IR 05000369/1989024
| ML19325D252 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 10/05/1989 |
| From: | Cooper T, Shymlock M, Vandoorn K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19325D250 | List: |
| References | |
| 50-369-89-24, 50-370-89-24, GL-88-17, NUDOCS 8910200003 | |
| Download: ML19325D252 (21) | |
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[p* #40 9'o NUCLEAR CEGULATC RY COMMisslON :
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ATLANTA GIORGIA 30323 j
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Report Nos.: 50-369/89-24 and 50-370/89-24 l
Licensee: Duke Power Company
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422 South Church Street
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Charlotte, NC 28242 i
Facility Name: McGuire Nuclear Station Units 1 and 2
Docket Nos.': 50-369 and 50-370 l
License Nos.: NPF-9 and NPF-17 i
Inspection Conducted: July 22, 1989 - September 12, 1989
Inspectors:
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.er 6 d E /fW
K. VanDoorn, priorResip.tInspector Date Signed
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W$h11&t}&$* A hef.Wl'?$9
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T. Cooper, Resjeent Inspeglor Date Signed
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Approved by:
N 4/mt/ M d O/ S' /90 M. B. Shym'~ocK)Section CMeT Date '
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Reactor Projekts Section 3A I
Division of Reactor Projects
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SUPMARY h
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Scope:
This routine unannounced inspection involved the areas of operations safety verification, surveillance testing, maintenance activities, and follow-up on previous inspection findings, followup of Licensee i
l Event Reports, followup of 10 CFR Fart.21 reports review-of modifications program, and evaluation of~ licensee self-assessment capability.
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Results:
In the areas inspected, two violations, one. unresolved item and one weakness were identifi.ed, j
i One violation involved inadequacies in documentation of minimum shift staffing and inadequacies in shift duty assignments (paragraph 3.).
A second apparent violation involvesiinadequacies in design control
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of the Control Room and Annulus Ventilation Systems leading to the i
systems being inoperable under certain conditions (paragraph 9).
The cecond apparent violation is being considered for escalated j!
enforcement.
These ventilation problems and a related problem with '
the Hydrogen Skimmer System (Report 369,370/88-24) indicate an j
apparent weakness in the licensee design control for ventilation j
systems.
One unre,olved item was identified involving access to
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Control P,oom (paragraph 3.d.).
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rep 0RT DETAILS
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Persons Contacted i
licensee Employees G. Addis, Superintendent of Station Services
'D. Adkins, Security Specialitt i
- D. Baxter, Support Operations Manager
- J. Boyle, Superintendent of Integrated Scheduling
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D. Bumgardner, Unit 1 Operations Manager
- B. Dolan Design Supervisor l
J. Foster, Station Health Physicist
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M. Funderburke, Station Chemist t
- G. Gilbert. Superintendent of Technical Services
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- C, Hendrix, Maintenance Engineering Services Manager
- T. Mathews, Site Design Engineering Manager l
- T. McConnell, Plant Manager
- D. Murdock, McGuire Design Engineering Division Manager
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W. Reeside Operations Engineer
R. Rider, Mechanical Maintenance Engineer
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- M. Sample, Superintendent of Maintenance i
- R. Sharp Compliance Manager
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- J. Snyder Performance Engineer J. Silver, Unit 2 Operations Manager i
A. Sipe. McGuire Safety Review Group Chairman
- B. Travis, Superintendent of Operations f
- R. Weidler, Design Supervisor (
R. White, IAE Engineer l
r Other licensee employees contacted included construction craftsmen,
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technicians, operators, mechanics, security force members, and office personnel, j
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- Attended exit interview 2.
Unresolved Items An unresolved item (UNR) is a matter about which more information is required to determine whether it is acceptable or may involve a violation
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or deviation. An unresolved item is identified in paragraph 3.d.
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PlantOperations(71707,71710)
The inspection staff reviewed plant oper3tions during the report period to i
verify conformance with applicable regtlatory requirements. Control room
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logs, shift supervisors' logs, shif t turnover records and equipment
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removal and restoration records were routinely perused. Interviews were conducted with plant operations, maintenance, chemistry, health physics, and performance personnel, j
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Activities within the control reorr were monitored during shifts and at
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shif t changes. Actions and/or activities observed were conducted as prescribed in applicable station administrative directives.
