ML20236F609
| ML20236F609 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 10/23/1987 |
| From: | Elsasser T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20236F591 | List: |
| References | |
| RTR-NUREG-0660, RTR-NUREG-660, TASK-1.C.1, TASK-2.B.2, TASK-TM 50-029-87-06, 50-29-87-6, GL-82-33, IEB-84-03, IEB-84-3, NUDOCS 8711020234 | |
| Download: ML20236F609 (36) | |
See also: IR 05000029/1987006
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-29/87-06
Docket No.
50-29
Licensee No.
Licensee:
Yankee Atomic Electric Company
1671 Worcester Road
Framingham, Massachusetts 01701
Facility Name:
Yankee Nuclear Power Station
Inspection at:
Rowe, Massachusetts
Inspection Conducted: April 27,1987 - July 20,1987
Inspectors:
H. Eichenholz, Senior Resident Inspector
D.-Haverkamp, Project Engineer
C. Carpente , Reactor Engineer
Approved By:
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T. El
Chief, Reactor Projects Section 3C
Date
Inspection Summary: Inspection on April 27, 1987 - July 20, 1987 (Report No.
50-29/87-06
Areas Inspected: Routine onsite regular and backshif t inspection by resident
inspectors (418 hours0.00484 days <br />0.116 hours <br />6.911376e-4 weeks <br />1.59049e-4 months <br />). Areas inspected included licensee action on previous
findings, operational safety verification, radiological controls, events
requiring telephone notification to the NRC, plant events, maintenance
observations, surveillance observations, onsite review committee activities,
licensee response to IE Bulletins, licensee corrective actions on fitness for
duty concerns, licensee action on NUREG 0660, changes to the licensee organi-
zational structure and an onsite meeting to discuss operator licensing issues.
Results:
No violations were identified by the inspector; however, three
inadequacies involving the untimely notification of a 50.72 reportable event
(Section 6), the failure to initiate a maintenance request for safety-related
maintenance on an emergency diesel generator (Section 8), and the incorrect
implementation of a containment boundary modification (Section 9) were classi-
fied as licensee-identified violations.
8711020234 071023
ADOCK 05000029
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Details
1.
Persons Contacted
Yankoe Nuclear Power Station
B. Drawbridge, Assistant Plant Superintendent
T. Henderson, Technical Director
N. St. Laurent, Plant Superintendent
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The inspector also interviewed other licensee employees during the
inspection, including members-of the operations, radiation protection,
chemistry, instrument and control, maintenance, reactor engineering,
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security, training, technical services and general office staffs.
2.
Summary of Facility Activities
At the start of the inspection period on April 27, 1987, the plant
was continuing normal coastdown operations from Cycle XVIII.
From
approximately 72% of rated power on May 2, 1987, a plant shutdown was
initiated for the Cycle XVIII-XIX refueling outage.
Cold Shutdown (Mode
5) was achieved on May 4, 1987, and the plant was in the Refueling Mode
(Mode 6) on May 12, 1987.
The licensee's planned six-week refueling
outage stretched into a nine-week outage principally due to: an overly-
optimistic schedule; discovery of a severely damaged fuel assembly that
required fuel reconstitution, modification to core components and
implementation of additional radiation protection measures to reflect
hot particle concerns; the expansion of steam generator eddy-current
inspection and tube plugging activity in response to a re-evaluation of
1984 inspection data; and repair activities for a body-to-bonnet leak on
the No. 2 main coolant cold leg stop valve, which also resulted in valve
disc replacement due to identified disc cracking.
On July 1,1987, the licensee initiated Core XIX Physics Testing.
Initial
criticality was achieved with testing being satisfactorily completed on
July 2, 1987.
The turbine was phased to the grid on July 3,1987. While
performing main turbine overspeed testing on July 6, 1987 with the plant
in the startup mode (Mode 2), a reactor scram on main coolant high pres-
sure occurred due to personnel error.
Successful completion of the
turbine testing and phasing to the grid occurred on July 7, 1987.
Power
escalation was halted and a plant load reduction to 50% of rated power
occurred on July 9,1987 to allow the conduct of condenser tube leak
checks. The plant achieved 100% of rated power on July 11, 1987 and
remained essentially at that power level through the remainder of the
inspection period.
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3.
Licensee Action on Previous Inspection Findings
(Closed) Inspector Follow Items 83-EP-01, EP program assessment to
review NUREG 0737 Item II.B.2, Post Accident Site Access Assumptions;
and 83-03-01, Review licensee evaluation of individuals coming onsite
with postulated high radiation conditions in vapor container and deter-
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mine if consistent with emergency plan / staffing guidelines.
The inspector reviewed the following documents: Analysis / Calculation for
YR-LOCA Evacuation Dose dated August 10, 1979; Shielding Design Review
letter dated July 6,1981; Shielding letter dated November 30, 1981;
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Status of TMI Action Plan Item Implementation letter dated May 14, 1982;
and.TMI Action Plan Item II.B.2.2, Design review of plant shielding -
corrective actions for access to vital areas dated April 11, 1983.
The inspector noted that the licensee has implemented a Post-Incident
Cooling (PIC) System which permits the safe shutdown, cooldown and
maintenance of cold shutdown conditions from within shielded areas of
the plant. The implementation of the PIC, emergency procedures and
personnel actions have been reviewed against the guidelines of NUREG 0737, Item II.B.2 and.found to be acceptable.
The inspector also reviewed personnel dose rate criteria for site access
during period of high Vapor Container (VC) shine. Administrative and
emergency procedures are in place to assure unfettered access to the
plant during emergencies. Assumptions include transit to the plant in
a vehicle traveling at 40 MPH and travel starting at the Forward Control
Point (Furlon House).
The inspector traveled the access road at various
speeds (40, 30, 20, and 10 MPH) and noted that 40 MPH travel can safely
be performed in dry weather conditions.
Travel at 20 MPH should be
possible under adverse weather conditions (snow).
Dose rate criteria
can be met at travel speeds as low as 20 MPH with maximum calculated VC
shine. As the VC shine decreases, a function of time after shutdown,
dose rate criteria can be met at progressively lower travel speeds.
Based upon the above review, these items are acceptable.
(Closed) Unresolved Item (50-29/87-04-01) Lack of formality and profes-
sionalism in the control room.
This item reflected NRC observations of
activities involving shift relief practices and inadequate control of
personnel in the control room who enter the immediate area surrounding
the operators and the control boards.
The first item was corrected by
counseling the individual involved to insure that his future actions or
statements do not appear to reflect a shift relief action when none was
intended.
Subsequent observations on a routine basis of shift relief
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practices and protocol by the inspector have not detected recurrence of
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NRC concerns in this area.
Regarding control room decorum, licensee
corrective actions have included: issuance of Special Order (S.O.) 87-77
by the plant operations manager that provides a memorandum to the plant
operators on expected control room conduct; submittal of this memorandum
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to all department heads with a request that they discuss the expectations
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for proper control room decorum with their respective staffs; discussion
of the. issue at a plant operations review committee meeting; and revision
to the general employee training program to include management policy on
personnel conduct in the control room. Additionally, the plant operations
manager has-informed the inspector that the management policy and ' guidance
provided in S.O. 87-77 will be included in an operations department
procedure.
Based upon observed improvemer,ts in control room decorum that reflect a
professional conduct of operations, this item is considered closed.
(Closed) Unresolved Item (50-29/87-04-04) Change Technical Specifications
to conform to 10 CFR 50.54 minimum shift manning requirements. On June
10, 1987, the licensee submitted to NRC:NRR Proposed Change (PC) No. 203
to the Technical Specifications (TS) that will modify Table 6.2-1 to be
consistent with 10 CFR 50.54.
This item is closed.
-(Closed) Unresolved Item (50-29/87-04-03) Review plant procedures to
ensure conformance with 10 CFR 50.54(m).
The licensee issued revision 5
to procedure AP-2009, Control Room Area Limits for Control Room Operators
and Senior Licensed Operators.
The current revision no longer contains
an allowance for the supervisory control room operator to respond to a
plant fire or alarm, when he is the fire brigade leader without a shift
supervisor in the control room.
The procedure is now in conformance with
10 CFR 50.54(m)(2)(iii) which requires a person holding an SRO license to
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be in the control room at all times.
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4.
Operational Safety Verification Reviews
a.
Daily; Inspection
During routine facility tours, the inspector checked the following
items: shift manning, access control, adherence to procedures and
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limiting conditions for operations (LCOs), instrumentation, recorder
traces, protective systems, control rod positions, containment
temperature and pressure, control room annunciators, radiation
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monitors, radiation monitoring, emergency power source operability,
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control room and shift supervisor log, tagout log and operating
orders.
No inadequacies were identified except as noted in certain
sections below.
1.
Prior to plant shutdown for the refueling outage, plant
operators opened, one at a time, the loop bypass valves.
This
action occurred between April 28-30, 1987 and was done for the
purpose of flushing the bypass lines.
During the prior SALP
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period (SALP Report 50-29/85-98), the NRC specified the need
for the licensee to insure that procedures are developed for
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all planned operations.
The inspector noted that procedure
OP-4721, Rev. O, Operation with Loop Bypass Valves Open, w e
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issued in January, 1987.
This was an indication that the
licensee has demonstrated responsiveness to NRC recommendations
for improved operations.
The inspector verified that the open-
ing of the bypass valves was in accordance with the established
procedure, which upon implementation insures that the licensee
is within its safety analysis envelope and technical specifica-
tions requirements.
2.
During the prior refueling outage the inspector expressed
concern relative to the licensee's planned use of demineralized
water in the shield tank cavity (STC) for the purpose of rinsing
components.
These concerns were documented in Inspection Report
50-29/85-18.
On May 25, 1987 the assistant technical director
submitted a safety evaluation to the PORC that describes the
consequences of a controlled addition of demineralized water
to the STC for the purpose of equipment decontamination. This
safety evaluation was generated to support procedure OP-1901,
Rev. O, Controlling the Addition of Demineralized Water to STC
Water for Equipment Decontamination.
The inspector identified
no deficiencies as a result of reviewing the licensee's activ-
ities in this area.
3.
The inspectors reviewed a sampling of the licensee's refueling
procedures and observed refueling operations in progress in the
vapor container (VC), control room and spent fuel pool (SFP)
building. The major activities observed by the inspector
were:
1) removal of Cycle 2 fuel from the core, placement in
the upender, transfer to the SFP and placement in the spent
fuel racks; 2) removal of cycle 1 fuel, ultrasonic inspection
and installation into cycle 2 fuel position; and 3) movement of
a new fuel assembly into the SFP and subsequent tran'sfer to the
VC. Observations and discussions with personnel in the VC and
SFP indicated that personnel were knowledgeable of their job
and requirements and work progressed in an orderly, profes-
sional manner.
A review of the licensee's procedures demonstrated that
provisions were provided:
1) to verify that minimum water
level requirements are monitored during fuel handling opera-
tions; 2) to verify that the radiation monitors for the SFP
and VC are operable and checked; and 3) to ensure that the
SFP cooling and cleanup system is operable and that applicable
technical specification requirements were specified in the
refueling procedures.
