following concept All valves in the upgrade area were purchased under the same purchase specifications as those valves presently used on Safety Class 2 systems. The entire PCC system was designed and constructed in accordance

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with ANSI B3 The PCC system was constructed under the same QA requirements throughout. The upgraded piping and PCC-M-219 motor operator i

are seismically qualified and the motor operator is powered from a safety class bus. The only exception in this concept is that QA requirements in effect during original corytru: tion concerning; weld testing would not conform to present-day Safety Class 2 requirements. An example is that

, during original construction lonly 20 percent of the welds were radio-graphed whereas current' requirements, specified in Yankee General Specifications, YA-GEN-1, require 100 percent radiographing for Safety Class 2 butt and branch welds. Considering that the plant was built without reference to safety classes, and that the boundary between Safety Class 2 and 3 was later established arbitrarily at the four check valves, the inspectors had no further questions in this regar The modification of PCC lines that penetrate containment was based on a commitment to the NRC (MYAPCo. to USNRC dated January 11, 1984, GDW-84-10), to revise the Class C test methods for liquid filled piping

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penetrations such that thay conform to 10 CFR 50, Appendix The modifications consisted of the installation of test boundary valves and drain and test connections near eight PCC containment penetrations to i facilitate the use of air or nitrogen as the leak test medium and to

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permit testing of boundary valves in the same direction as would be i required to perform their safety functio The following design output document elements were reviewed by the inspector to verify conformance with design input requirements:

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piping design to conform to ANSI /ASME B3 special precautions to drain and vent sections of the PCC system due to carcinogenic effects of sodium chromate

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pipe stress / seismic evaluation required

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drain valves to be located at low points to permit proper drainage

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low point drain / test connections to branch off main piping as shown in a generic drawing

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where possible, ratio .of main pipe diameter to branch line diameter equal or greater than 4:1. Where not possible, branch connection to

> use a pipe fitting such as a tee

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all branch connections to be capped Based on drawings, specifications, and other design documents, the

inspector verified that the above design input requirements were adequately translated into the final desig Installation instructions appeared appropriately detailed and included QC and field engineer holdpoints. Discrepancies between as-installed and as-designed config-urations were dispositioned with engineering change notices (ECN). This provided a controlled, formal method for assuring that the final hardware configuration matches the as-built drawings. The inspector reviewed the i nine ECNs issued against the original design and confirmed that each was properly controlled and did not change the scope of work or safety evaluation criteri <

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EDCR 85-24 was independently reviewed by individuals from the Plant Engineering Department (PED) and Quality Assurance who were not involved with the. development of the design package. Other reviews and approvals included the Plant Operations Review Committee (PORC), Plant Manager, Operations Manager, and the Yankee Nuclear Safety Department (YNSD).

Several of these reviews generated comments and contingent modifications to the design. In each case, it appeared that review results were taken into consideration and incorporated into the final design packag The '10. CFR 50.59 safety evaluation concluded that an unreviewed safety question did not exi This was based primarily on the assertion that only negligible losses would occur regarding the delivery of PCC water to its cooling loads and that the seismic qualification of the PCC system would not be affected by this design change. The inspector reviewed the calculations which determined the increased PCC system resistance to flow due to the installation of the new test boundary valves. The assumptions and methods employed appeared consistent with typical fluid dynamics models. A spot check of computational accuracy revealed no errors. The predicted flow degradation was clearly negligible as a safety considera-tion. The seismic evaluation results were presented in ECN No. 2 for each penetration and provided either clear acceptance of the design or conditional acceptance based on a modified design or dimensional limi For example it was determined that the maximum distance between new valve PCC-524 and support H-96, to ensure the validity of the seismic analysis, was 35 inches. The actual installed dimension was 34 7/8 inches. Based on the design documents, it appeared that all seismic evaluation contingencies were cleared in the' final design package.

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The inspector reviewed the process whereby engineering provided technical input to the procurement functio This included a review of material issue cards for part retrieval from the storerocm, procurement specifi-cations, letters to and from vendors, and material purchase request Based on this review, engineering support of procurement was adequate to ensure qualit \

, Each design change was reviewed to assure that the required changes to  !

operating procedures and other technical documents were mad With I respect to EDCR 85-24, the inspector specifically verified that the new valves had been added to the Inservice Inspection (ISI) program and that the penetration leak testing procedures had been modified to reflect the new boundary valves. Procedure 1-15-1, Revision 12, was verified to in-troduce new valve PCC-524 to the ISI valve list. The licensee demons-trated that Procedure 3.17.4.11, Component Cooling In Leak Test, had been revised to reflect the new test boundary valves and associated changes to step-by-step leak test instructions.