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Staffing Requirements
Licensee Technical Specification Table 6.2-1 reouires a minimum shift crew, when either unit is in Modes 1 through 4. of one Shift Supervisor, one Senior Reactor Operator three Reactor Operators.
three Auxiliary Operators, and one Shift Technical Advisor Table
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B-1 of the ifcensee Emergency Plan requires that when in the Emergency Plan, the minimum staff requirements for the affected unit are one Shift Supervisor, one Assistant Shift Supervisor (SRO). two Control Room Operators, and two Auxiliary Operators.
NUREG 0654, i
Revision 1. Criteria for Preparation and Evaluation of Radiological l
Emergency Response Plans And Preparedness in Support of Nuclear Power i
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Plants requires that in addition to the minimum staff required for the affected unit, each unaffected unit in operation shall have at L
least one Shift Foreman. One Control Room Operator, and one Auxiliary l
Operator (except that units sharing )a control room may share a Shift Foreman if all functions are covered.
These requirements delineate the fact that the Technical Specifi-cation minimum staffing requirements constitute the minimum staff necessary to assure safe shutdown of the two units in an emergency situation.
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Selected Licensee Consnitment (SLIC) 16.13-1. Fire Brigade requires:
A site Fire Brigade of et least five members shall be maintained
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onsite at all times.
The Fire Brigade shall not include three
members of the minimum shift crew necessary for safe shutdown of
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the unit and any personnel required for other essential functions during a fire emergency.
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Station Directive 2.11.1. Fire Brigade Organization and Program, states that the primary Fire Brigade will consist of one Reactor
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Operator / Senior Reactor Operator to function of the Fire Brigade Leader and four Non-licensed Operators to function as brigade members.
f While reviewing the licensee ;.ocedures for verifying that the
minimum staffing requirements were being met. the inspector noted i
several discrepancies in the content and performance of these
procedures and their adherence to the regulatory requirements.
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following discrepancies were noted by the inspector during his review I
of June through August shift documentation.
Attachment la of Operations Management Procedure (OMP) 2-2 l
Shift Turnover, is meant to be cumpleted to verify that the minimum Technical Specification requirements for control room i
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shift manning were being met. On at least four occasions during July and August 1989 the Shift Supervisor was signed as meeting both the Shift Supervisor and the Senior Reactor Operator positions.
From this documentation review. no objective
evidence exists that the Technical Specification minimum
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staffing requirements were being met.
I On one case in June 1989. the attachment verifying the minimum
shift crew composition was not completed.
No objective evidence exists that the minimum staffing requirements were being met
during that shift.
On attachment la OMP 2-2. Where the assignments are logged for
i Fire Brigade members. the coninon practice during July and
August 1989, has been to assign five non-licensed operators.
contrary to Station Directive 2.11.1. which requires four-
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non-licensed operators and one licensed reactor operator or
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senior reactor operator.
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On multiple instances during July and August 1989, either the Senior Reactor Operator or one of the three Reactor Operators or l
one of the three Auxiliary Operators, that were part of the
minimum staff. were designated as members of the Fire Brigade, i
On several occasions, several of the minimum control room staff i
personnel were simultaneously designated as members of the Fire l
Brigade.
During these occasions, no objective evidence exists that the requirements for both minimum staffing and Fire Brigade were being met, j
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On several occasions during June and July 1989, no members of
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th? operating staff were designated as Fire Brigade members on attachment la of OMP 2-2.
For these instances, no objective I
evidence exists that the Fire Brigade was being manned, as
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required by licensee procedures and SLIC 16.13.
These examplu together demonstrate a failure to provide documenta-i tion that the Technical Specification requirements for minimum shift i
staffing and the SLIC requirements for Fire Brigade manning were
being met.
Licensee procedure OMP 2-2. requires that each shift perform this verification on Attachment la, of OMP 2-2.
This failure to follow procedure constitutes an example of violation i
369.370/89-24-01. Failure to Adequately Document Shift Staffing -
Multiple Examples.
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These occurrences indicate a lack evidence a general lack of familiarity with the different procedures that delineate the
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requirements for shift manning verification and fire brigade manning requirements. The Fire Brigade assignments usually consisted of five
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non-licensed operators instead of the one licensed R0/SR0 and four
non-licensed operators, as required by the licensee proccdures. The
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minimum shift crew composition checklist was routinely used to verify
who was covering which position i.e. who the SR0 in the control room
was (even if it was the Shift Supervisor), and was not being used to
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verify that minimum staffing requirements were being met, i
The occurrences also appear to demonstrate a lack of coordination i
between the Emergency Plan (and associated liUREG). Technical
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l Specifications, and the Selected License Commitments when the various procedures were written dealing with manning requirements.