The following procedures were reviewed:
Operating Procedure (0P) 1100, Revision 10, " Dismantling
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and Reassembly of the Reactor Systems for Core XIX
Refueling"
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OP-1209, Revision'11, Operation of the Vapor Container
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(VC) Fuel Handling Equipment
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OP-1214, Revision 13, " Component Movement within the New
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Fuel Vault'(NFV). and Spent Fuel Pit (SFP)
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OP-1700, Revision 13, " Cycle XIX Reactor Refueling.and
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' Component Inspection"
LOP-4226, Revision 14, " Testing of Fuel Handling Equipment
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with'the Dummy. Fuel: Assembly
The inspectors also ~ noted that SFP water level was higher '
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than'the minimum level required, personnel handling fuel were
properly supervised and an accurate map of fuel location
changes was being' maintained.
No' violations or safety concerns
were identified.
4.
On July 1, 1987 at approximately 11:35 p.m. with the plant in
Mode 3 the inspector noted during a control' room panel walkdown
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that the control rod primary position indicator 87" and 90"
light. emitting diodes for rod B-5 were essentially dark.
The
secondary position indicator (demand indicator) for the Group B
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-rods was indicating that the rods were at 90".
The. inspector
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questioned control room personnel and the reactor engineering
manager pertaining to this. observation.
They indicated that
this situation'was not a oroblem because the diodes were glowing-
faintly,
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The inspector reviewed TS 3.1.3.2, which requires in Modes 1 and
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2 that all control rod primary and secondary position indicator
channels be operable and capable of determining control rod
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positions within +/- three inches. Additionally, TS 4.1.3.2
specifies that each rod position indicator channel shall be
determined to be operable by verifying the primary position
indicator system and that the secondary position indicator
channels agree within three inches.at least once per four hours,
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Based upon these'TS requirements, the inspector informed the
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reactor engineering manager that it would be considered
unacceptable to the NRC to maintain the current. conditions and
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enter Mode 2.
At 12:20 p.m. the control room operators stepped
in the Group B rods to 87".
This action was allowed by TS 3.1.3.4 and procedure AP-7104, Rev. 68, Core Operational Limits.
The inspector verified that the licensee had not entered Mode 2
operations with inoperable rod position B-5 87" and 90" diode
indicators.
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In' addressing'the. inspector's concerns,~.the reactor engineering-
-manager noted that TS 3.1.3.1, which states that all control
rods which'are inserted in the core shall be operable and'posi-
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tioned within +/ .eight inches (indicated position) of every
.other rod in'their' group, is the limiting concern and the--
licensee will attempt to obtain a-TS change.to prevent.TS:
3.1.3.2 from being the limiting factor. . On' July 2,~1987, the.
licensee 11ssued Maintenance Request.No.87-959 to document the
'inoperability of the 87" and 90'.' diode indicators for~ control
. rod B-5 .-At-8:50'p.m. on July 2, 1987 the positioniindicating.
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-coil stack on control rod B-5 was-rep 1 aced.
The inspector had no further questions on-this item.
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b.
' System. Alignment Inspection
Operating' confirmation was ma'e of selected piping system trains.
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Accessible-valve positions and status were examined.
Power supply
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and breaker alignments were checked.
Visual' inspections of major
components were performed. Operability of instruments. essential to
system performance was assessed? The'following systems were checked
during plant tours.and control room panel status observations:
Emergency diesel generator unit
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Spent fuel cooling system
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Charging system (control board status observations)
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Low pressure and high pressure injection systems
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No unacceptable conditions were observed.
c.
Biweekly and Other Inspections
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(1) General Facility Observations
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During plant tours, the inspector observed shift turnovers,
compared boric acid tank sample analyses and tank levels to
technical specifications requirements and reviewed the use of
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radiation work permits and radiation protection' procedures.
Area radiation levels, air monitor use'and operational status
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were reviewed.
Verification of tagouts indicated the action
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was properly conducted.
No inadequacies were identified.
(2)
Fire Protection and Housekeeping
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Inadequacies noted regarding licensee housekeeping or fire
protection practices were minimal and tended to reflect the
outage condition of the plant for a majority of the inspection
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interval. A strong commitment to proper housekeeping conditions
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and practices by the plant staff.is routinely observed by the
Jinspector. Station management tours of the-facility minimized
.the period that housekeeping conditions.were. degraded.
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.(3) Observations'of Physical Security
Selected aspects of:plantJsecurity were reviewed.during. regular'-
- and backshift hours'to verify that. controls'were in accordance
'with the security: plan and approved procedures. Additional
details'regarding the. scope and findings of this review.aren
described in a separate special: inspection report, NRC Region I,
Inspection Report No. 50-29/87-08.
' Subsequent:to:a June"23, 1987 Enforcement Conference held at
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the:NRC:RI office, the inspector observed improvementsLin
station management oversight of security activities.
Items
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that reflected.an improving trend in'their concern for security
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were.
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Security training'will be scheduled for licensed -operators.
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The initial sessions of the segmented operator requalifi-
-cation program will' review the importance of security in
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facility operations.
The lesson plan developed will be
reviewed by the manager administrative services.
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2.
On July-2, 1987, the-manager administrative services
submitted to the chief of security suggestions for
increasing security force professionalism.
3.
A Daily Security Status report form was issued and became
effective on July 6,-1987.
This report provides the
manager administrative' services, who is responsible to
provide a closer interface between the licensee and the
security contractor, with a daily status of:
scheduled
activities, events, or postings; equipment problems and
maintenance;
scheduling problems; information on compen-
satory measures; and other items of interest or concerns,
4.
The vice president and manager of operations (VP/M00)
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issued on July 8, 1987 the licensee's implementation
matrix for security program improvement initiatives.
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Responsibility assignments and deadlines were specified.
The manager administrative services will provide a monthly
progress report to the VP/M00.
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5.
On July 16, 1987 a security advisory committee was formed.
Their function is to review and provide guidance in the
revision of the Security Plan, Contingency Plan, Traineg
and Qualification Plan, and security procedures.
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6.
With respect to security equipment maintenance, the
licensee's maintenance department has been assigned
responsibility for maintenance oversight on all security
systems.
This oversight will include provisions for
engineering support for security maintenance or modifi-
cation work, and development of a comprehensive security
preventive maintenance program.
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During this inspection p'eriod the inspector noted increased
attention by the station personnel to minimize the number of
control room door alarus.
Until equipment repairs and upgrades
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can be accomplished, the licensee is relying on station personnel
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cooperation to reduce unnecessary alarms.
Station management
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re-emphasized the prevention of unnecessary door alarms at their
July 9, 1987 general plant safety meeting.
The inspector noted
an improving trend in response to prior inspector concerns on
this item.
d.
Backshift Inspection
The inspector conducted backshif t, weekend or holiday inspections
on May 7, 12, 16, 18, 26, June 6, 13, July 3, 12 and 13.
No
unacceptable conditions were found.
5.
Review of Radiological Controls
Radiological controls were observed on a routine basis during the
reporting period.
Standard industry radiological work practices and
conformance to radiological control procedures and 10 CFR Part 20
requirements were observed.
Independent surveys of radiological
boundaries and random surveys of non-radiological areas throughout the
facility were taken by the inspector.
Based upon a review of licensee activities in this area the inspector
noted the following:
Early in the inspection period the inspector noted that licensee
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personnel were not logging out of the radiation controlled areas
(RCA) access control point at the end of the day.
The licensee did
not require logging out of a radiation work permit (RWP) each time
a person goes through the control point.
In response to the inspector's
concerns relative to the licensee's practices and the failure to log
out at the end of the day for a number of personnel, the radiation
protection department revised its access control practices. As of
April 29, 1987 the licensee required all personnel to log in and out
each time they go through the access control point for the RCA.
The
inspector determined that the licensee aggressively pursued correc-
tive action in response to NRC concerns in this area.
No further
problems were noted on this item during the inspection period.
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On'May.15,.1987 the; inspector observed the i: 7nsee's activities
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involved in lifting the reactor. head off the reactor vessel.- The
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inspector reviewed-ALARA Job Review No.87-024 dated May 14, 1987,
which consisted of.a job exposure estimate, pre-job checklist and an
.ALARA controls _ summary. 'As a. result of reviewing the reactor' head
, lift activities, the inspecto_r was able to verify that the preplanned
controls were' properly implemented in accordance with procedure
OP-8020,.Rev.-2, Implementation and Documentation of ALARA Job
Reviews. .The inspector.noted the.use of an extensive briefing o'f
. workers prior'to initiating the activity.
No unacceptable con -
ditions were.noted during the inspector's review of-this activite
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-On May 22,:1987 preparation for and. performance of tube plugging in
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the No. 2 steam. generator was in. progress. Westinghouse workers had
been selected to enter the steam generat'or channel head to clean and
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plug designated degraded tubes.
Due to dose rate gradients within-
'the channel head,. multiple dosimeters were issued to each worker to
insure-that' appropriate. areas of the body were accurately monitored
'(a total-of 13,were used). The work proceeded without: incident and
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in accordance with pre-established stay times.
Subsequently, the
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TLDs worn by the workers were read and the results were reviewed.
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Initial licensee ~ evaluation of the data indicated that one of the
workers had .two TLD results (right upper. arm and right thigh) which
appeared to have exceeded regulatory limits for whole body dose.
The two TLDs.in. question had results of 3.118 and 3.399 rem as the
assignment of whole body dose, which was based upon the licensee's.
exposure calculation assumptions that the individual's dose had been
delivered through 300 mg/cm squared density thickness tissue.
Because this was a potential overexposure situation, the plant
radiation protection department conducted further evaluations that
' included technical assistance provided by the YNSD radiation pro-
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tection group and environmental laboratory personnel.
The licensee
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determined that the individual's exposure to the whole body, using
the TL0s in question, should be calculated based upon the dose being
delivered through 1000 mg/cm squared density thickness tissue. This
subsequent determination brought the dose to 1.234 and 1.151 rem for
the right thigh and right upper arm, respectively.
Following the
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licensee's identification of this issue to the inspector on May 27,
1987, the event and the licensee's actions were discussed with
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NRC:RI Division of Radiation Safety and Safeguards (DRSS) specialist
inspectors. The event and the licensee's evaluation will be reviewed
by the NRC during a subsequent routine facility inspection of the
radiation protection program (Unresolved Item 50-29/87-06-01).
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Contamination was discovered by a licensee radiation protection
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contractor on May 30, 1987 as he frisked himself at the RCA.
This
event involved hot particle control issues and a potential for the
individual's whole body skin exposure to be in excess of the
regulatory limits.
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After learning that a contamination incident involving.a hot
particle had occurred, the inspector notified NRC:RI (DRSS) cognizant
personnel of the occurrence on June 4, 1987.
Subsequent to several
NRC:RI-initiated conference calls with the licensee, a special NRC
safety inspection was conducted on June 15-19, 1987.