. The inspector reviewed the design change to determine conformance to weld testing and hydrostatic testing requirement Requirements for weld ,

testing are contained in YA-GEN-1 and ASME Code Section I Test l procedure requirements are dependent upon safety class designation and

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joint type. The inspector reviewed a broad sample of weld data sheets and found that all required weld tests were performed and that Quality Control (QC) involvement was comprehensive. All welds failing inspection were documented on a rework / repair sheet as having been corrdcted and reinspected. Liquid penetrant tests for Safety Class 3 built welds were performed despite not being required, displaying good engineering practice beyond specific requirement Hydrostatic test requirements are delineated in ASME Section XI and YA-GEN- The inspector reviewed hydrostatic testing following instal-lation of valve PCC-524. Although it appeared that the test setup and methodology conformed to the requirements, the required test pressure of 165 to 175 psig could not be obtained. Due to excessive . leakage beyond the boundary valves, the maximum test pressure attained was 125 psig, which was held for greater than ten minutes. The licensee decided that the new welds on PCC-524 were acceptable for service based on verification at a pressure greater than the PCC pumps shutoff head and the non-destructive testing (NDT) examinations performed. Greatly influencing this decision was the need to restore the PCC system to service to avoid ,

certain remedial actions per the (TS) and the need to restore cooling water for waste processin The inspector reviewed this issue for possible impact on hydrostatic test standards, and it was established that prior to startup both the upstream and downstream welds on PCC-524 (with PCC-524 shut) were tested at 165-175 psig in separate tests, thus satisfying in intent the requirements of ASME Section XI and YA-GEN- Other hydrostatic tests were reviewed and verified to be satisfactory. A discrepancy existed between YA-GEN-1, which required a test pressure hold time of 15 minutes times the thickness on uninsulated piping (or 10 minutes, if greater), and ASME XI which requires 10 minutes in all cases.

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The licensee was following ASME XI requirements and YA-GEN-1 was being revised to be consistent with ASME XI.

j The training program for Engineering personnel and the training given to Operations personnel specific to EDCR 85-24 were reviewe '

Plant Engineering Department (PED) personnel receive general plant training and PED indoctrination training delineated in Procedures 0-00-06, General Plant Training Program, and 17-200-2, Indoctrination of PED Personnel Program. The PED training program is described in Procedure 17-200-1, PED Training Program. The training is divided into four major categories:

Operations, Technical, Quality Assurance, and Management / Supervisory Training. The Manager of Plant Engineering is responsible for implement-ing the program and schedules training on a core cycle basis. Based on a brief review of program documents, it appeared that training for PED personnel was satisfactor The inspector reviewed the training package generated for Operations personnel for EDCR 85-24. The package contained a description of each new valve installed in the PCC system and drawings showing valve location It also included revised valve lireup sheets. The package was signed by

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all Operations personnel which indicated they had reviewed it, but it was not clear whether a briefing (which is optional, dependent on scope) had occurred. This activity is controlled by Procedure 1-200-13, Operations Department Interface with Design Change Activities, which appear's to be a comprehensive program for ensuring continuity of system operation following design change Boundary valves PCC-520 and PCC-523 and drain and vent valves PCC-521 and i

PCC-525 were installed in the 6 inch line supplying reactor coolant pump q cooling loads including the pump seal coolers. This portion of the PCC system is designated non-safety class; therefore QA requirements for safety class 3 work were not applie Interpretation of ANSI N18.2-1973,

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Section 2.2.3(1), which assigns a Safety Class 3 (SC-3) rating to components whose failure could lead to the release of radioactive gases to l the environment, suggests that an SC-3 rating may apply to this portion

! of the PCC system. A primary to PCC leak in the pump seal coolers would not cause an automatic isolation (by solenoid - operated valve PCC-A-254)

and could result in a radioactive gas release to the environmen The basis for classifying this position of the PCC system as non-safety was not readily available for the inspector's revie This item remains unresolved pending further review of the licensee basis for the above

system classification (50-307/87-01-01).