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These examples demonstrate an inadequacy with OliP 2-2.
The procedure
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allows personnel to be assigned to the Fire Brigade who are members
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This is contrary to the requirements of
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the Technical Specifications, the SLIC. and NUREG 0654 and
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constitutes an example of violation 369.370/89-24-01.
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Plant Tours
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Plant tours taken during the reporting period included, but were not l
limited to, the turbine buildings, the auxiliary building. Units 1 l
l and 2 electrical equipment rooms. Units 1 and 2 cable spreading rooms, and the station yard zone inside the protected area.
During the plant tours. ongoing activities, housekeeping. security.
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equipment status and radiation control practices were observed.
j The inspector performed inspections of various parts of the plant and
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noted housekeeping problems in several locations.
In all four diesel i
generator rooms, trash was noted in many locations, particularly under the floor gratings and under the solid covers over the floor cable runs, Gloves, pens, and paper trash were among the items
observed by the inspector.
At several locations throughout the
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Auxiliary Buildings. it appears that locations that are not easily
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accessible like areas where personnel must climb over equipment to reach accumulations of debris is occurring.
For example, on the 767 i
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foot elevation of the Auxiliary Building, ducting vents and dampers
removed as part of the recent VC/YC modification were left on the
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floor near where the work was performed.
Drop-lights were lef t a
sitting under the fans where this work took place for several weeks
following the completion of the work.
Scaffolding has been I
disassembled and randomly stacked at various locations throughout the
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Auxiliary Building. such as in the 2A NS Heat Exchanger Room and on the 750 foot elevation. by the WE pumps.
The licensee was informed of these items for action.
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During a tour of the Unit 2 lower containment, the inspector i
identified several discrepant conditions, which he brought to the i
attention of the licensee.
These items included examples of shielded l
cables where the armor shielding had been pushed back from the
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terminal connections. loose cable trays, crimped air supply lines to
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instrumentation, and loose cable connections on post accident
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I monitoring components.
These items were inspected and evaluated by
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i the licensee, who determined that no item was of significant impact.
Work requests were written to correct these items at the next
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available time.
The resident inspector reinspected the loose cable tray and determined that it' had been returned to satisfactory
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condition and examined the PAM instrumentation, along with
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documentation of preventive maintenance performed on it after the t
initial inspection and determined that no outstanding concerns i
existed in this area, c.
Plant Operation l
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Unit 1 Operations l
Unit I remained on line until a reactor trip occurred on i
August 26, 1989, at 9:34 a.m. due to a failed logic card which
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resulted in a low reactor coolant flow signal.
The unit was l
back on-line on August 29.
Technical Specification (TS) 3.0.3
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was entered on August 25. due to the Annulus Ventilation System i
(VE) being inoperable.
The licensee had failed to consider the i
l air column in the annulus relative to outside air temperature
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Discretionary enforcement was granted by NRC Headquarters and RII until 2:20 p.m. on August 26, 1989 to allow completion of design analysis and system retesting.
Retest was satisfactory and VE I
was declared operable at 10:50 a.m. on August 26.
An Unusual
l Event was declared on September 1. at 1:40 p.m. due to a
hydrazine spill.
Several persons had to be evacuated from two
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trailers. The spill was cleaned up and the event was terminated
at 4:30 p.m.
- Unit 2 Operations
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Unit 2 was in a refueling outage the entire period.
Extensive
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Eddy Current inspections of Steam Generator tubes and plugs were l
performed, the results of which were presented to NRC Head-
quarters and Ril during several phone conversations and during a
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meeting with NRC-in Washington on August 31. 1989.
All tubes were full length inspected.
No tube defects similar to the i
Unit I defect which caused a tube rupture in March 1989, were i
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found.
All hot leg tube plugs and a sample of cold leg plugs I
were tested.
Indications were found in the heel (lower) region of 58 plugs, primarily in heat number W592-1 which had i
previously been identified as being susceptible to stress i
corrosion cracking.
All W592-1 plugs were replaced along with
other plugs which showed indications.
The licensee confirmed
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through laboratory analysis of a sample of plugs that cracking had occurred in the heel region.
Further documentation of this inspection will be contained in the meeting minutes to be documented by NRC/NRR.
The unit was scheduled to enter Mode 4
on Septemt;er 1. 1989, however a Main Turbine bearing was severely damaged on August 31. due to a failure to remove a i
blank flange preventing lubrication' oil from reaching the bearing.
Corrosion was found on the containment vessel during
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inspection prior to Integrated Leak Rate Testing (ILRT)
(See paragraph 4).