The NRC's
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review of the incident and inspection findings are documented in
Inspection Report 50-29/87-10.
6.
Review of Events Requiring Telephone Notification to the NRC
The circumstances surrounding the following events, which required NRC
notification via the dedicated emergency notification system (ENS)
telephone line, were reviewed. A summary of' the inspector's review-
findings follows or is documented elsewhere as noted below:
At 8:58 a.m. on May 6, 1987 the NRC was notified in accordance_with
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50.72(b)(1)(ii) that the Containment Integrated Leak Rate Test.
(CILRT) was declared a test failure. This resulted from identifica-
tion of a significant,.but not quantifiable leak at a blank flange in
the fuel cycle dewatering system.
Subsequent investigations into
this event by the licensee concluded that a loss of containment
integrity had not occurred.
This matter is discussed further in
Section 9 of this report and is also the subject of LER 50-29/87-11.
On May 11, 1987 at 12:52 p.m. and 1:52 p.m. the licensee made
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notifications to the NRC via the ENS line pertaining to an event
involving a failure of a security officer to receive prior plant
shift supervisor approval for a visitor entry into the vapor
container on May 10, 1987. NRC review and inspection findings for
this item are documented in Special Security Inspection Report
50-29/87-08.
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At 11:15'a.m. on May 22, 1987, the NRC was notified of a condition
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involving fuel degradation in accordance with 50.72(b)(2)(i).
This
event involved the visual observation of failed fuel pins on fuel
assembly A731 as it was being removed from core location H-9.
The
observation was made at 9:05 p.m. on May 21, 1987.
Following a plant
management review of the event at 7:45 a.m. on May 22, 1987, plant
management determined that this was a reportable event.
Since the
primary coolant chemistry data for the prior operating cycle indi-
cated the presence of failed fuel, and considering that broken open
fuel pins were experienced on the prior core, the observed condition
was not considered to be highly unusual.
In the past, the operations
department has issued instructions prior to performing fuel moves for
operating personnel to place a four-hour ENS call when a damaged fuel
assembly is discovered.
To provide a more consistent deportability
determination technique for failed fuel related events, the inspector
recommended to the technical services manager (TSM) that additional
guidance on this subject be included in OP-MEMO-2A-1, Policy for
Immediate Notification of the NRC of Significant Events.
The TSM
!
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acknowledged the inspector's comments and concerns related.to pro-
.
- viding a more timely deportability determination'for this type of
1
event and agreed to incorporate the.NRC recommendation in the
l
OP-Memo.
Further details on the event and the' licensee's correc-
tive; actions are discussed in Section 7 of.this report.
On May 25, 1987 at 10:30 a.m. the licensee made a 24-hour ~ noti-
---
fication to;the NRC.via the ENS line in accordance with 10 CFR.
'73.71 that an event involving confirmed tampering with security
equipment had occurred on.May-24, 1987 at 10:57 a.m. .The licensee
j
submitted..to NRC Region I their five-day written report describing
l
'
the event on May 29,'1987._ At the time,of. occurrence, the inspector
discussed the event details ~and licensee actions with cognizant-NRC
d'
Region I personnel.
This event and the licensee's response actions.
'will be reviewed as part of.a future routine regional safeguards and
security inspection.
At 10:40 a m..on June 1, 1987 the NRC was notified in accordance
--
.
withz50.72(b)(2)(ii) that at 9:20 a.m. a loss of power event
occurred.that resulted in the start of the'Nos. 1 and 2 emergency
' diesel generators.
During this notification, the NRC was also-
informed that another event had occurred on May 31, 1987 at'1:10
p.m. that resulted in an unplanned automatic start and phasing-of
the No. 1 emergency diesel generator onto the 480V emergency No. 1
bus.
These events are discussed in Section 7 of this report.
Regarding the May 31, 1987 event, the inspector determined that
i
the licensee failed to make a timely notification of the event in
accordance with 50.72(b)(2)(ii), which required a four-hour report.
The licen'see attributed this failure to personnel error on the part
'
of the shift-supervisor to make the required determination of
deportability,
Counselling of the individual was conducted by licen-
.
see management to preclude recurrence.
Because this appears to be
an isolated incident and the criteria of 10 CFR 2, Appendix C, have
been met, no Notice of Violation will be issued.
At 2:30 a.m. on June-20, 1987 the NRC was notified in accordance
--
with 50.72(b)(1)(v) that a major loss of emergency communications
capability had occurred at 2:10 a.m. as a result of the Ames Hill
(Marlboro, Vermont) section of the Public Notification System (PNS)
l
being out of service. .The system was returned to service on June
21, 1987.
The inspector noted that the control room operators
,
utilized OP-MEMO 20-5, Public Notification System, to assess the
deportability of thi.s one-hour non-emergency event.
At 8:25 a.m. on July 6,1987, the NRC was notified in accordance
--
with 50.72(b)(2)(ii) of an automatic reactor scram that occurred
at 7:38 a.m. as a result of high main coolant system pressure that
occurred while performing a turbine overspeed test.
This event is
discussed in Section 7 of this report.
,
No inadequacies were identified, except as noted above.
L
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_
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7.
Review of Plant Events
a.
'Coastdown Operations
'
From April 27, 1987 u'ntil-May.2,1987,theplantoperatNcontihually.
in Mode-1 end-of-cycle coastdown.
During this' period' preparations:
,
were made by-the licensee'for Cycle XVIII-XIX refueling,
During the
' shutdown process on May 2, 1987, control room operators l shut the non-
return valves (NRVs) on the main steam lines in preparation;to break;
condenser vacuum. -At this time a control room operator reported that
the-train A control switch:for the No. 2 NRV would not close the
valve. The valve was subsequently closed u' sing the train B switch.
Maintenance request'(MR)87-552 was issued by the operations staff
.
,
to provide corrective maintenance for the occurrence.
~
The inspector noted that an identical occurrence was experienced on
February 18, 1987 following a plant trip. .The licensee's actions
were' described in Section 4.a (4)-of. Inspection-Report 50-29/87-02.
Because this appeared to be a recurrent. situation, the inspector
discussed the planned troubleshooting activities with the. instrument-
and controls (I&C)' supervisor on May 4,-~1987, who, demonstrated his-
concern'and~ understanding for the need to determine root cause..
However, unknown to the I&C department at-the time, the maintenance
' department issued MR 87-560 on May 2, 1987 to facilitate the v'alve's
disconnection and removal for scheduled maintenance overhaul by the
manufacturer (Rockwell International) without initial troubleshooting
being performed to attempt to' isolate the cause of.the occurrence.
The inspector's concern that outage scheduling activities may have-
taken priority over.the conduct of a root cause investigation was -
discussed: o'n May 7,1987 with the plant superintendent who acknowl-
edged the inspector's comments and concerns on.this matter. At the
May 8, 1987 PORC meeting, which the. inspector attended,.the plant
superintendent discussed with the committee members the importance
of conducting a thorough investigation for safety-related equipment
failures, especially when there is a' recurring failure being
experienced.
On June 25, 1987 the I&C Supervisor issued a memorandum that
discussed the NRV's recurrent failures and possible causes, which
included:
1) the electrical circuit developed a malfunction which
prevented a valid close signal from reaching the train A dump
solenoid valve; 2) the train A dump solenoid valve received a valid
,
close signal but did not operate; and 3) the hydraulic fluid ports to
'
the train A dump solenoid were plugged.
The manufacturer's. overhaul
conducted per MR 87-560 resulted in finding no evidence of hydraulic
port plugging.
The I&C department was also able to discount the
,
L
other two potential causes. On May 8,1987 the manufacturer informed
the licensee of a 10 CFR Part 21 condition that involved inadequate
,
?
capacity of the thermal compensating accumulator in the NRVs.
This
condition could, under certain high hydraulic pressure conditions
l
caused by elevated temperatures, cause the dump solenoid valves to
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14
fail to operate.
Licensee MRs87-780, 781, 782, and 783 were
initiated on June 3, 1987 to implement the manufacturer's recom-
,
'
mended modifications to resolve the 10 CFR 21 concerns.
The I&C
>
Supervisor's memorandum specified that the most likely cause of the
recurrent malfunctions was the 10 CFR 21 identified concern. This
memorandum was reviewed by the PORC at its June 26, 1987 meeting
with the committee noting that the NRV will be cycled during startup,
when the system is at operating temperatures, to assure operability.
Subsequent to overhaul and modification work performed to correct the
10 CFR 21 deficiency, the No. 2 NRV was time tested on June 14, 1987
using procedure OP-4654, Rev. 4, Calibration of the NRV Automatic
'
Control System.
However, due to unrelated problems with this valve,
g
MR 87-852 was issued which resulted in the replacement of the train A
,
dump solenoid.
Because this maintenance could affect valve closure
timing, further retest was warranted.
The retest was conducted on
.g
July 3, 1987 with the plant in Mode 2.
The control room log entry for July 3,1987 acknowledged the plant
operator's awareness for the need to time test the valve per OP-4654.
The log also indicated that the action statement of TS 3.6.2, Con-
tainment Isolation Valves, was entered for the period that the valve
was opened to facilitate the test.
However, the inspector noted
that according to Note 11 in TS Table 3.6-1, the action statements
of both TS 3.6.2 and 3.0.3 do not apply.
What does apply is TS 4.6.2.1.b, which specifies that the NRVs shall be demonstrated
operable immediately prior to returning the valve to service after
maintenance, repair or replacement work is performed on the valve
of its associated actuator by operating the NRV through at least
one complete cycle of full travel and verification of isolation time.
This surveillance requirement was recognized as being applicable in
!
the control room log. According to TS 3.6.2, the NRVs are to be
3
[
operable in Modes 1-4 to provide a containment isolation function.
'
Furthermore, TS 3.7.1.5 provides main steam NRV LCOs and surveillance
requirements when the plant is in Modes 1-3.
The inspector was aware of past unsuccessful licensing efforts by the
licensee to remove the NRVs and other secondary side valves from the
containment isolation provisions of the TS. As a result of apparent
confusion pertaining to NRV TS, the inspector discussed the matter
i
with the NRC:NRR project manager who agreed that it would be appro-
priate for the licensee to resubmit for NRC review a licensing action
to remove the NRVs from the containment isolation provisions of the
TS.
The licensee concurred with the inspector's observations for the
need to modify the TS as enumerated above, and planned to initiate
the licensing action in the near future.
The inspector had no further questions of the licensee on this
matter at this time.
.
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CycleXVIII-XIXRefuelingOpegplps
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and plant nduelin,# operations commenced. 'The ' }
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control' rods' inserted 4
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- reactor remained in Modes'S op 6 until June- 27, 1987..
