IR 05000293/1982029
| ML20028E103 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 12/17/1982 |
| From: | Eichenholz H, Elsasser T, Jerrica Johnson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20028E096 | List: |
| References | |
| 50-293-82-29, NUDOCS 8301200348 | |
| Download: ML20028E103 (17) | |
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50293-820929 50293-821111
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50293-820930 50293-821115 50293-821004 U. S. NUCLEAR REGULATORY COMMISSION 50293-821006
REGION I
50293-821008 50293-821012 50293-821028 50293-821031 Report No.
50-293/82-29 50293-821102 50293-821109 Docket No.
50-293
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License No. DPR-35 Priority Category c
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Licensee:
Boston Edison Company 800 Bo_Ylston Street Boston, Massachusetts 02199 Facility Name:
Pilgrim Nuclear Power Station Inspection At:
Plymouth, Massachusetts Inspection Conducted: October 19, 1982 - November 15, 1982 Inspectors: hk
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1.' Johns'dn, Senior Resident Inspector date dezL~d4 utuin p.Eichenholz,ResidyInspector
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e date Approved by:
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T. ElsasW, Chief, ReMtor Projects date Sectior. No. 18, Projects Branch No. 1 Insp ction Summary:
Inspection on October 19, 1982 - November 15, 1982'(Report No.'50-293/82-29).
Areas Inspected:
Routine unannounced safety inspection of plant operations, including followup of previous findings, an operational safety verification, followup of events, LER's, a review of surveillance and maintenance activities and a review of actions in response to the Perfonnance Improvement Program.
The inspection involved 161 inspector hours by two resident inspectors.
Results:
Five violations were identified in three areas (Failure to properly control equipment tagging / maintenance, Paragraph 6; Failure to properly set a main steam safety valve, Paragraph 4; Failure to control distribution of the Q-List, Paragraph 6; Failure to use proper methods for access control, Paragraph 8; and Failure to properly followup on a security deficiency.
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Paragraph 8).
8301200348 830107 gDRADOCK05000
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Rsgion I Form 12
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(Rev. February 1982)
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DETAILS 1.
Perscns Contacted J. Aboltin, Sr. Reactor Engineer G. Anderson, Watch Engineer N. Brosee, Assistant Chief Maintenance Engineer A. Caputo, Fire Prevention and Protection Engineer W. Deacon, Senior Electrical Engineer R. DeLoach, Group Lea &;r - Operations QC F. Famulari, Acting Q.A. Manager G. Fiedler, Watch Engineer J. Fulton, Senior Licensing Engineer W. Hoey, Senior Radiation Protection Engineer G. James, Reactor Engineer R. Kennedy, Sr. Q.A. Engineer G. LaFond, I&C Supervisor G. Larson, Q.C. Inspector P. Mastrangelo, Chief Operating Engineer C. Mathis, Station Manager J. McCann, Watch Engineer J. McEachern, Security Supervisor A. Morisi, Nuclear Operations Support Manager B. Nolan, Emergen v Planning Coordinator L. Olivier, Wutch Jngineer L. Oxsen, Director of Operations Review K. Roberts, Chief Maintenance Engineer
L. Rucker, Systems & Safety Analysis Group Leader R. Sherry, Sr. Mechanical Engineer P. Smith, Chief Technical Engineer K. Taylor, Watch Engineer G. Whitney, Plant Engineer S. Wollman, Shift Technical Advisor E. Ziemianski, Management Services Group Leader The inspector also interviewed other members of the health physics, operations, maintenance, security, and technical staffs.
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2.
Followup on Previous Findings A.
(Closed)FollowItem(50-293/82-15-01).
Changes to Special Nuclear Material inventory and control procedure. The inspector verified that procedure No. 4.0, SNM Inventory and Control, Revision 7, was approved by the ORC on September 29, 1982 and addressed the coments discussed in Report No. 50-293/82-15. This item is closed.