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Overall, the design control program appears to be well-controlled in both program and execution with only minor problems eviden Management

} appeared conscientious, highly professional, and dedicated to q'ua l i ty.

j 3. Operations l

The licensee's quality assurance effectiveness in the area of plant

{ operations was assessed through an in-depth review of surveillance procedures, calibration records, discrepancy reports, ar.d repair order The inspector witnessed the performance of Surveillance Procedure 3.17.

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4.6, Personnel Air Lock Leak Test-Type B, and Surveillance Procedure 3-12-5, INCA Core Power Distributio Procedure adherence was maintained during both surveillances thus indicating that surveillance procedures have been written in sufficient detail for performance of surveillance activitie The personnel performing the surveillances performed their duties in a professional manner and were able to discuss in detail the various aspects of the procedure.

The inspector reviewed various surveillance procedures in order to insure l l that they were meeting FSAR and Technical Specification requirement The l following Procedures were reviewed; '

Procedure N .17. Personnel Air Lock Leak Test Type-B, i Revision 11 Procedure N .1 INCA Core Power Distribution, Rev. 8

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! Procedure N . Emergency and Auxiliary Feed Pump Test, l Rev. 25'

l Procedure N . ECCS Routine Testing

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Procedure N .5 - Plant Calorimetric, Rev. 9 Procedure N .6.2.2.16 - Daily Calorimetric Adjustment, Rev. 7

[ The above listed surveillance procedures were sufficiently detailed to

insure.the FSAR and Technical Specification requirements are met.

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After the initial review of Procedure No. 3.1.2, ECCS Routine Testing and

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12.5, Plant Calorimetric, the following instrument calibration records were reviewed in order to insure that surveillance testing was being performed with calibrated instrumentation.

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1 PI 1604C Service Water Discharge Pressure PI 1624C Service Water Suction Pressure PI 3404 PCC Discharge Pressure

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PI 3406 PCC Discharge Pressure PI 3403 PCC Suction Pressure PI 3405 PCC Suction Pressure PI 1709 SCC Discharge Pressure PI 1711 SCC Discharge Pressure

PI 1708 SCC Suction Pressure i PI 1710 SCC Suction Pressure

PTID 1603 Feedwater Flow PTID 1604 Feedwater Flow PTID 1605 Feedwater Flow

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PTID 1606 Feedwater Flow

. PTID 1607 Mainsteam Flow 1 PTID 1608 Mainsteam Flow PTID 1609 Mainsteam Flow

. PTID 287 Feedwater Header Pressure PTID 403 Steam Generator Pressure 4 PTID 404 Steam Generator Pressure

! PTID 405 Steam Generator Pressure l PI 1213 Feedwater Header Pressure (MCB)  ;

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PI 1013A Steam Generator Pressure (MCB)

P! 1023A Steam Generator Pressure (MCB)

j PI 1033A Steam Generator Pressure (MCB)

l PI 1202A Feedwater Pressure (Local)

TI 1203A Feedwater Temperature (Local)

TI 1203B Feedwater Temperature (Local)

TI 1203C Feedwater Temperature (Local)

! TI 1203D Feedwater Temperature (Local)

i The above listed instruments are either used for pump testing per ASME XI or are used in the computer input or manual input to the calorimetric j adjustment which adjust the excore nuclear instruments on a daily basi ,

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All instruments were found to be in the station calibration program and

! all instruments were in calibratio It was noted that the flow instrument technical manual listed calibration frequency at six month It was found by plant staff that these instruments have a large effect' on the calorimetric for minor inaccuracies. Plant procedures now require

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calibration on a quarterly basis. Management has taken the initiative in this area to insure the quality of the nuclear instrument adjustment.