On September 5. approximately 15.000 gallons of
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t Reactor Coolant and Refueling water Storage Tank water, were spilled into the Auxiliary Building.
Containment Spray System (NS) Train A was over pressurized during a valve stroke time test of an NS valve.
An NRC Augmented Inspection Team was dispatched on September 6. 1989.
Details of the event will be
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documented in Report 369.370/89-31.
On September 8. testing of
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Main Steam Isolation valves without air assist resulted in valves not meeting stroke time requirements.
Extensive re-analysis and re-testing was performed to verify operability
for both units (see paragraph 7.e.)
d.
Security
i While on a routine plant tour, the inspector noted a problem with
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access control of the Unit 2 Control Room.
Security personnel were notified, but the condition continued until the inspector notified
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the Operation's Shift Supervisor.
The licensee has taken action to prevent recurrence.
Details of the incident have been given to the
NRC Region II security group and will be tracked as Unresolved Item j
369.370/89-24-02: Control Room Access Control Problem.
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Generic Letter Review
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In response to Generic Letter 88-17. Loss of Decay Heat Removal, the
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. inspector reviewed the implementation of the items discussed in the
Generic Letter.
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To address the training requirements outlined in the Generic Letter.
the inspector interviewed selected licensed Reactor Operators and Senior Reactor Operators and determined that the licensed personnel
were familiar with the events' leading to the issuance of the Generic t
Letter and the licensee actions taken in, response to those events.
Training received by these personnel was determined to be adequate.
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During the review of the actions taken in response to the Generic Letter requirements for control of containment closure, the inspector reviewed licensee procedure OP/2/A/6100/23 Controlling Procedure for
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Containment Closure. The procedure established a containment closure
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coordinator. during outage situations, who is tasked with the
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tracking of breached penetrations and the assurance of coordinating
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the closure of these penetrations in an emerger.cy situation.
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inspector reviewed several penetration packages and interviewed l
selected closure cocrdinators and determined that in an emergency
situation, the licensee would have adequate assurance that all j
penetrations could be sealed in the required time frame.
The Generic Letter required that for mid-loop operation, multiple
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indications of core exit temperature and RCS water level must exist.
The inspector reviewed the instrumentation arrangement in the main control room and interviewed selected Reactor Operators.
Instrumentation necessary for the detection and mitigation of a loss of decay heat removal event has been provided, including multiple indications of level and temperature.
The operators demonstrated a thorough knowledge of the various instruments, the alarms associated with the instrumentation, and the procedures associated with the l
various instruments.
These instruments were monitored by control room operators on a routine basis.
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The inspector reviewed various procedures and documentation associated with mid-loop operation.
Operating procedures had been
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revised to provide direction for minimizing operations that could
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lead to perturbations in the reactor coolant system.
Procedures have
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been provided that list alternate means and flow paths for adding inventory when needed.
A Special Order was issued that assures adherence to these requirements.
Steps have been added to procedures
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which assure that all hot legs are not blocked simultaneously by
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nozzle dams, until the reactor vessel head has been removed, plant operations personnel have demonstrated to the inspector a thorough knowledge of all these procedures and their contents.
l The plant does not have loop stop valves.
Therefore, the require-ments associated with them do not apply.
The licensee has addressed the requirements of the Generic Letter 1r.
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l their response and the inspector verified that the' items covered in i
that response have been adequats!1y implemented, with no adverse L
findings.
The inspector verified that all requirements of the l
Generic letter have been implemented.
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The inspector identified that the working copies, in the main control room, of licensee procedure. AP/1/A/5500/19. Loss of Residual Heat
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Removal, were of an earlier revision than the controlled copy. After
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I notifying licensee personnel of this, a review of all abnormal
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I procedures (APs) w35 performed and this was determined to be an
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isolated incident.
i No violations or deviations were identified.
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4.
Surveillance Testing (61726)
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Selected surveillance tests were analyzed and/or witnessed by the l
inspector to ascertain procedural and performance adequacy and
conformance with applicable Technical Specifications.
Selected tests were witnessed or reviewed to ascertain that current
written approved procedures were available and in use. that test i
equipmerit in use was calibrated, that test prerequisites were met.
that system restoration was completed and acceptance criteria were met.