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B.
(0 pen) Unresolved Item (50-293/81-08-04).
Control Rod Drive Hydraulic Control Unit Accumulators.
The inspector reviewed the licensee's procedures (alarm, system operating 2.2.87, and abnormal condition 2.4.8) which address CRD HCU accumulators.
The inspector expressed concern that the present procedural guidance was unclear as to the following:
1) conditions constituting an inoperable accumulator, 2) the relationship between an inoperable accumulator and an inoper-able control rod, and 3) the requirements for reporting various degraded conditions. The licensee has initiated a review of the appropriate safety analysis report and Technical Specification bases and has been in contact with the NSSS to resolve these concerns. The licensee stated that written guidance would be issued in the near future to resolve these concerns.
Pending a review of the licensee's written guidance to resolve the above concerns, this item remains open.
3.
Operational Safety Verification A.
Scope and Acceptance Criteria The inspector observed control room operations, reviewed selected logs and records, and held discussions with control room operators.
The inspector reviewed the operability (including valve positions)
of the High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems.
Tours of the reactor building, turbine building, station yard, switchgear rooms, SAS, diesel generator rooms and the control room (daily) were conducted.
The inspector's observations included a review of equipment conditions, control room annunciators, potential fire hazards, physical security, housekeeping, radiological controls, equipment control (tagging) and radioactive release rates from the station.
The inspector reviewed records of radioactive liquid and gaseous
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release from the station and sampling of the Standby Liquid Control System boron concentration.
i These reviews were performed in order to verify conformance with the facility Technical Specifications and the licensee's procedures.
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B.
Findings (1) On October 20, 1982 the inspector noted that the licensee had been assigning personnel who were not the on-watch Watch Engineer to the
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position of Fire Brigade Leader. This appeared to be contrary to Figure 6.2.3 of the facility Technical Specifications. However, the NRC's current guidance in this area, 10 CFR 50, Appendix R, recognizes the potential needs of the Watch Engineer to direct activities from other than the scene of the fire, namely, the control room, and re-comends that the " shift supervisor" not be assigned the concurrent position of Fire Brigade Leader.
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The licensee issued LER No. 82-03/99X-2 on October 21, 1982 describing recent changes in station organization including a change in the Fire Brigade Leader designation, and has included this in a proposed change to the Technical Specifications.
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to the Fire Brigade Leader's position had been properly trained, and had no further questions.
No violations were identified.
(2) The inspector reviewed activities involved in a plant shutdown on November 2,1982 and the subsequent startup on November 4,1982.
The licensee identified a failure to follow the startup procedure No. 2.1.1, Startup from Cold Shutdown, Rev. 38 dated September 7,1982.
In parti-cular, Step 9.g.1 required a drywell personnel airlock door seal test prior to exceeding 2120F. The test had been completed satisfactorily at about 6:05 am on November 4,1982 after the reactor had been taken critical at 5:33 am and 2170F.
Primary containment integrity was not violated. The facility Techni-cal Specification 4.7.A.c.(2) requires testing the door seals once eai:h operating cycle. However, 10 CFR 50, Appendix J, Section III, requires testing air locks after each opening.
Station procedure No. 8.7.1.7, Leak Rate Testing of Personnel Air Lock, Revision 8, describes the previous NRC position on this Appendix J requirement as requiring a door seal test within 3 days of each opening for multiple entries,
and an air lock test with the mechanical strongback installed at 6 month intervals.
The inspector noted that the door seals had been tested during the plant depressurization on November 2,1982 and that the test at 6:05 am on November 4,1982 during the startup was within the 3 day requirement described above.
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The inspector concluded that this item was a licensee identified violation of station procedure No. 2.1.1.
The licensee promptly filed Failure-Malfunction Report No.82-216. describing this event. No other
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violations were identified.
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(3)
During various times in this period the inspector reviewed compensatory measures being taken for equipment problems in the security area. No violations were identified.