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Maintenance activities were reviewed to insure that adequate corrective action. was taken in problem areas and that adequate post maintenance testing was performe The following discrepancy report / repair orders

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were reviewe I DR 4909-86 "C" Emergency Feedwater Thrust Shoe Replacement DR 3363-86 "A" Emergency Feedwater Pump Repair

DR 5464-86 Pressurizer Relief Value PR-S-14 Overhaul DR 5465-86 Pressurizer Relief Value PR-S-15 Overhaul DR 7094-86 Containment Spray Valve CS-122 and 123 Handwheel Modification Test Report 47927-0 Pressurizer Safety Valve Testing Adequate post maintenance testing was performed after each of the above evolutions. It also appears that adequate corrective action was take Surveillances were reviewed to insure compliance with ANSI and ASME I requirements and that the various acceptance criteria had been met. In l this area one violation was identified. 10 CFR 50, Appendix B, Criterion

! XI states in part, " Test results shall be documented and evaluated to

! assure that test requirements have been satisfied." The licensee's QA program Section II.F requires implementation of 45.2-1977, Quality Assurance Requirements for Nuclear Power Plants. Section 18, of this standard Quality Assurance Records Section, states in part, " Inspection and test records shall, as a minimum, identify the date of inspection or test." Additionally, Section 12, Test Control, states in part, " Test results shall be documented, and evaluated by responsible authority to assure that test requirements have been satisfied."

The inspector reviewed surveillance procedure 3.1.2, ECCS Routine Testing, and found that the various individual surveillances were not dated when performed as required by ANSI 45.2-1971. The procedure includes ECCS valve lineups, pump testing and valve testing required by Technical Specifications and due to the amount of testing required may take in excess of four days to complet The only dates found on the procedure were the initial start date and the individual review dates for each r surveillance. The review dates in some instances were all completed on l the last day within a matter of hour This way of documenting also i

conflicts with ASME XI requirements that state "All test data shall be analyzed within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> after completion of a test."

i The inspector also reviewed surveillance procedure 3.1.5, Emerge'ncy and Auxiliary Feed Pump Test, and found the following discrepancies:

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! Surveillance Procedure 3.1.5, dated March 12, 1986: The ' pump-differential pressure was recorded as 1234.5 psid while the accep-l tance range was between 1068 and 1220 psi ,

I Surveillance Procedure 3.1.5, dated August 6, 1986: The pump 25A

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suction pressure was recorded as 5.2 psig while the acceptance criteria was greater than 7 psig. A retest was conducted later in i the day and a value of 11.5 psig was obtained, i

t Surveillance Procedure 3.1.5, dated December 10, 1986: The pump 258 discharge pressure was originally recorded at 1239 psig while the acceptance range was between 1075 and 1229. The value was changed to 1229 psig without explanation.

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j In subsequent discussions the operations management acknowledged that

all of these items were unacceptable. The problem appea-s to be caused by erratic suction pressure indication which tends to be af fected by initial j oump starting with instrumentation connected. Operations has taken steps to reduce this problem by isolating the suction gauges on pump start. The discrepancies did not appear to have affected pump operability.

l The failure to date the individual surveillance at the time of performance l and the failure to assure that test requirements had been satisfied, are identified as a violation (50-309/87-01-02).

j The overall assessment of the operations and associated support activities i is good. Management appears to be in control of daily activities and make adjustments when required.

i 4.0 Quality Assurance / Quality Control

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4.1 Program Review Selected Quality Assurance (QA) program documents were reviewed to verify that the following administrative controls have been

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established for the QA/QC overview effort.

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Independence, qualification and training of QA/QC personnel

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Documentation and review of corrective actions

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Auditing, including checklist preparation and implementation l

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4.2 Organization and Training i

The. offsite Manager Quality Assurance reports directly to the l Executive Vice President and is assisted by two Project Engineer The Quality Assurance Department (QAD) onsite organization reporting

,~ to the Manager-Quality Assurance (QA) consists of a QA Section Head, three QA Engineers, and a QC Supervisor with a staff of engineers and

! inspectors (all inspectors hold engineering degrees). The QAD staff i

are augmented by Yankee Atomic Electric Company (YAEC) and or con-I tracted personnel as needed. YAEC is responsible for conducting all t audits and most vendor control. An example of additional staffing is the proposal for Combustion Engineering, Inc. (C-E) to supply eleven mechanical and seven electrical inspectors, three auditors and a supervisor during the 1987 refueling outag Training continues to be provided in areas, such as, Design Control (presented by the Plant Engineeria Department), new fuel inspection,

ASME Section XI visual testing, Environmental Qualification (presented by YAEC) and Non Destructive Examination (on-the-job).

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Training is planned during 1987 in nine subjects (e.g. chemistry

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analysis, plant primary and secondary systems, plant trip circuits).