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In@ector[J
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observations:are noted bd ow:
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_ During the.' performance ,df ultrasonic inspectiop on ricycled
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cfuel~on May' 22,;1947, the licensee discoverea that fuel 'assemb3 -
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A-731 from coreVr> cation 4 9 was found to hdy'e sustained fEetting-
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, damage to the fuel rods a'nd spacer grids- on the side l adjage)f to :
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the core baffle WJ. JAll ' fuel rocs in the assembly hat edd)
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current testing peformed. with thirteen rods determined to/ r
have frettin'g danio ,e
rcarrodsv;cev..ibly;observeddehave
through-wall dafect;. llnedamao'.) assembl,wasreconsytuted
f
in Cage A5v
The litu see awisted Combustion Engineering r
,
'
.personnelwiththerepaliwhichwasconductedinaccordanced<'k,
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with pro:odure ranber40k QJ2, dated May 18, 1987.
The damage
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fuel rods: ware.repi ped with inert. rods.
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No additional fuel assor$@es were found by ultrasonic esamin-
]{
.
ation to have. failed ft.si rods.
The licensee believes that the
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'
frettingLdamage to the 'od cladding and spacer grid of assembis ?
f A;
A-731'is attributabichte e. ore baffle spacer flow Jetting.
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' Corrective ' action in this refueling outage for recurrent fuel
-
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assembly damage in c d positions susceptible.to the' flow jot-
1
ting anomoly consisted \\of 1) . installation of eighteen b'affle / .
j
spacer plugs to reduce the pressure behind all remaining ba'fie
i
spacersinorderto'e$minateanybaffle'spacerflowjettingin
-
the core periphery, and 2) eight core positions that had previ-
ously. experienced the llow ind/ced fretting were loaded with-
fresh cycle XIX ful assemblies that were constructed to reduce
!
the' possibility of damage due to flow co6dit Gns.
The design
,
.and installation of the core baffle spacer plugs were controlled
by Engineering Desigb Change Requsst 87-301.
The 1Mensee's use
of core baffle plugs,and strengthened fuel design at core posi-
tion C-9 during the 1985 refueling outage in response to fretting
induced fuel failure was effective in precluding recurrence at
'
this location.
On June 20, 1987, the licensee reported the fuel degradation
of fuel assembly A-731 in core position H-9 as LER 50-29/87-09.
31, 1987 at 1:10 p.m'. with the plant in Mode 6,
L
On Sunday May
--
the Nos. 3 and 6 station service atranformers (SSTs) were
de-energized by relcy action. This condition resulted in the
automatic starting and loading of the Noc 1 ercergency diesel
generator (EDG). Normal electrical power lineup was restored.at
approximately 5:00 p.m.
The licensee attributed the c wse.of
this event to excessive vibration of a control room pknel, which
was induced by cutting a panel opening for new relaying that was
1
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being instal, led as part of a plant modification.
The details of
the event, the licensee's-investigation and their corrective
4
action to preclude recurrence are contained in the licensee's
I
LER 50-29/87-08 submitted to the NRC on June 30, 1987.
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At the time of the event, YNSO projects engineering personnel
were onsite providing cognizant engineering duties as part of
modification responsibilities.
The inspector noted that the
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licensee was able to effectively integrate this engineering
~
expertise into their efforts to conduct a thorough investi-
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gation of the event.
--
At 9:20 a.m. on June 1, 1987, with the plant in Mode 6, the
3-
licensee was making preparations to remove the 115kV Cabot
!
Station transmission line (Y-177) from service to allow mainte-
nance on the line and inspection of oil circuit breakers in
j
the plant switchyard. As a result of a control room operator
,
switching error the Zone 1 impedance protective relaying was
'
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actuated.
This resulted in the opening of switchyard and onsite
distribution breakers.
Because 480V ac buses 4-1 and 6-3 were
tied together, both EDG Nos. I and 2 automatically started and
loaded onto their respective emergency buses.
Normal electrical
power lineup was restored at 9:35 a.m.
This event resulted in
,,
,
b
the plant having less than the minimum electrical buses energized
as required by TS 3.8.2.2, and in response to this condition
_
control room operators initiated the action statement requirement
7
to establish containment integrity.
The details of the event, the licensee's investigation and
their corrective action to preclude recurrence are contained
in the licensee's LER 50-29/87-10 submitted to the NRC on June
30, 1987.
The inspector had no further questions of the licensee regarding
the above events.
No violations of regulatory requirements were
identified by the inspector.
ko
.,
c.
Plant Startup and Power Operations
'
!
Plant heatup to Mode 4 was initiated on June 27, 1987.
Initial
criticality of Core XVIII was achieved per OP-2103, Reactor Startup
and Shutdown, and OP-1701, Core XIX BOL Zero Power Physics test. The
y/
's
measured Critical Boron Concentration at hot zero power with all rods
out was 1914 ppm, which was within +/- 10% of the calculated design
value (1873 ppm) and within the acceptance criteria of the applicable
operating procedure. At 1:55 a.m. on July 3, 1987, the reactor
,
engineering department completed all startup physics testing and,
following data reductions and review, determined that all tested
parameters met required acceptance criteria.
The inspector found
that the requirements stipulated in procedure OP-1701 for prereq-
uistites, procedure steps, verification signatures and acceptance
,
'
criteria were met.
No discrepancies were identified.
_ _ _ _ _ _ - _ _ _ _ _
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.
17
On July 3, 1987 the plant entered Mode 2 operations at 4:38 a.m.
following the reactor being made critical.
Plant power escalation
continued through July 5,1987 with a maximum power level of 60% of
rated power being obtained.
Subsequent to the licensee's investi-
gation and resolution of equipment problems experienced with the
turbine governor control system and the left hand turbine throttle
valve (see discussion that follows), the licensee performed proce-
dure sections A-F of OP-2102, Rev. 11, Turbine Testing and Startup
and decided to perform section G, the turbine overspeed trip test,
on July 6, 1987.
At 7:38 a.m., on July 6, 1987, while the plant was at about 10% of
rated power, a reactor trip was initiated due to high main coolant
system (MCS) pressure.
Just prior to the trip, the turbine had been
taken off line after two days of post-outage testing, and operating
,
personnel were preparing to conduct an overspeed turbine trip test
in accordance with OP-2102, section G.
Those preparations required
normal reactor operation from the control room and local (turbine
floor) operation of the main turbine throttle valves and control
valves. When the left-hand throttle valve failed to open properly,
the right-hand throttle valve was used for turbine startup. While
increasing the turbine speed the turbine operator opened the throttle
valve too fast, which resulted in an excessive steam demand.
Between
the reactor operator's response in pulling control rods to match
reactor power to steam demand, and the excessive throttle valve
,
opening that subsequently resulted in turbine overspeed and control
valve closure, conditions were established to effect a rapid main
coolant system heat up and corresponding reactor pressure increase.
The licensee attributed this event to 1) excessive throttle valve
operation caused by personnel error, and 2) using the Gaitronics
plant -announcing system in lieu of direct " head set" communications
between the control room panel operator and the operator on the
turbine floor.
The latter condition resulted in poor coordination
of operator activities during the turbine startup. All equipment
functioned normally in response to the reactor trip. The turbine
overspeed test was completed satisfactorily on July 7, 1987.
The
licensee plans to document the event, their investigation, and
corrective actions in LER 50-29/87-12.
8.
Maintenance Observations
The inspector observed and reviewed maintenance and problem investiga-
tion activities to verify compliance with regulations, administration and
maintenance procedures, codes and standards, proper QA/QC involvement,
safety tag use, equipment alignment, jumper use, personnel qualification,
radiological controls for worker protection, fire protection, retest
requirements and deportability per Technical Specifications.
The
following activities were included:
_ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _
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18
~ Maintenance Request (MR) 85-1495, Repair MOV-302 Leaking Bonnet
--
MR 86-1189,-Incore Thermocouple D-5 Indicates Open
--
MR 86-1746, Incore Thermocouple D-4 Faulty Indication
--
MR 87-552, No. 2 NRV Train A - Failed To Close On Manual Signal
--
MR 87-560, MS-NRV-405B (No. 2) - Overhaul With Factory Rep
--
MR 87-780, MS-NRV-405A (No.1) Actuator - Add Thermal Compensating
--
MR 87-781, MS-MRV-405B (No. 2) Actuator - Add Thermal Compensating
--
.
'
MR 87-782, MS-NRV-405C (No. 3) Actuator - Add Thermal Compensating
--
Accumulato-
MR 87-783, MS-NRV-405D (No. 4) Actuator - Add Thermal Compensating
--
MR 87-598, Fuel Chute Dewatering System - Blind Flange Leaking
--
During Type A CILRT
--
MR 87-738, Tube Plugging of Steam Generators Nos. 3 and 4
MR 87-755, No. 1 LPSI Pump ACB Will Not Close
--
MR 87-959, No. 5 Rod Indicating Stack - 87" and 90" Light Out
--
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MR 87-8-17, No. 3 EDG Day Tank Level Indicator Power Lamp -
Indicator Blinking
--
MR 87-958, No. 1 EDG Day Tank Level Control - Day Tank Fills To High
--
MR 87-1032, FO-L-1 No.1 EDG Day Tank Level Control System - Will
Not Meet The Acceptance Criteria of OP-6475
Based upon a review of licensee activities in this area the inspector
noted the following:
In Inspection Report 50-29/87-02, Section 11, the inspector reviewed
--
the licensee's repair activities associated with a malfunction of
the air circuit breaker (ACB) for the No. I low pressure safety
injection (LPSI) pump that occurred on March 13, 1987.
On May 31, 1987 another failure of the ACB to close occurred. The MR
87-755 was issued to control the repair activities. An investigation
was initiated by the licensee to determine root cause, which included
an independent evaluation by a vendor representative (General
Electric).
The cause of the equipment malfunction was determined to
)
be an out of adjustment trip free latch.
This condition caused the
trip bar to stick in mid position.
Following re-adjustment per the
j
applicable plant procedure, the ACB was retested and returned to
i
service.
The licensee discovered from the vendor representative that
their lubrication practices for the AK series breakers, which covers
)
this and other ACBs used at the plant, were outdated.
Vendor docu-
)
ments dated 1979, 1984, and 1985 that provided new lubrication
recommendations were never received by the plant's maintenance
department.
In response to this new information, and as a precau-
tionary measure, the licensee issued a purchase order to the vendor
j
to repair / refurbish ten of the AK25 breakers.
This effort consisted
of installing new trip bar bearings and latch rollers, and rebuilding
of the overload current trip devices.
Prior to startup from the
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19
!
i
refueling outage eight of the ACBs were refurbished. Two ACBs which
'
are manually-operated backup 480VAC supplies to the manual throwover-
switches, are scheduled for refurbishment during the next refueling
outage.
The licensee is also considering at the next refueling
outage the refurbishment of AK50 emergency diesel generator output
breakers. Additional inspector findings pertaining to the mainte-
nance activities for the breakers are as follows.
(1) A maintenance request to control the repair / refurbishment
effort on the AK25 ACBs was not issued.
The maintenance
department used the material and s'ervice procurement program,
which is described in procedure AP-0211, Rev. 15, to control
.
the work activities.