Details surrounding access to the Reactor Building Trucklock doors on November 3, 1982 are described separately in Paragraph 8, below.
(4) On November 9,1982 the inspector noted that the 4' Traversing Incore Probe (TIP) machine control panels were red tagged in the control room.
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The inspector further noted that the tags were issued on June 26, 1982 for an entry into the ventilation system in accordance with procedure
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No. 1.4.5.
However, the inspector also noted that the TIP machines had been subsequently operated and therefore reviewed the Watch Engineer's (W.E.) Red Tag log index. The inspector determined that the 4 TIP machines had been tagged out about 10 times since June 26, 1982, each entry having been authorized and cleared in the log. The inspector noted that there was no entry indicating approval of the 4 tags currently hanging on the control room control panel.
This is a violation of station procedure No. 1.4.5.
The on-watch Supervisor immediately corrected this error. This example of a violation involving adminstrative and opera-tional control of tagging is included with further details in Paragraph 6 below.
4.
Followup on Events; Trips, and Licensee Event Reports (LER's)
A.
Review of LER's
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(1)' LER's submitted to the NRC:RI were reviewed to verify that the details were clearly reported and that the corrective actions were adequate. The inspector determined whether generic implications were involved and whether onsite followup was warranted.
The following LER's were reviewed:
LER No.
Subject 82-03/99X-2
, Station Organization Change 82-41 Reactor Building Damper Inoperable 82-42 HPCI M0V 2301-35 Inoperable 82-43 RCIC MOV 1301-25 Inoperable 82-44 Control Rod Accumulator No. 22-51 Inoperable 82-45 High RCS Conductivity
- 82-46 HPCI MOV 2301-3 Inoperable 82-48 Inoperable Fire Door
- 82-49 Incorrect Main Steam Safety Valve Setting 82-51 Loss of Offsite Power (FollowupofLER's 82-42, 82-43, 82-45, and 82-51 are documented in NRC Report No. 50-293/82-24).
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(2)
For the LER's selected for onsite review (denoted by asterisks), the inspector verified that appropriate corrective actions were takea or responsibilities assigned and that continued operation of the facility was conducted in accordance with the T.S.
82-46; High Pressure Coolant Injection (HPCI) System Steam
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Supply Valve (MOV 2301-3) inoperable. On October 28, 1982,
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a ground developed in the 'B' 125 v. dc battery bus. Trouble-shooting identified the grounded component as the internal limit
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switch circuitry in the MOV 2301-3 valve operator.
(This event
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isalsodescribedin4.Bbelow) The cause of the ground was moisture from a packing leak on the valve. The HPCI system was
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declared inoperable, isolated for repairs, the valve stem repacked, and the system returned to service. The inspectors verified that conditions specified in the Technical Specifications (T.S.) were
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met.
This LER also describes a similar ground that developed on the 250 v. dc battery bus for the same valve on November 1,1982.
The licensee tightened the packing to limit the amount of steam
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blowing. Again, the requirements of the T.S. were met.
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The inspectors verified that the requirements of the T.S. were
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met for these two events. However, they noted that this HPCI
steam supply valve stem packing had been leaking for many months i
prior to these grounds developing and that steam leaks were the cause of two more instances of inoperability gn November 10, 1982.
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The licensee had previously recognized the concerns from this leaking valve and had a new valve stem on order.
Following a lengthy procurement process the station has been provided with
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j a spare stem and plans to perfom repairs in the near future.
Other comments are provided in Paragraph 4.B below.
82-49; Incorrect Main Steam Safety Valve Setting.
On Novemb'r 2,
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e 1982, following an insurance company audit, the licensee determined that the installed 'A' main steam safety valve had been previously set to lift with nitrogen at the steam setting of 1240 psig (vice the nitrogen setting of1185 psig). The NRC was immediately
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notified, the plant was shutdown, the installed valve was re-i l
placed with a properly set valve and the plant was subsequently l
returned to power.