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1 4.3 Implementation i

t All twenty YAEC audits scheduled for 1986 were completed and eighteen YAEC audits are currently scheduled for 1987. Twenty two technical i specialists assisted on more than 10 audits in 1986 and it is planned to use sixteen technical specialists on 12 of the 1987 audits, j

The checklists of the selected audits (see Attachment A) did contain j attributes requiring observation of physical conditions, verification-l of calculation accuracy, acceptability of test results and the

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technical adequacy of work instructions. Documented in audit reports were items, such as, system walkdowns, comparison of as-built documents (e.g. sketches, drawings) to the as -installed items and rechecking of calculation Audit reports also now include an assessment and executive summary section.

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QA Evaluations (i.e. surveillance, in process monitoring) are scheduled on a semi-monthly, bi-monthly, quarterly, semi-annual and i annual basis in 15 functional areas (paralleled to 10 CFR 50,

Appendix B criteria II-XVI) and a miscellaneous category. One 4' hundred and forty four reports were issued for 159 evaluations done in 1986. The inspector reviewed the QA evaluations listed in Attachment A on a sampling basis. The applicable work procedures are almost always used to conduct the evaluations so that this effort is i

mostly observation of ongoing work or physical plant conditions.

3 Examples are checking the position of selected valves to the Safe-

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guards Locked Valve List, examination of warehouse storage conditions l and testing of inservice radiation monitors. There was also evidence b .

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of comparing maintenance procedures to vendor recommendations and an evaluation that identified an instance where out of tolerance results had .been accepted by maintenance personnel (similar to the violation discussed in paragraph 2). -

Inspection hold and notification points are established in the plant procedure A second type of in process monitoring (" Independent Inspection") is also performed. This is broader in scope than normal inspection points so that it approaches the evaluation methodolog The August 8, 1986, semi annual Trend Analysis Report considered nonconformance reports and audit and evaluation reported deficien-cies. The analysis, presented in a pie and bar chart format with accompanying narrative, was brief and precis It identified seven causes of deficiencies, responsible organizations, a 1985 versus 1986 deficiencies comparison, normalized the distribution and provided recommendation .4 Conclusions QA/QC staff levels appear to be adequate as YAEC conducts the audit program and the additional personnel are readily made availabl Personnel are well qualified and motivation is high and there is an increased emghasis on supplementary trainin Audits were comprehensive, the checklists were well developed with obvious technical input. Attributes did include observation of ongoing activities or physical conditions and verification of the accuracy of calculations etc. There was extensive use of technical expertise, corrective action was timely and appeared adequat QA evaluations were almost entirely observation of ongoing work or checks of physical conditions (e.g. valve positions). Inspection points were well chosen and at an adequate level. The concept of

" Independent Inspections" was further emphasis on monitoring in-process work.

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The trending analysis was well done although it did not consider non licensee identified deficiencies such as NRC identified violation Plant management has begun to request certain evaluations and inspections and this indicates the creditability of QA/Q The QA/QC overview effort appears effective and management support was eviden . Unresloved Item Unresolved items are matters about which more information is required to ascertain whether they are acceptable or violations. An unresolved item is discussed in paragraph 2.

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f Management Meetings I

j Licensee management was informed of the scope and purpose of the 1 inspection at the entrance interview on January 26, 1987. The findings of

! the inspection were discussed with licensee representatives during the

! course of the inspection and presented to licensee management at the

January 30, 1987, exit interview (see paragraph I for attendees).

l At no time during the inspection, did the inspectors provide written material to the license The licensee did not indicate that proprietary information was involved within the scope of this inspection.

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ATTACHMENT A Documents Reviewed *

Audit Packages My-86-07, Plant Changes My-86-11, Measuring and Test Equipment My-86-15, Technical Specifications My-86-18, QA Program Evaluations 86E-013, Maintenance Department Surveillances 86E-029, Environmental Qualification 86E-035, Measuring and Test Equipment 86E-101, Material Control 86E-106, Testing of inservice Radiation Monitors 86E-115, Packerizing of Reactor Vessel Spare Studs 86E-128, Final Design Package 86E-131, Safeguards Locked Valve List i

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MAME HARHEE AT0ml0 POWER 00mPARUe avau,,ay,?n" ewe (207) 623-3521