The licensee's maintenance and quality
'
control personnel acknowledged the inspector's concerns that
the use of the AP-0211 process did not provide the same level
i
of programmatic control required by AP-0205, Rev. 11,
1
Maintenance Request, which is required when corrective
maintenance is performed.
In addition, procedure AP-0214,
Rev. 10, Maintenance and Change of Safety Classified Systems,
Components, or Structures specifies that: corrective maintenance
performed on safety classified systems,. components, or struc-
tures, utilizing materials, parts or item replacement or docu-
i
mentation, shall necessitate the generation of a Maintenance
Request in accordance with AP-0205, Maintenance Request.
The
inspector noted that: 1) the licensee planned to issue a Non-
conformance Report to document and provide corrective action for
the failure to issue an MR and 2) the inspector identified no
unacceptable equipment conditions because of the failure to
issue an MR.
(2) Extensive involvement by the licensee's quality control (QC)
group was noted during the repair activities involved with the
ACB and the subsequent refurbishment efforts. As part of
reviewing the work activities being conducted under MR 87-755,
the QC group performed an equipment history review.
They
.
determined that during the conduct of procedure OP-4506, Rev. 9
I
Inspection of ECCS Circuit Breaker No. LPSI-1, on May 23, 1987
maintenance personnel replaced the Phase B overcurrent trip
device, but no MR was initiated.
The QC supervisor initiated a QC
Resolution Request (RR) to document the deficiency and sent it
for resolution to the maintenance supervisor.
The RR is
controlled by YNSD quality assurance Department procedure
l
0QA-X-5, Quality Control Inspection, which was intended to be
used for specific conditions requiring resolutions that are not
j
directly applicable to an Inspection Report.
The inspector
'
expressed concern that both plant management and the Plant
Operations review committee would not, in a programmatic
,
1
manner, become aware of and participate in the resolution of
conditions adverse to quality when an RR is issued. The
failure to issue a MR, as identified by the QC group, warranted
!
^
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20
the issuance of an NCR.
Plant procedures AP-0206, Rev. 8
Nonconforming materialsi Parts, and Components; AP-0209, Rev.
8, Inspections; and AP-0227, Rev. 4. Corrective Action, support
the need for issuance of an NCR for the identified condition.
Additionally, the release of the replacement overcurrent trip
device without an MR number is contrary to the requirement
established in procedure AP-0213, Rev. 9, Identification and
Control of Materials, Parts, and Components. Although this
condition did not result in unacceptable equipment conditions,
it does illustrate the need to strengthen the knowledge level of
station personnel to these important programmatic requirements
involving maintenance activities.
The inspector was informed by
the quality assurance department that an NCR would be issued to
resolve the concerns involved in failing to issue an MR and
procedure OQA-X-5 will be reviewed and revised to clarify the
relationship between an RR and an NCR.
(3) NRC Information Notice (IN) 87-12, Potential Problems with
Metal Clad Circuit Breakers, General Electric type AKF-2-25,
was issued on February 13, 1987.
Initial licensee
screening performed on February 27, 1987 on the Evaluation of
Operating Experience Form, APF-0020.1, indicated that the
maintenance support department reviewed the event and
determined that the subject breaker was not used at the plant.
The evaluation form was placed in concurrence routing
including PORC review.
The inspector noted that: 1) as of the
end of the inspection period the evaluation of this operating
experience feedback was not completed, and 2) the IN specified
that GE has issued Service Information Letter Number 448 to
address special maintenance practices developed for the entire
AK type breaker series.
During a June 3-7, 1985 inspection
conducted to address the concerns identified in NRC Generic Letter (GL) 83-28 (Salem-ATWS), as documented in Inspection
Report 50-29/85-12, the licensee's program to provide
assurance that vendor information is current and complete as
required by paragraph 2.2.2 of the GL was determined to be
unacceptable (Unresolved Item, 50-29/85-12-2).
The failure of
the licensee to establish and implement a program to ensure
that vendor information for safety-related components is
complete, current and controlled throughout the life of the
plant warrants additional management attention to resolve this
issue.
Inspector review of the event details and maintenance activities
--
associated with MR 87-598, issued on the fuel chute dewatering
system are contained in Section 9 of this report.
,
_
_ _ _
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Regarding:MRs87-847,.958, and'1032 the-inspector was-informed by
--
the licensee-on July 8, 1987 that an improper solenoid. valve
(F0-LCV-701) was installed in the.No.1 EDG fuel oil system. :Imme-
"'
diately following this discovery.the EDG was declared inoperable.
.)
. with"the actions'specified in TS 3.8.1.1'being implemented.
The
,
- incorrect installation of'a'non-environmentally' qualified (EQ): valve
'
occurred on June 13, 1987'with a safety' class EQ valve installed'on
- July 10, 1987. . The. plant's three.EDGs were; required t'o'be operable'
when the plant entered Mode'4~ operations on June.27, 1987. A YNSD
engineering evaluation (YRP 678187) dated July'13,'1987 concluded
that the improperly installed valve would have functioned as required.
(1.e., automatically open on a low level in the No. 1 EDG-day tank to
allow gravity-fill from the fuel. oil storage tank) and should be-
- considered as'EQ.
The licensee viewed this: incident as one involving a major breakdown
in imp.lementing their maintenance system for safety related equip-
ment.
In response to this condition, the inspector noted additional
oversight was provided by corporate and station management to
- identify both the root causes of the occurrence-and appropriate-
corrective actions.
As a result of their investigation, the
licensee attributed this occurrence to 1) confusion as to the equip-
ment that required maintenance and poor verbal communications- between
supervisors, QA,; maintenance and operations personnel; 2). assignment
{
- of the repair activities to the maintenance department in lieu of the
..I&C department; and 3) failure to implement requirements of procedure
AP-0205 for the maintenance activity as noted below.
(1) MR 87-847 was issued,with an improper. determination that QA and
EQ were not applicable. A subsequent control room personnel
determination,that QA was required only resulted in a change to
control room logs and not the MR itself.
(2) The work performed by the maintenance department was done
without an MR in their possession, and no QA group review of
the intended activity was performed prior to initiation of the
repair activity.
' (3) The requirement that the MR be reviewed by the cognizant depart-
1
ment supervisor was not implemented and no special instructions
or specification of maintenance procedure to be utilized to
control the performance of the corrective actions was provided.
In effect no MR was initiated for the work that was performed.
In response to their investigation, the short term corrective
actions implemented by the licensee were:
(1) Respective department managers conducted briefings of
maintenance and operations personnel as to the causes,
l
.
.
________m
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_
_ _ _ - _ _ _ _ _ - _ - _ _ _ _ - _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ . _
. . _ . .
,
,
22
problems, consequences of the occurrence, and discussion
on the requirements and need for compliance with the
provisions of procedure AP-0205.
(2) The assistant plant superintendent was to conduct special
l
training sessions for maintenance, maintenance support, I&C,
and appropriate training personnel.
Besides reviewing this
occurrence and other recent incidents at the plant involving
maintenance, the training sessions will include the following
elements: review procedure AP-0205 and the basis for the
procedural steps; proper use of the safety class manual;
proper job planning and briefing prior to work; supervisor
observation of job performance; quality in work practices;
proper job assignment and close out; and maintenance
department responsibilities versus QA department
responsibilities.
Longer term corrective actions planned by the licensee included:
(1) The assistant plant superintendent, maintenance manager, and
maintenance and I&C supervisors will provide specific recommen-
dations to the training department so that the lessons learned
from this event can be incorporated into future training.
(2) The assistant plant superintendent and the operations manager
will provide specific recommendations to the training department
so that the lessons learned from this event can be incorporated
into the SRO training program.
(3) The assistant plant superintendent will take a more active role
in future development of the training program for the
maintenance and operations department personnel.
The licensee intends to issue a plant information report to
facilitate PORC review of the subject occurrence.
In response to
this and other occurrences enumerated earlier, the inspector
held discussions with the plant superintendent and the assistant
plant superintendent about their intended corrective measures.
The
licensee characterized their corrective action efforts into four
areas: 1) upgrade of facilities for the maintenance and maintenance
support groups that would improve needed communications between
maintenance personnel, supervisors and support engineers; 2) identify
manpower deficiencies in the engineering positions within the main-
tenance support department and first line maintenance supervisory
positions, and provide input in the upcoming budgeting process; 3)
upgrade maintenance department procedures to aid in training of
replacement personnel and insuring the performance of quality work;
and 4) provide improvements in training of personnel involved in
maintenance activities.
The licensee's evaluation of resource needs
- - - .
..
,
23
and areas requiring strengthening is consistent with the inspector's
observations in the maintenance area. The licensee has been candid
with the inspector about their concerns for the apparent' decrease in
maintenance performance, and have exhibited a proper level of criti-
cal introspection to ascertain the underlying reasons for their cur-
i
rent performance in this area.
'
The inspector considers this occurrence to be a licensee identified
violation, and in accordance with the provisions of 10 CFR 2,
,
Appendix C, a' Notice of Violation will not be issued.
!
As part of the repair of the leaking bonnet on the loop 2 cold leg _
--
stop valve, which was implemented during the refueling outage by MR
85-1495, the valve was disassembled and cracks were found on both
disks.
Inspections of the as-found condition of the disks were made
by Plant and YNSD mechanical engineering services group personnel.
Preliminary evaluations, as documented in licensee memorandum MSM
53/87 dated June 26, 1987 indicates that the cracks were most
l
likely attributed to original fabrication, materials and/or
techniques with the cracks existing for most, if not all, the past
27 years of operations.
The' licensee has replaced the cracked disks
with spare parts.
Licensee memorandum YRP 619/87 dated June 28,
1987 provided a safety evaluation for operation of the plant with
the possibility of similar cracks in other loop stop valves. This
safety evaluation was reviewed by the PORC at a meeting on June 26,
1987. This meeting identified the need to address the consequences
of spalling off of small segments of the disk's stellite hard face
as a result of disk failure due to crack propagation. Additional
information obtained from the YNSD Yankee projects group, as
contained in memorandum TS 87-68 dated June 27, 1987, addressed the
PORC concerns and was reviewed by the committee on June 29, 1987.
The inspector discussed the issues associated with the disk crack-
ing with the cognizant NRC: Region I Division of Reactor Safety
]
specialist. All referenced evaluations were transmitted to the
l
specialist for his review.
No unacceptable conditions were
l
identified.
The licensee's corrective actions, in addition to those
j
mentioned above, consists of 1) sending the cracked disks to a vendor
for metallurgical examinations; 2) procuring spare disks and stems;
3) developing an inspection program for the next refueling; and 4)
the issuance of Special Order No. 87-71 to the Plant Operators that
makes them aware of potential valve problems due to disk cracking and
provides notification guidance as a result of experiencing unusual
valve operation.
The inspector observed that the maintenance performed on the
unisolable main coolant stop valve was in accordance with appropriate
plant procedures; considered the guidance provided by NRC Information Notice 87-23, Loss of Decay Heat Removal During Low Reactor Coolant
Level Operation; and incorporated recommendations from INPO evalua-
tions which addressed similar events.
The licensee's approach to
.