Further description of this event is described l
in Paragraph 4.B below.
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B.
Events / Plant Trips (1)
' A' Recirculation Pump Trip
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On October 28, 1982, at 4:17 pm, plant personnel were trouble shooting a ground on the 125 v. dc "B" Battery System in accordance with Maintenance Request (MR) 82-1544. As a result of this activity, the
"A" Recirculation Pump was inadvertently tripped when 125 v. de control
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?ower was temporarily removed from the Recirculation Pump MG Set #1 Auxilliary 011 Pump (P228A) at 250 v. dc Motor Control Center (MCC)D9.
As a result of this event, reactor power rapidly decreased from full power to approximately 65 percent.
Directly following the event, the inspector observed the use of tile procedures listed below:
2.4.17; Trip of one recirculation pump
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2.1.9; Reactor recirculation pump operation, Section B
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2.1.15; Test No. 32, recirculation pump speed and jet pump operability
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2.2.84; Recirculation system
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The inspector observed no unusual conditions in plant pararreters, and other equipment appeared to be functioning properly as evidenced by a review of the various control room indicators and recorders. At 5:04 pm, following the detemination as to the cause of the trip and verification of allowable recirculation loop temperatures (4840F idle
loop vs. 5000F operating loop), the "A" recirculation pump was placed in service so that the plant could return to nomal operations.
In terms of equipment perfomance or operator actions taken sub-l sequent to the "A" recirculation pump trip, no inadequacies were identified by the inspector.
The inspector reviewed plant documentation and held discussions with l
various licensee personnel to ascertain the facts associated with l
the inadvertent pump trip. Based upon initial infomation provided
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by the licensee and inspector review of the licensee's activities, the event followup would require additional review in the following areas:
Maintenance troubleshooting activities and problems encountered,
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and Conditions leading to developing a ground on the "B" 125 v. dc
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station battery.
In the area of maintenance troubleshooting, it appeared that initial efforts to track the ground through the D.C. buses were proving successful until the 125 v. de control power was removed at Compart-ment 931 on 250 v. dc MCC D9 (Aux. pump P228A), which resulted in
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the recirculation pump trip. The maintenance personnel were unaware that this would occur due to the belief that removing the 125 v. dc control power at a prior distribution point resulted in removal of the
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ground indication on the battery with no adverse affects on plant
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operation.
Preliminary investigation into the inadvertent trip of
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the recirculation pump by the licensee determined that problems may have existed with plant documentation reflecting the as-built con-
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figuration of equipment,
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Following the inadvertent recirculation pump trip, the licensee was l
successful in determining the cause of the 125 v. dc "B" battery
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ground, which was attributable to moisture accumulating inside the High Pressure Coolant Injection (HPCI) system steam supply valve's (MOV 2301-3) motor operator. At 11:30 pm the licensee declared the HPCI system inoperable ar.d issued M.R. 82-1545 to remove the moisture from inside the valve motor operator. The HPCI system inoperability and the
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cause of the moisture are discussed in Paragraph 4.A above as part of LER 82-46.
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i The inspector discussed several questions with-the licensee's manage-ment concerning troubleshooting activities, configuration control, and equipment qdalification. ThE Station Manager stated that a report
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would be prepared within the next few weeks that would provide addition-al details. The inspector will review the licensee's report concerning this event in a future routine inspection.
(50-293/82-29- 01 ).
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(2) On October 31,1982, at 7:30 am, a half scram signal was received
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on Channel 'A' Reactor Protective System from high APRM channels A, C, and E.
Investigation revealed a faulty recirculation loop flow signal from a proportional amplifier which feeds the APRM flow converter network. The faulty unit was replaced, testing was performed, and the system returned to normal at 10:16 am on
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October 31, 1982.