Detailed below are selected tests which were either reviewed or
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witnessed:
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PROCEDURE EQUIPMENT / TEST
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PT/0/A/4350/28B 125 Volt Vital Battery Quarterly Inspection f
(completed procedures for EVCA. EVCB. EVCC l
and EVCD)
PT/0/A/4450/08C Control Room Area Ventilation Performance Test PT/2/A/4200/09A ESF Actuation Periodic Test (Sections 12.3 and 12.4)
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f PT/1/A/4450/03C Annulus Ventilation Performance Test PT/1/A/4250/04 PreStartupTurbineTesting(Section12.3)
Also NRC/ Region 11 personnel witnessed local Leak Rate testing during
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this period (See Report 369.370/89-20).
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Containment vessel corrosion was identified-by the licensee during i
visual inspection prior to Integrated leak Rate Testing.
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corrosion was located where the annulus concrete floor meets the (
outside containment wall:
Some corrosion appeared to have occurred
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for the entire containment circumference.
Approximately one-half of r
the area was not visible due.to installed duct work. The worst area
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in the visible portion was approximately 30-feet circumferentially and up to 4 inches above the floor with less severe corrosion in both l
directions circumferential1y toward the inaccessible areas.
A thorough inspection of this area showed maximum depth of 1/10 inches near' the concrete / containment interface and general corrosion of
1/16-inch, Nominal wall thickness is 1-inch with a calculated f
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minimum wall thickness, using the most conservative Code methodology, f
of.864-inch.
Recent Code Analysis methodology allows.53-inch minimum wall.
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Root cause appeared to be standing water which was not removed during a number of years of operation.
During early years of operation the l
licensee had left the construction guard pipes around the drain f
lines.
These were 4 inch high' guards used to protect ' debris from
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entering the drains during construction which should have been removed prior to operation.
Also the licensee had allowed standing water to remain in the annulus.
This water was apparently from
instrument line leaks. and floor ~ irregularities prevented complete
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drainage.
The'11censee. cleaned and recoated the accessible area and i
implemented a weekly surveillance to prevent standing water.
Longer
term corrective actions planned include repair such as welding and recoating of the. entire circumference during the -next outage.
Observation of Unit 1 indicated similar degradation.
Plans are to
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make repairs during the next outage.
The inspector, along with
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NRC/NRR personnel inspected Unit I and reviewed the licensee'
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methodology for determining minimum wall thickness.
The licensee analysis appeared adequate and no concerns were identified relative to licensee proposed corrective actions.
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MaintenanceObservattens(627.3)
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Routine maintenance activities were reviewed and/or. witnessed by the mident inspection staff to ascertain procedural and performance adequacy conformance with applicable Technical Specifications.
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14 "w February 1989. Work Request 56105 was performed on the Unit 2 contvinment Purge Exhaust Fan.
Following maintenance the fan was returned
to service with no retest having been performed. The Performance section received the WR on August 9.1989 and identified no retest had been performed.
Between February and August, the Unit 2 core had been unloaded and reloadad.
Technical' Specification 3/4.9.4 requires that Containment
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Purge Exhaust Fans be demonstrated operab'ie prior to Core Alterations.
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The licensee determined that the Technical Specification had not been violated, since a routine surveillance on the system was performed prior.
t to the beginning of Core Alterations.
The licensee notified the. inspector that during a routine review of -
documentation, it was revealed that following maintenance on the 2B Charging Pump, the pump was declared operable without a retest.. This pump
was used to meet Technical Specification requirements for'a Charging Pump being operable prior to Core Alterations.
Following determination of the missed retest, the. pump was successfully retested.
During the core _
reload, however, the pump had not been demonstrated operable, as required by Technical Specification 3.1.2.3.
The inspectors are currently evaluating these probius to determine if they are repetitive and 'if j
requirements were. exceeded.
The licensee is submitting an LER on the
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charging pump.
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i A maintenance team inspection completed its work with an exit interview on l
July 28, 1989.
(See Report 369.370/89-15).
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No violations or deviations were identified.
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Licensee Event Report (LER) and 10 CFR Part 21 Report Followup (90712, 92700)
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The below listed Licensee Event Reports (LER) were reviewed to
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determine if the information provided met NRC requirements.
The determination included: adequacy of description, verification of
compliance with Technical. Specifications and regulatory requirements.
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corrective action taken. existence of potential generic problems.
i reporting requirements satisfied, and the relative safety l
significance of each event.
Additional inplant reviews and
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discussion with plant personnel, as appropriate, were conducted for i
those reports indicated by an (*).
The following'LERs are closed.
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369/88-04: Unit 1 and 2 Entered TS 3.0.3 when EVCB Vital Battery
Charger was Deenergized due to Unknown Reasons while EvCA Vital Battery was Inoperable. -TS 3.0.3 was entered for only 14 minutes and licensee ' corrective actions appeared appropriate.