_ _ - _ _ - - _
_-
.
. - _ - - -
- ,
<
y
+ , ,
M
'
' +
24
,
n
.'
.
implementingithe work. activity was conservative in nature, as evi-
'
- denced by the selfLimposition of maximizing the time that refueling;
'. integrity was maintained and utilizing temporarily installed reactor
head thermocouple.
,
%~
"
_The inspector identified no deficiencies on this item-and.will:
,
review the'results of the metallurgical examination of~the-
-cracked disks during' future inspections'(Inspector _. Follow Item
50-29/87,-06-02).
9.
Surveillance Observations
.The inspector observed tests-or parts of' tests to assess performance.in
accordance 'with approved procedures and LCOs, test results (if completed),
removal.and-restoration of equipment and deficiency' review and resolution.
The'following tests were reviewed.
OP-4611,-Rev. 17, Nuclear Instrumentation'and Reactor Protection.
--
.
System Precritical Check.
.0P-4212,.Rev. 4, Main Coolant System Residual Heat Removal
--
. Availability Verification
OP-1708, Rev. 1, Vapor Container' Type A Leakage Test
--
OP-4572,fRev. 2, 3'and.4, No. 3 Station Battery Service Test
--'
'0P-4570, Rev. 2, No. 1 Station Battery Service Test
---
--
.0P-4571,. Rev. 3, No. 2 Station Battery Service Test
0P-4703,.Rev. 11,-Control' Rod Drop Time Measurement
---
-OP-7106, Rev. 6, Calibration and Operation of the Westinghouse
- - -
Reactivity Computer Model NBSU 8094'
OP-1701, .Rev. 9, Core XIX BOL Zero Power Physics Test
--:
OP-4702, Rev. 12, Vapor Containment Type B&C. Penetration Tests
---
' 0P-7222,: Rev. 8, Reactor Rod Control System Precritical Check
--
--
0P-7203, Rev, 8, Eddy Current Examination and/or Repair of Steam
Generator No. 2,.3'and.4
OP-4522, Rev
7, Inspection and Maintenance of Station Battery No. 3
--
OP-4500, Rev. 9, Weekly Check of the Station Batteries
--
OP-4501, Rev. 9, Quarterly Check of the Station Batteries
--
OP-4519, Rev
6, Station Battery Discharge Test Battery No. 3
--
a
OP-4201, Rev. 13, Power Range Channel Calibration Heat Balance
--
OP-4708, Rev. 11, Determination of Shutdown Margin
l
--
OP-4215, Rev. 6, Surveillance of the Boron Injection Flowpath
--
OP-4239, Rev. 9, Verification of Containment (VC) Integrity and
--
Operability Check of the VC and Spent Fuel Pit (SFP)
>
OP-4220, Rev.13, Primary System Water Balance
!
-
--
OP-4271,-Rev. 2, Leakage Check of the Neutron Shield Tank
--
- Based upon a review of licensee activities in this area, the inspector
i
noted the following:
1
!
l
x
- _ - -
-
i
_ .
-
._
_
.
- - - -
,
,
!
-
r
'
u<
25
l
!
<
l'
1--
. Containment Integrated Leak Rate Test'(CILRT)
'
l
Prior to the conduct of the CILRT, the; inspector held. discussions.
with licensee and NRC:RI. cognizant personnel'about the' upcoming. .
. test.
The inspector was informed by an NRC:RI. specialist inspector-
that since the licenseeL util_izes the mass point method for leakage .
calculations ~,.._the, licensee would need an exemption from the require-
'
- ments-of Appendix J to 10 CFR Part 50,: Paragraph III. A 3
This
paragraph requires the use of either point-to point or. total time
-
methods. A request for exemption to the aforementio'ned paragraph
was submitted-by the licensee'in their letter FYR 87-41 dated
April'17,.1987. The NRC issued the requested' exemption on
May 26, 1987.
I
On several occasions during May 4-7, 1987 the inspector observed
-l
. portions of the licensee's preparations a'nd conduct of the Type A'
'
containment integrated leakage rate test (CILRT), performed as
required by 10 CFR 50 Appendix J. . At 5:00 a.m. on May 6, 1987
licensee test personnel discovered significant, but non quantifiable,
' air leakage from the fuel chute dewatering system blank flange. The
containment pressure.at that time was 32.3 psig. At 7:00 a.m. that
morning the leaking flange was' inspected by plant technical personnel
and supervisors who determined that the leak rate appeared toLbe in
excess of technical specification limits. Although the leakage could
not be ' measured, they assumed the leakage to be large enough to fail
the Type A' test. 'At 8:58 a.m. the, failure was reported to the NRC~
duty' officer as:a 4-hour notification pursuant to 10 CFR 50.72. At
9:00 a.m. temporary repair of the blank flange was completed,- stop-
ping the leakage, and the Type A test was' resumed.
'Subsequ'ently, as a result of large swings in test data and indica-
tions of abnormal operation of the containment pressure monitoring
instrumentation, licensee personnel performed troubleshooting and-
i
discovered water in the containment pressure instrument line. After
correcting the problem, official test data collection began at 9:15
p.m. and was terminated at 9:28 p.m. on May 7, 1987.
The licensee's
initial assessment of test data indicated that the test results were
satisfactory, and the final calculations of test data will be sub-
mitted to the NRC in a written report following the outage.
The inspector reviewed the 1987 CILRT log of test activities and the
licensee's actions taken in response to the testing problems.
The
inspector considered the actions to be appropriate.
However, the
inspector questioned licensee management regarding the amount and
cause of leakage from the fuel chute dewatering system blank flange.
In response to these questions, as well as similar concerns expressed
by plant management, licensee quality control personnel prepared
special inspection document QCIR (quality control inspection report)
No.87-166, '.' Inspection Plan for MR #87-598," approved May 9,1987.
The purpose of that inspection plan was to provide inspection
instructions for removal of the blind flange from the fuel chute
_ _ _ _ _ _ - _
_ - _ _ __-
-
1
i
26
!
,
dewatering line,.in order to determine the probable cause of boundary
failure. The flange was inspected on May 12, 1987 and inspection
personnel then found that, contrary to the flange design that speci-
fied the use of two concentric resilient 0-rings, no 0-rings were
installed.
I
As a result of this discovery, the inspector questioned licensee
' personnel regarding the flange installation during the last
refueling outage in November 1985, and the inspector reviewed the
following documents relevant to this matter.
l.
(1) Plant design changt: request (PDCR) 85-09 " Modification for Type
C Testing"; job order completed December 9, 1985.
-(2) Procedure OP-5100, Revision 12, dated May 1985, " Valve, Fitting
or Pipe Section Replacement, Installation and/or Repair - Job
No. PDCR 85-009, Installation of Testable Flange on Fuel Chute
Pumpback line at CS-V-679;
(3) Procedure OP-4702, " Vapor Container Type B&C Penetration Tests"
Revision 11, Attachment E.
Component- Fuel Chute Pumpback
Blank Flange, completed November 24, 1985.
(4) Yankee Atomic Electric Company (YAEC) letter to USNRC dated
January 8, 1986 (FYR 86-004), " Replacement of a Manual
.
Containment Isolation Valve with a Blank Flange in Table
'
3.6-1 of the Technical Specifications", Proposed Change
No. 198.
(5) USNRC letter to YAEC dated June 9, 1986 (NYR 86-128), " Changes
to Technical Specifications", Amendment No. 96 to Facility
Operating License No. DPR-3.
In response to the apparent containment Type A test failure and
the licensee's disco; ry on May 12, 1987 that no 0-rings were
installed in the bl- ( flange, by memorandum dated May 21, 1987,
licensee corporate management directed the manager, plant engi-
i
neering department to investigate the factors leading to, and the
resolution of this event.
That memorandum specified the manner for
performing the investigation, including:
1) determine the underlying
factors leading to the incident; 2) recommend to plant and corporate
{
management short term corrective actions; 3) recommend long term
corrective actions to preclude recurrence of this incident and
incidents of a similar nature; and 4) a test will be conducted in
order to duplicate the original conditions to determine a leak
rate.
The ensuing investigation was conducted in an expeditious
i
and aggressive manner, and the findings and recommendations of the
!
investigation were detailed in a memorandum dated June 5,1987
1
- _ _ _ _ - . . _ _ .
-
_ _ _ _ _
.,
.
27
from the manager, plant engineering department to the vice presi-
dent and manager of operations.
Both the licensee's investigation *
l
and the inspector's independent review identified the following
information about the blank flange 0-ring problem.
(1) Plant design change request (PDCR) 85-09 included modifications
i
to the fuel chute dewatering system and the vapor container
(VC) service air system to reduce leakage and improve the
results of Type C testing.
The modification of the fuel
chute dewatering system consisted of removing a flanged valve
and machining the flange on the pipe to allow installation of
a blank flange with testable 0-rings.
The blank flange would ..
remain in place during plant operation and become a containment
isolation system barrier.
During refueling operations, the
blank flange would be removed and the valve reinstalled.
(2) The PDCR and Job Order called for the use of a generic
installation procedure, OP-5100.
The testing procedure was
written as Attachment E to OP-4702.
(3) The modification was installed and tested during the refueling
outage in November 1985.
(4) The leakage from the fuel chute dewatering line blank flange,
as detected and assessed on May 6, 1987 while performing the
Type A CILRT, was assumed, but not verified, to be large enough
to preclude passing the Type A test.
(5) As noted above, subsequent investigation revealed that the
0-rings had not been installed with the blank flange as
intended by PDCR 85-009.
(6) The licensee's investigation identified the specific detailed
chain of events that contributed to installation of the flange
without the 0-rings.
By YAEC letter dated July 11, 1987, the
licensee submitted Licensee Event Report (LER) 50-29/87-11,
" Type B&C Test Combined Leakage Exceeded Technical
Specifications." The LER included a discussion of the blank
flange leakage, including a summary of the sequence of events
resulting in improper flange installation.
Therefore, the
specific chain of events is not detailed further in this
report.
In summary, the licensee's investigation into the
blank flange 0 ring problem showed that the major cause of the
- The findings and recommendations of the licensee's investigation, as
described in the June 5, 1987 memorandum, are restated in the paragraphs that
follow.
The investigation findings were substantiated by the inspector's
independent review and discussions.
..
.
..
,_ -
r
.
- .
>
,
~28
1
1
. failure to install-the 0-rings was a breakdown in the transfer
~ '
s f the" details of the blank flange sealing function in the'
o
-
design package to the personnel performing the installation'.-
l Another issue identified in .the licensee's investigation was
the apparent failure in the design. process to. identify.the need
1
-to evaluate the' support for the pipe after.the valve'is removed
'l
H
and the blank-flange-is installed. 'Although this problem was.
'
!
. identified prior to the installation, it should have been
-
l
Levaluated during the design or design: review process.