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The inspector reviewed this event and determined that the NRC was properly notified and that the requirements of the T.S. for APRM scram capability were met.
Coments concerning the quality of l
repairs / testing are described in Paragraph 6 below, i
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(3) On November 2,1982, at about 11:10 a.m., the licensee notified the NRC that a plant shutdown from full power was in progress because of an in-correct Main Steam Safety Valve setting for one of the tw installed valves.
The inspectors held discussions with licensee personnel, reviewed sta-tion documents, and monitored the licensee's corrective actions. The licensee determined that the cause of the incorrect setting was per-sonnel error compounded by procedural inconsistency. This is described in LER No. 82-49.
The licensee stated that during a recent insurance company audit of the main steam system, the auditor questioned the basis for the setpoint of the spare safety valve (No. 6262) currently being tested.
The licensee's staff engineer identified an erroneous nitrogen setting (which had not yet received final licensee staff review) and immediately reviewed rec-ords of the installed valves. The licensee determined that the 'A' safety valve (No. 6302) had been set on February 15, 1982 with nitrogen at the steam setting of 1240 psig vice the nitrogen setting (1170-1185 psig).
TS Limiting Safety System Setting 2.2 requires the safety valves to be set to lift at 1240 psig +_13 psi.
If an overpressure iondition had occurred in the main steam system, the incorrectly set Mr.;n Steam Safety Valve would not have lifted at 1240 psig but rather at some higher pres-sure.
LER No. 82-49 states that this valve will be sent to a testing laboratory to determine the as found setting, and that a complete re-vision to prc:edure No. 3.M.4-7 will be made to clarify the intended test method.
The plant was shutdown on November 2, 1982.
Persor.nel were counselled on procedural adherence and an innediate change to procedure No. 3.M.4-7, Main Steam Safety Valve, was made adding cautions referencing original nitrogen test pressures stamped on the nameplate.
Re-setting of the l
spare safety valve (No. 6262) was performed on November 2,1982 and wit-
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nessed by the inspector. A review of documentation for the installed
'B' safety valve (No. 6309) verified that it had been set correctly.
The improperly set 'A' valve (No. 6302) was replaced with the spare (No. 6262) and a plant startup was performed on November 4, 1982.
The failure to follow procedure No. 3.M.4-7 in setting the safety valve (No. 6302) with nitrogen on February 15, 1982 to ensure the setting was in accordance with TS 2.2 is considered a violation (50-293/82-29-02).
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(4)' On November 2,1982, at about 4:56 pm, the licensee received noti-fication from the local police that a bomb threat had been received by the Boston Globe newspaper earlier in the day. The caller to
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the newspaper indicated that there were several bombs placed in the Boston area. The inspector verified that the licensee took the appropriate actions, that the station was secure.
Further followup
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was perfomed by state and federal agencies.
(5) On November 4, 1982, at 2:10 pm, the licensee suspended plant start-
up to repair a minor steam leak in the drywell. This was reported to the NRC, repaired, and nomal startup procedures continued at 7:00 pm. No inadequacies were identified.
(6) On November 9,1982, at 11:09 am, the licensee took the ' A' Conden-i sate Demineralizer out of service following the observation of in-
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creasing reactor coolant system conductivity. The licensee's actions
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prevented the conductivity from exceeding the T.S. limit.
Investigation revealed that several lateral screens were damaged and that future
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changes to the size of several laterals would need to be performed.
The licensee has implemented administrative controls to more closely monitor and control the 'A' Condensate Demineralizer during its future use, which is planned to be restrictive in nature and only during the backwashing of other demineralizers. No violations were
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identified.