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July 29, 1986 HN-86-100 GDH-86-183 Region I United States Nuclear Regulatory Commission Office of Inspection and Enforcement 631 Park Avenue King of Prussia, Pennsylvania 19406 Attention: Dr. Thomas E. Hurley, Regional Administrator References: (a) License No. DPR-36 (Docket No. 50-309)

(b) USNRC Letter to HYAPCo dated June 26, 1986 - Inspection Report No. 86-07 Subject: Response to Notice of Violation in Inspection Report 86-07 Gentlemen:

This letter is in response to the Notice of Violation contained in Appendix A of Reference (b). The Notice of Violation has been repeated below for completenes Notice of Violation l 10 CFR 50 Appendix B Criterion XII states: "Heasures shall be established to assure that tools, gages, instruments, and other measuring and testing devices used in activities affecting quality are properly controlled, calibrated, and adjusted at specified periods to maintain accuracy within ,

necessary limits". l Item No. 7 of the licensee's controls and limitations for the repair order (DR/R0 #6184-85) for containment spray pump No. P-61A specifies that the final pump shaft to motor shaft total indicator runout readings be correct within .002".

Contrary to the above, as of May 14, 1986, measures were not established to control, calibrate and adjust the dial indicators used to measure, within .002", the shaft runout for the containment spray pump No. P-61 Also these dial indicators did not have any identification or other evidence of being properly controlled, calibrated, and adjusted at specific periods to maintain accurac This is a Severity Level V Violation (Supplement I).

G/ /n.ir 7879L-SDE pgpypyg ,

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MAINE YANKEE ATOMIC POWER COMPANY

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United States Nuclear Regulatory Commission Page Two Attention: Dr. Thomas E. Murley, Regional Administrator MN-86-100

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Maine Yankee Response

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Maine Yankee requests reconsideration, in part, of the Notice of Violation. The dial indicator used to indicate the pump shaft to motor shaft runout indicator readings on P-61A was not used as an absolute measuring device or for a final acceptance criterion during the replacement of the mechanical seal Contrary to the statement in the Notice of Violation, Step 7 of the Controls and Limitations to Discrepancy Report / Repair Order (DR/RO)

No. 6184-85 did not specify "...that the final pump shaft to motor shaft

total indicator runout readings be correct within .002"." Step 7 of DR/R0 No. 6184-85 instructed maintenance personnel to

"After assembly is

complete, record final pump shaft to motor shaft indicator runout ,

reading (Manufacturer recommends to correct within .002" total indicator)...."

This step was included in this DR/R0 package for two reasons:

1). To obtain a before and after maintenance comparison of pump shaft to motor shaft total runout indicator reading ). To obtain diagnostic information on runout of the seal to optimize

the mechanical seal lif The .002" tolerance for total runout indication was a manufacturer's recommendation to minimize seal face wear but it was not an absolute requirement or safety acceptance criterion to be achieved in order to declare the pump operabl Procedure 3.17.6.6 " Inservice Testing of Safeguards Pumps" requires parameters such as pump leakage, vibration, pump amps, and pressure to be j evaluated. The acceptance criterion for each of these parameters must be achieved prior to returning safeguards pumps to service. The testing required by Procedure 3.17.6.6 was performed on P-61A following mechanical j seal replacemen He do agree, however, that dial indicators which are used in safety-related applications should be controlled and calibrated. He have re-examined Maintenance Department practices where dial indicators may be used in safety-related applications. As a result, one procedure has been modified to require the use of a calibrated dial indicato l 7879L-SDE

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4* M AINE YANKEE ATOMIC POWER COMPANY

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l I United States Nuclear Regulatory Commission Page Three l Attention: Dr. Thomas E. Murley, Regional Administrator MN-86-100 l To prevent a recurrence, Maine Yankee will, by October 31, 1986, place dial indicators to be used in safety-related applications under the Measuring and Test Equipment Program which satifies 10 CFR 50, Appendix B, Criterion XI Full compliance will be achieved by October 31, 198 Very truly yours, MAINE YANKEE ATOMIC POWER COMPANY

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bfAV G. D. Whittier, Manager Nuclear Engineering and Licensing

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GDH/bjp cc: Mr. Ashok C. Thadani Mr. Pat Sears Mr. Cornelius F. Holden i

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