(7) ~This incident occurred'during the 1985 refueling outage, in the
same time frame as another incident (improper overload' devices); .
which involved similar deficiencies related to the adequacy of
.
installation and test p.rocedures, manpower adequacy and' design
l
- control adequacy. A licensee Task Force performed a detailed ;
review of'the other incident and made' recommendations to address
'
deficiencies in these areas, as described in a memorandum from'
B. L. Drawbridge to L. H. Heider, " Implementation Matrix;of Task
-
>
~ Force Recommendations-Status", dated March 6, 1987. The licen-
see's inv~estigation determined that a-number of the resolutions 1
~
,
for these' deficiencies are equally applicable to the 0-ring:
!
problem. The' overload device incident is addressed.in previous
NRC Region I inspection report No. 50-029/86-09, and was the
subject of-an enfo'rcement conference held on July 22, 1986.
(8) -Lastly,;the licensee's investigation identified the specific
underlying factors leading'to the incident and recommended
-- corrective actions.
The various causes of the event are also
described.in LER 50-29/87-11, in which the corrective actions
are modified and expanded to prevent recurrence of design change
installation errors.
!
Subsequent to and independent of the licensee investigation of the
-flange 0-ring problem and also following fuel transfer activities,
the blank flange was tested with no resilient seals.in place by
j
internal pressurization of the line, which was essentially equiva-
D
lent to the test conditions of the Type A test.
During that test
!
the system yielded a leak rate of 0.048 wt. %/24 hours, which was.a
!
significant contributor to the combined leakage rate of Type B and C
penetrations testing exceeding the Technical Specifications limit,
as discussed in LER 50-29/87-11.
However, this amount of leakage
.;
was not considered sufficient to result in failure of the Type A
'
test. A follow-up test performed with resilient seals installed
produced zero leakage, which constituted a verification of the
validity of the flange design.
The licensee's investigation and the inspector's independent review
of this incorrect flange 0-ring installation identified apparent
violations of the licensee's implementing procedures for assuring
quality implementation of design changes to safety-related struc-
tures, system and components.
However, both the Type A and the
w-
_
_ - - _
- - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _
--__-
- _ - _ _ - _ _ - _ _
.
.
29
maximum Type 8 and C combined leakage rates were determined to be
within the technical specification limits for containment integrity.
Also, the installation error occurred during the same time frame as
a problem (improper overload devices) that resulted from similar
inadequacies. As noted above, the licensee and NRC actions in
regards to that problem are described in the previously referenced
NRC inspection report.
In addition, in response to the 0 ring
installation problem, the licensee plant and corporate management
(
l
initiated prompt follow-up review and investigation and identified
additional substantive corrective actions to prevent recurrence.
Therefore, no separate noncompliance are being cited in a Notice of
Violation with regard to his matter.
However, licensee completion
of the corrective actions described in LER 50-29/87-11 is considt J
an open item, pending NRC verification of the completed actions and
effectiveness of licensee implementation of design change installa-
tion procedure improvements during a subsequent NRC inspection
(Inspector Follow Item 50-29/87-06-03).
Core XIX Startup Test Program
--
Comments pertaining to inspector observation are contained in
Section 7 of this report.
Steam Generator (SG) Inservice Inspection
--
During the inspection period, the licensee contracted for SG
l
inservice inspection services.
Eddy current testing of SG No. 2
was scheduled, and was to consist of examination of the hot and
cold leg tubes. The inspector periodically reviewed the operations
l
conducted per OP-7203 to verify the results for TS 4.4.10 compliance.
'
The licensee's testing program this outage was designed to resolve
the issue of SG No. 2 magnetite interference and testability of the
hot leg tubes, as document in Inspection Report 50-29/85-18.
As a result of the performance of testing activities on SG No. 2,
the licensee discovered that not all previous indications of tube
defects had been identified during the 1984 examinations of SG Nos.
2, 3 and 4.
This occurrence is documented on the licensee's July
2, 1987 submittal of LER 50-29/87-06.
The NRC's review of the licensee's activities in this area by an
NRC:RI region based specialist inspector will be documented in
Inspection Report 50-29/87-07.
During the specialist inspection, which covered the period of May
11-29, 1987, additional information was requested to help clarify
issues raised as a result of the licensee's testing experience.
On
June 15, 1987, the licensee submitted its letter FYR 87-65 to NRC:RI
that provided the requested clarifications, and stipulated a commit-
ment to adopt a primary-to-secondary leakage limit during Cycle 19 of
500 gallons per day through any one steam generator not isolated from
~
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_
.
-
.
4 R c.:
n
L
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30r
. .
- ) .
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~
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,
.. ..
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.,
.
c
'.the'~ main- coolant system (MCS). - This . limit is. in . addit' ion)to. the ~
- existing TS 'limitsof 1_. gpm through'all' steam generators not isolated-
from the MCS. The inspector'.verifled that procedure OP-9105,lRev. 6,
j
' Detection and' calculation' of primary to secondary leakage,, was revised
a
in accordance'with'the licensee commitment, -The inspector.had no:
l
,<
further. questions of the. licensee pertaining to activities 'in this
'
,
area.
-- -
Testing ~of'the 125'VDC Station Batteries
As a~ result of performing procedure OP-4272, Rev. 2,'on.May 29,.
'
1987 the-licensee declared the No. 3 station battery service' test
ca failure.
The test is performed to satisfy the 18-month TS- 4.8.2.3.2.d requirement which.provides verification that the battery.
capacity is adequate to supply and maintain in an' operable status
all of the actual emergency loads.
The. test used this refueling .
Loutage' incorporated a load profile that was higher than past tests.
The basis for the new profile was contained in licensee memorandum _
YRP 487/87 dated May 22, 1987.
No indications of weak cells or. weak
connections were identified following the performance of the. test.
~
In response _-to.the test results, the licensee's YNSD cognizant
'
projects engineering personnel conducted an evaluation,.as documented
in a June l1, 1987 memorandum YRP 507/87.
This evaluation recommended
--that: 1) a performance discharge test be conducted to verify that
_
battery capacity-is still 80%'or greater and 2) conduct a service
test with a revised load profile that results in a reduction of 200
amperes for the first minute.
This reduction is accomplished by
administratively-establishing operating guidelines that result-in the
emergency turning gear oil pump (ETGOP) being started and .run follow-
ing a-turbine trip on shutdown and when rolling the turbine until the
turbine is up to 1800 rpm.
This action would preclude ETGOP starting
current from overlapping the No. 3 EDG-inrush currents during a' loss
of- ac. power.
The inspector noted that the load profile maintains the
inrush current constant for the first minutes, which represents a
very conservative approach and is in accordance with IEEE standards.
The licensee accomplished the performance discharge test in accord-
ance with procedure OP-4519 on June 3, 1987, which demonstrated that
A
Battery No. 3 has 83.3 percent capacity and is in excess of the 80%
required by TS 4.8.2.3.2.e.
Subsequently on June 8, 1987, a service
test utilizing the reduced load profile was successfully performed in
accordance with procedure OP-4572, Rev. 3.
Following the collection
of all test data, the licensee's engineering staff conducted a review
at the battery vendor (C&D) facilities on June 18, 1987 to discuss
.
the No. 3 battery capability and projected life. All relevant
'
'
battery performance tests between 1977 and 1987 were utilized by the
vendor's evaluation.
The vendor's evaluation indicates that Battery
No. 3 has a 90% capacity and that it now has a 40% design margin over
the design load profile currently used.
2-- _ _ _ - _ _ - - __ _
_
_-_ -
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,
31
The licensee's 1985 capacity determination for the No. 3 battery
was found to be 95.1%.
Because the 1987 test determined an 83.3%
capacity, the inspector was concerned that the rate of. decrease
could result in substandard performance of the battery sometime
during the next 18 month operating cycle.
This resulted in the
inspector reviewing the test results with cognizant specialist
inspectors at MRL:RI Division of Reactor Safety on June 16, 1987.
On June 18, 1987 the inspector and NRC:RI personnel discussed with
the licensee their test program and planned corrective actions.
.During this discussion, the licensee indicated that they plan on
replacing the existing No. 3 battery.
It is the licensee's intention
to be ready to place the new battery in service sometime in the next
cycle.
They indicated to the NRC that it would require a cold shut-
down of approximately two weeks to perform final installation and
testing.
On June 25, 1987, the licensee's YNSD projects group issued memor-
andum YRP 612/87. This document specifies that the testing on
,
No. 3 Battery done during the refueling outage and previous outages,
i
when evaluated by the manufacturer, demonstrates that it has the
required capability to meet emergency loads for Cycle 19.
During
the conduct of the procedural reviews the inspector noted the
following areas of concern or discrepancies:
(1) The method contained in procedure OP-4519 to calculate the
battery capacity is inconsistent with the stipulated method in
IEEE-450, 1975 Edition, for a two hour test.
It appears that
!
the formulas stipulated in the procedure is usable only for
Nos. 1&2 batteries, since they have only a one hour test. YNSD
engineers on site to review the discharge test performed
independent calculations, did not use or review the station
generated procedure, and therefore did not identify the
]
procedural inadequacy.
Additionally, the procedure does
'
not show all calculational work, which minimizes the
effectiveness of supervisory review.
1
(2) The licensee has never committed to comply with the IEEE
Standard 450-1975, Recommended Practice for Maintenance,
Testing and Replacement of Large Lead Storage Batteries for
Generating Stations and Substations.
Battery service testing
l
in accordance with IEEE Standard 450-1975 is endorsed by
The NRC's Inspection and Enforcement
'
Manual, Part 9900-Technical Guidance, specifies that battery
testing conform to IEEE Standard 450-1975.
Since the licensee
was unaware of this guidance, the inspector provided them with
j
its contents.
!
As a result of discussion with representatives of the licensee's
maintenance organization, the licensee indicated that they would
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revise procedure OP-4519 to address the inspector's concerns, and
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evaluate battery performance and service testing procedures to
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ascertain their conformance to the NRC's guidance.
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Other than the deficiencies noted above,-the inspector had no
.i
,
further questions on this item.
'
10 .' Onsite Review Committee Activities
On May 5, 8, 11, 14, 15, 22, and June 5, 9, 16, 18, 25, 26, 30, and July
2,1987 the inspector observed the meetings of the Yankee'NPS onsite
review committee to ascertain that the provisions of TS 6.5.1 were met.
As.a result of reviewing the licensee's activities in this area, no
violations were identified.
However, one area that required plant
management attention to resolve inspector concerns involved the
licensee's processing of procedure. change notices (PCNs).
Additional
I
management attention was determined to be warranted to insure that when
1
the PCN were used to revise Plant Procedures, the requirements estab-
!
lished in procedure AP-0001, Plant Procedures, would be adhered to.
Specifically, the PORC needed to ensure that if the PCN resulted in a
i
change to'the intent of the procedure there would be a pre-implementation
PORC review. Throughout the inspection period the inspector observed
improvements in PORC's attention to detail on this matter, and in this
regard is reflective of the licensee's cooperative spirit usually
demonstrated in resolving inspector concerns.
The inspector had no
further questions of the licensee on this matter.
,
11-
Licensee Response To IE Bulletins
.