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(7) On November 11,1982, at about 10:40 am, the licensee identified a
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failure of relay 5AK-3G to deenergize during routine testing of the
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'D' inboard Main Steam Isolation Valve position switches. Power was reduced to about 60%, the inboard and outboard MSIV's on the
'D' line were closed, the licansee notified the NRC, and the station management reviewed the event. The station On-site Review Committee (ORC) approved a temporary modification (TM 82-60) to remove the power supply fuse to relay 5AK-3G to fail the relay in the ' tripped mode'. The two MSIV's were then reopened and nomal power operations resumed. The licensee plans to investigate the 'D' inboard MSIV position switches at a future drywell entry.
The inspector discussed this event with. licensee personnel, observed the deenergized relay / fuse status and had no further questions at this time. No violations were identified.
(8) On November 15, 1982, at about 6:02 am, while performing alternative testing of redundant components prior to a planned maintenance outage on the HPCI turbine steam supply valve, the Reactor Core Isolation Cooling (RCIC) turbine tripped and a system isolation was received.
Investigation revealed no steam leaks. The system was declared in-operable and the NRC was notified; The inspector verified that
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j alternative testing of the HPCI system was perfomed as required l
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by the Technical Specifications. At the end of this inspection period the RCIC system was still inoperable.
Further followup of actions to restore the RCIC system to an operable status will be performed during routine inspections of the facility. No viola-tions were identified.
5.
Surveillance Activities The inspector reviewed the licensees actions associated with surveillance testing to verify that the testing was performed in accordance with approved station procedures and the facility Technical Specifications.
Portions of the following tests were reviewed / observed:
alternative testing for an inoperable HPCI system between October 28-29,
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1982 and post maintenance operability testing.
setting of the spare Main Steam Safety Valve on November 2, 1982
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routine load test of an EDG on November 4,1982
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routine I&C functional testing of RPS/PCIS inputs from high reactor l
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water level signals on November 8,1982 alternative testing and post maintenance testing of the HPCI MOV 2301-3
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following inoperability declaration at 2:45 am on November 10, 1982 alternative testing and post maintenance (packing adjustment) testing
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of the HPCI MOV 2301-3 at 2:26 pm on November 10, 1982, and alternative testing of the HPCI system following the identification of
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an inoperable RCIC system on November 15, 1982.
No violations were identified.
6.
Maintenance Activities and Operational Control of Tagging A.
The inspector reviewed the licensee's actions associated with maintenance activities in order to verify that they were conducted in accordance with station procedures and the facility Technical Specifications. The inspec-tor verified for selected items that the activity was properly authorized
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and that the appropriate radiological controls, equipment control tagging,
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j and fire protection were being implemented.
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The Boston Edison Company (BECo.) Operations QA Manaual, Section 2.3, dated June 15,1981, states that the Q-List is a list of safety related systems, structures, and components maintained under the authority of the Nuclear Engineering Manager and issued in a controlled manner by the Nuclear Operations Support Section.
On October 8, 1982 the Nuclear Engineering Department Manager issued a revision to the Q-Li+,t (NED 82-661). This issue was not distributed in a controlled manner.
The inspector discussed the control of the October 8,1982 version of the Q-List with the Operations QC Group Leader, the acting QA Manager,
and the Nuclear Operations Support Department Manager. The licensee representatives acknowledged the failure to properly distribute this
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issue and stated that a newer version had already been sent to controlled
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copy holders on November 2, 1982. The licensee further stated that a review would be perfomed to verify the quality of activities performed between October 8, 1982 and November 4, 1982.
The failure to distribute the October 8,1982 version of the Q-List in a controlled manner is a violation (50-293/82-29-03).
(3) On November 8,1982, the inspector r' viewed an active M.R. (82-1555)
which documented the tagging /isc' Ao af the jockey fire pump for repairs on November 5,1982. L N time of this review, the section of the M.R. required to be signci h;/ ine on-watch Watch Engineer (authorizing pemission to start the isolation) was blank. This dis-crepancy was imediately corrected by the licensee.
The failure to properly authorize the isolation is a violation of the station procedures controlling maintenance (No. 1.5.3,1.5.7).
l (4) On November 10, 1982, at about 2:26 pm, the licensee declared the HPCI system inoperable to repair a steam leak on the turbine steam supply
valve (MOV2301-3). The inspector questioned the on-watch Watch Engineer concerning the administrative controls used during these activities.
The Watch Engineer stated that no Maintenance Request was used, that the packing was tightened, and that a Watch Engineer's red tag was used to control the isolation, although no entry had been made in the tag log.
The inspector verified that a post work operability test of the valve had been perfomed but expressed concern to the Chief Operating Engineer of the apparant lack of proper administrative controls for this activity.
The failure to log the issuance of a Watch Engineer's red tag is a viola-tion of the station procedure for tagging (1.4.5).
The violations described in Paragraphs 6.B.(1), 6.B.(3), 6.B.(4), as well as Paragraph 3.B.(4) above, concerning operational control of tagging /
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maintenance activities are collectively considered a violation. (50-293/82-29-04).
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The items / documents revierd included the following:
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Maintenance Request (M.R.) 82-1544; Ground on 'B' 125 v. de bus
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M.R. 82-1545; HPCI valve 2301-3 steam leak causing ground
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M.R. 82-1547A; Repair 'A' 125 v. de ground detection teter
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M.R. 82-1547; Repair APRM flow control circuit proportional
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amp 1ifier 260-8A M.R. 82-1-56; Reset spare safety valve No. 6262
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M.R. 82-1-57; Replace installed
'A' safety valve with spare
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M.R. 82-1555; Repair jockey fire pump
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M.R. 82-1549; HPCI valve 2301-3 steam leak causing ground
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M.R. 82-1560; HPCI valve 2301-3 steam leak causing ground, and
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M.R. 82-1562; Install temporary modification No. 82-60 to
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remove fuse powering relay 5AK-3G.
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Findings
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(1) On October 29, 1982, the inspector reviewed the two M.R.'s (82-1544, and 82-1545) in progress to correct the ground and steam leak on HPCI 2301-3 valve.
Neither M.R. referenced a red tag deenergizing the 125 v. de control power (125 v. de control power knife blade in breaker panel C944).
The inspector veri-fied by observation that the control power was deenergized and red tagged for worker protection.
The on-watch operators immediately corrected the M.R. by adding the appropriate red tag
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reference.
The failure to reference this component as requiring isolation prior to starting work is a violation of station procedures for tagging and maintenance (No. 1.4.5, and 1.5.3).
(2) On November 1, 1982 the inspector reviewed the calibration and maintenance activities performed on October 31, 1982 during the replacement of proportional amplifier 260-8A (M.R. 82-1547).
The inspector reviewed the calibration records for test equip-ment used, the calibration procedure used, and the proportional amplifier material qualifications.
The inspector noted that the amplifier was installed in the APRM cabinet panel No. 937.
The BEco. list of safety related systems and components (Q-List)
dated October 8, 1982, section 26.1112 (Page 2 of 61) lists panel No. 937 as a safety related structure. The inspector noted that the replacement amplifier module, although cali-brated, did not have a material receipt inspection report (MRIR)
number which could be used to trace its quality.
Following discussions with the on-site QC Group Leader on November 3, 1982, a station nonconformance report and deficiency report were issued to resolve questions pertaining to this replacement. On November 3, 1982, the amplifier was replaced under M.R. No. 82-1552 with a component whose quality was traceable.
The inspector also noted that the licensee was in the progress of addressing spare parts storage and handling deficiencies in general and had plans to return other question-able spare parts to the warehouse to properly control their iuture use.
The inspector determined that the licensee had promptly resolved the questions concerning the amplifier, however, a violation concerning the distribution of the Q-list was identified.
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Response to Order for Modification of Licensee / Performance Improvement Program Boston Edison Company has responded to the NRC's January 18, 1932 Order by submitting a revised Performance Improvement Program (PIP), Revision 1, on July 29,1982 (BECo. letter No.82-203). The NRC: Region I's tentative accept-ance of this program is described in a letter from the NRC to BECo dated September 27, 1982.