The licensee's response to the following IE Bulletin (IEB) was reviewed.
This review included: adequacy of the response to IEB requirements,
timeliness of the response, completion of identified corrective actions
and' timeliness of completion.
IE Bulletin No. 84-03:
Refueling Cavity Water Seal, dated August 24,
1984.
This bulletin notifies licensees of an incident at another
a
facility in which the refueling cavity water seal failed and rapidly
drained the refueling cavity, and requested certain actions to assure
.
that fuel uncovering during refueling remains an unlikely event.
The
i
licensee was to evaluate the potential for and consequences of a refuel-
,
ing cavity water seal failure and provide a summary report of these
i
actions prior to beginning refueling or within 90 days of receipt of the
bulletin. The evaluation was to include consideration of gross seal
failure; maximum leak rate due to failure of active components; makeup
,
capacity and emergency operating procedures. This issue was reviewed in
i
Inspection Report 50-29/85-02, Section 10. As a result of the inspector's
review and comments on this issue, the bulletin remained open until the
licensee's on going analysis was completed and reviewed.
l
The inspector reviewed licensee letter FYR 84-114 to NRC:RI dated
i
December 5, 1984 and licensee letter FYR 85-83 to NRC:RI dated August
!
6, 1985, using the guidance provided in Temporary Instruction 2515/66.
,
Additionally, in the licensee letter dated August 6, 1985, the licensee
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33
detailed changes that would be accomplished to'further reduce the poten-
tial of a loss of shield tank cavity water.
The inspector reviewed these
commitments to ensure they had been completed.
The following reports and
procedures were reviewed:
--
Impell Test Report " Experimental Verification of the Presray Reactor
Cavity Liner Seal for the Yankee Nuclear Power Station" dated July,
1985
Memo from B. L. Drawbridge to N. N. St. Laurent dated December 4,
--
1984 " Review of Connecticut Yankee Cavity Seal Failure"
--
Memo from B. L. Drawbridge to N. N. St. Laurent dated December 27,
1984, " Notes of INP0 Workshop on the Connecticut Yankee Shield Tank
Cavity Seal Failure"
Memo from B. L. Drawbridge to J. A. Kay dated January 28, 1986
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" Summation and closecut of SOER 85-01, " Reactor Cavity Seal Failure"
OP-1100, Rev. 10, Dismantling and Reassembly of the Reactor Systems
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For Core XIX Refueling
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OP-1203, Rev. 13, Filling of the Shield Tank Cavity (STC)
--
OP-1209, Rev.11, Operation of the Vapor Container (VC) Fuel
Handling Equipment
OP-1214, Rev. 13. Component Movement within the New Fuel Vault
--
(NFV) and Spent Fuel Pit (SFP)
--
OP-1524, Rev. 1, Reactor Cavity Seal Ring Installation and Removal
OP-1700, Rev. 13, Cycle XIX Reactor Refueling and Component
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Inspection
OP-3117, Rev. 12, Refueling Accidents
--
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OP-1507, Rev. 10, Reactor Head - Removal, Handling and Storage
.
Special Orders 42 and 44, dated April 24, 1987 and May 1, 1987
--
respectively
--
Rowe Station Shutdown Log
The licensee performed a detailed evaluation of the potential for loss of
refueling water in the shield tank cavity.
This evaluation included the
cavity seal ring, shield tank cavity, cavity penetrations, operating
procedures and failure consequences. Also, during a subsequent refueling
outage, the licensee performed a dimensional verification of the cavity
seal annulus to confirm that the annulus was within the dimensional
parameters which were modeled during the Impell hydrostatic test program.
A new procedure (0P-1524) was issued to detail annulus cleaning instruc-
tions and OP-3117 was upgraded to include the essential actions to be
taken by refueling personnel in the event of a loss of shield tank cavity
water.
These documents and the licensee's response conform to the bul-
letin's requirements and satisfactorily addressed inspector concerns
enumerated during the prior inspection on this issue.
The inspector also
verified that the changes detailed by the licensee have been completed.
This bulletin is closed.
- - _ _ _ _ _ _ _ _ _ - - _ _ _ -
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"
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12 .-
Followup'on Fitness for Duty Corrective Actions
Inspection Report 50-29/86-08, Section 16,~ describes the inspector's
review of..an'i.ncident that involved the use of'a' medically prescribed drug
that' allegedly. impacted _on the ability of an armed security officer to
~
- perform required duties.
Inspector-identified deficiencies'that warranted
corrective actions were 1) a lack of' written policies and procedures' per-
taining to the use of medication,by security officers and 2) failure to.
provide to the security officers appropriate employee indoctrination or
training.
l0n' September 23, 1986 the security contractor, Green Mountain Security.
,
_
Services (GMSS), . issued a written' policy statement' pertaining to the use
of drugs and alcohol'. This policy statement, which includes fitness for-
' duty concern. associated with the use of legal medications by security.
officers, is signed by all present and future GMSS employees at the
. Yankee Nuclear Power Station that'they have read, understoodf and will
comply with the provisions of the policy. Additionally, on this date the
~
- licensee's Drug and Alcohol Policy was provided to all security officers.
During'a. December, 1986 GMSS training activity, all supervisors were
issued a copy of GMSS' Supervisor Guidelines for Adminisi.ering the
.
Company' Policy on Drug and Alcohol Abuse.
-
The inspector had no further questions of the licensee on this matter.
13.
Licensee Action'on NUREG-0660, NRC Action Plan Developed as a Result of
j
the TMI-2 Accident
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The NRC!s Region I' office has inspection responsibility for selected
-action plan items. These items have been broken down into numbered
descriptions (enclosure 1 to NUREG-0737, Clarification of TMI Action
- Plant _ Items).
Licensee letters containing commitments to the NRC were
used as the basis for acceptability, along with the NRC clarification
letters and inspector judgment. 'The following items were reviewed.
q
.!
NUREG-0737, I.C.1.3.B, Revise Procedures for Transients & Accidents
--
This item required the licensee to reanalyze transients and
i
accidents and prepare Technical Guidelines, with the associated
l
requirements specified in Supplement 1 to NUREG-0737 (Generic Letter
'
No. 82-33).
The status of this item was last updated in Inspection
- l
Report 50-29/87-02, and reflected the open status of the item due to
1) NRC concerns associated with allowing the licensee to implement
'
,
the' emergency operating procedures (E0Ps) at the start of cycle XIX
4
(June 1987) and 2) the need for further NRC:NRR staff review and
'
actions.
8
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Subsequent.ly, the ' licensee in fits' Mayy5,1987. letter (FYR 87-36)
f
requested that NRC:NRR-extend'the completion date'for this item
.
,
I
from prior-to startup of cycle'19 operation, which is specified in.
il
the Commission Order dated July 5,1985, until 'one year. after NRC
approval of the licensee's Procedure. Generation Package. . In' ~ the '
!"
June 4, 1987 NRC:NRR letter to the licensee, the Commission's Order
-
was modifled.to allow delay.of the: implementation-of the E0Ps until
i
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prior to.startup of Cycle XX Operation (late fall.1988). This~1 tem
remains:open.
14. Organizational Changes
'
'
During the. inspection period,~ the inspectorireviewed changes to the
!
licensee's staff or. organization. structure as described below. The-
review included verification that licensee's onsite organization
._
structure is as' described in the facility TS,;and verification'that
'
-personnel:qualifi. cation levels are in conformance with ANSI'N18.1-1971,
!
as described in TS Section 6.3.1.
On July 8,1987, _theilicensee announced that the security supervisor will
report .toLthe administrative service's manager, and the reactor engineering-
'
manager will report to the assistant ' technical director. Both' changes
were effective'on July;15, 1987.
The inspector noted that~the technical
' director's: organizational structure was last. changed on December 29,.1986,
which in part created the _new position of assistant technical ' director.
However,'.during that' organizational change the reactor engi'neering manager-
continued to. report to the technical director.
Because~of on going con-
- i
cerns in the area of initial' operator licensing and licensed operator--
requalification programs, the licensee's change affecting the reactor
engineering manager allows the technical director who has the training
department manager reporting to him, to focus additional management
attention in the development and implementation of corrective actions
to address the concerns:in the training area.
The licensee had submitted to the NRC on June 10, 1987 its Prop'osed
l
Change (PC) No. 203 to the plant's TS that reflected station organization
' changes. As'a result of the changes discussed above, the proposed change
is not reflective of the organizational structure currently in place.
The
inspector discussed the recent changes and their impact on the PC with the
licensee's senior licensing engineer, who acknowledged inspector comments
a'd concerns, and indicated that they will consider the issuance of a
.
n
supplement to PC No. 203,
1
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No deficiencies were identified.
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5
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15. ' Onsite Meeting To Discus's Operator. Licensing Issues -
.
I
At.the. request of-NRC: Region I, an informational meeting was conducted
!onsite.on,Junel3, 1987 between representatives of the Division of Reactor.
' Projects and'the licensee's management and training organization. 1This-
~
-meeting _was: conducted to ensure-that the< licensee fully' understood the
~
.
. issues raised by the NRC in.the Requalification" Programs Evaluation
Report-No. 50-29/86-20 (0L) and Examination Report No.: 50-29/87.-04 (0L).
Discussions had been held prior to the. meeting.between the licensee's-
c training staff and the NRC's examination representatives to only. resolve
misunderstandings or. quest. ions concerning the findings' contained within
1the examination reports.
Licensee' personnel indicated that-they con-
curred with the: findings and=that they had been identified' internally
prior to the conduct of the NRC examinations.
Further,.they' indicated
that. resources were being devoted to improving the requalification
-program and' acknowledged their-responsibility for_the apparent inade-
i
<
.quate training of SRO upgrade candidates. 'The licensee has developed
l
a program.for tracking and. addressing weaknesses of individuals'who are.
'
.
in al training program.
l
Discussions were-conducted on the use of learning objectives as a means
of" defining the ' knowledge and performance levels required by operators; to
safely operate:,the facility.
The importance of'a simulator as a training
and evaluation. tool was-discussed. '
1
Concern was raised by the licensee ever the additional stress-on the
' licensed operators produced by the NRC direct involvement'in the requal-
1
Eification program' annual examinations. lThe NRC responded that the goal-
i
of ensuring .that licensed operators were sufficiently' trained to safely.
j
operate a facility was the responsibility of both the . licensee and the-
NRC and the' continual interchange between organization's was needed to
help licensed personnel to. understand the purpose of the requalification
examinations.
16. Ma'nagement Meetings
During the inspection period, the following management meetings were
1
conducted or attended by the inspector as noted below:
The inspector attended an exit meeting held on April 29, 1987 by
--
a region based operator licensing examiner at the conclusion of
Operator Licensing Examination 50-29/87-04 to discuss results and
identify strengths and weaknesses in the licensee's operator
licensing training program.
The inspector attended an exit meeting on May 14, 1987, by a region
--
based specialist at the conclusion of reviewing the plant activities
portion of Inspection 50-29/87-07, review of the licensee's steam
]
generator in-service inspection progra;n.
1
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