The inspector met with the licensee on November 3, 1982 to review the status of selected milestones planned for completion in October,1982. These mile-stones are enumerated below and utilize the identification numbers described in the PIP, Rev. 1.
At this meeting the inspector was not presented with any objective evidence to indicate completion of items III.3.G.3 (Corrective Maintenance) and IV.2.4.6 (QA Program Traininc).
Furthermore, the documenta-tion presented as evidence of satisfactory completion for items II.8.A.3 Assessment)ganization Policias) and III.2.B.3.2 (Operational Experience,
(NuclearOr did not provide sufficient information to allow the inspector to consider the items completed.
Due to these conditions, the inspector and the licensee's representative met again on November 15, 1982 and resolved two of the Tour items.
II.3.A.5 Licensing Activities; The licensee h.as consolidated into the Nuclear Operations Support Department all nuclear licensing
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activities. The licensee has specified in Nuclear Operations Support procedure No. 1.01 " Organization", the functions respon-
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sibilities, and interfaces for licensing activities.
II.8.A.3 Nuclear Organization Policies; The licensee committed to develop a list of additional NUORG policy directives desired by management. The inspector was given a proposed Table of Contents for the Nuclear Organization Policy Manual.
From the l-documentation presented, the inspector was unable to determine j
if the milestone commitment was met. The inspector informed l
the licensee's representative that this item would be reviewed with NRC:RI management and discussed at the periodic PIP im-
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l plementation status meeting to be held on November 17, 1982.
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III.l.D Correspondence Review; The licensee has completed the corres-pondence review with respect to I&E Bulletins. This has re-sulted in the establishment of March 1, 1983 as the completion date for reviewing an appropriate sample of I&E Bulletins to reconfirm compliance.
I III.2.8.3 Operating Experience Assessment (OEA); The Plant Operating Experience Assessment Canmittee (P0AEC) has submitted its report with recommendations that provide significant improve-ments over the prior program.
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III.3.A Conficuration Management; The licensee has prepared a draft report that defines a conceptual scope including regulatory and BECo management requirements, program elements and inter-faces.
III.3.A.2 Equipment Lists; The licensee has determined the scheduled completion daid for revising and maintaining equipment lists (mechanical, electrical,andinstruments),Qlist,andFSARas July 29, 1983.
III. 3. P,. 3 Design Change Control Process; The licensee has prepared and issued Nuclear Engineering Department Procedure 3.02, Rev. 8, Preparation, Verificatior., Approval and Revision of Design Documents for Plant Design Changes / Modifications.
This pro-cedure includes design change criteria development to identify specified requirements for a system design change.
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III.3.G.3 Corrective Maintenance; The licensee has revised procedure l.5.3, Maintenance Requests, to improve the quality of correc-tive maintenance and fulfill the requirement of the PIP committment IV.2.4.6 QA Program Training; This PIP milestone required the Training Department to provide training to all BECo Nuclear Organization and contractor personnel in the BEco QA Program. The licensee has reported that the extensive training program has achieved
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approximately a 90 and 95 percent completion for BECo and Contractor personnel, respectively. The inspector informed the licensee's representative.that this item would be reviewed
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with NRC:RI management and discussed at the periodic PIP
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implementation status meeting to be held on November 17, 1982.
IV.2.4.7 Review QA Audit Program; The licensee has reviewed the QA Audit Program to detennine program improvements to increase effectiveness. This action was documented in a report trans-mitted to the Senior Vice President on August 3, 1982.
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Unresolved Items Areas for which more infonnatica is required to determine acceptablility are considered unresolved. Unresolved items are discussed in Paragraph 2,
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Exit Interview At periodic intervals during the course of the inspection, meetings were held
with senior facility management to discuss the inspection scope and findings.
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