- 369/88-12:
Investigation of Possible Valve Actuator Problems. This-i problem, involving operator motor mounting screws being too small, was originally identified at the licensee's Catawba facility.
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valve (ICF 1278) was found defective at McGisire and was fixed.
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Preliminary review showed no other problems should-exist, however.
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l inside containment inspections were planned as added assurance.
Recent inspection of Unit 2 identified no problems.
The inspector
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reviewed inspection documentation and verified the inspection was planned for Unit 1.
369/88-16:
A Containment. Isolation Valve was Inoperable Because of Damage which Occurred due to a Procedure Deficiency.
369/88-40. Rev. 1:
Containment Integrity Breached and Fuel Movement Suspended when Three Temporary Penetrations were Found Leaking (Violation issued in Report 369.370/88-33)-
369/89-14:
The Land Use Census Technical Specification Requirement I
was not Completely Fulfilled because of Poor. Interface Between'
Groups.
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370/88-01:
Manual Turbine Trip Followed by Reactor Trip due to U
Decreasing Steam Generator level - Feedwater Control Valve Closed due to Failed Air Regulator.
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370/88-05:
A TS Required Fire Door was Blocked Open without
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Compensatory Actions due to Personnel Error.
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- 370/89-05:
Pressurizer Pory 2NC-32B would not have Actuated in the.
Low Pressure Mode of Control because of an Inappropriate Action and Contributing Management Deficiency. The licensee's evaluation showed this to be. an error by an engineer in documenting a variation notice
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which included a wiring error.
The wires error were later changed
without performing adequate' functional ' testing.
Corrective actions for a previous violation relative to functional testing. appear
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appropriate for this circumstance (See Report 369.370/89-18, para.
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5.b).
A contributing cause may be the fact that electrical schematic
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drawings, do not contain interconnection wiring - information..
The licensee plans to incorporate String List (licensee interconnection wiring and termination information) information-into electrical-drawings.
However, during this period Residual Heat Removal System.
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relief valve was available for overpressure protection.
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- 370/89-06:
Technical Specification = 3.0.3 was Entered to Perform a Reactor Coolant System Thermal Mixing Test.
This.was a voluntary entry into TS 3.0.3 to perform a valuable test.
Test methodology and
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safety evaluation was reviewed in' advance by NRC and the inspector witnessed the test.
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The inspector verified that the licensee had received and evaluated 10 CFR Part 21 Reports applicable to the plant and' had taken corrective actions as necessary.
The following Part 21 items are
closed:
P2186-03 AFW Turbine Driven Pump Valves Fail to Open on Demand -
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Stronger Springs Installed.
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Contromatics Actuator W/ Emergency Override Operator
l W/ Jackscrew Handwheel Units May Suffer Weaving / Fouling of l
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P2188-06 (Unit 2 Only):
Inconel 600 Steam Generator Tube Plugs
' Susceptible to Stress Corrosion Cracking Supplied By B and W.
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l No violations or deviations were identified.
7.
Followup on Previous Inspection Findings (92701, 92702)-
The following previously identified items were reviewed to ascertain that l
the licensee's responses, where applicable, and licensee actions ~ were 'in l
compliance with regulatory requirements and corrective actions have been completed.
Selective verification included record review, observations.
I and discussions with licensee personnel.
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(Closed)
Licensee Identified Violation 370/87-02-02:
Failure to Perform 4-Hour Rod Position Ch:!cks Per TS 4.1.3.1.1. 4.1.3.2 and 4.1.3.6.
The licensee appropriately reported the event and
corrective actions were completed including procedural changes and personnel training (See LER 370/87-02).
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(Closed)
Inspector Followup Item 369.370/87-39-01:
Followup of Licensee Corrective Action for Vital Battery Team Inspection.
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licensee responded to violations issued in NRC Report 369.370/85-10 via letters dated March 12, 1985 and September 13, 1985.
Corrective actions included changes to surveillance procedures, personnel training, and improved communications through staff meetings.
The inspector reviewed procedure PT/0/A/4350/288, 125 Volt Vital Battery Quarterly Inspection to verify appropriate procedure changes were implemented.
In addition the inspector reviewed completed surveillance records of the most recent quarterly test for all four battery banks, and no further problems were identified, c.
(Closed)
Unresolved Item 369/87-39-02:
Review Administrative
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Control of Foreign Objects Entering Refueling Canal.
This ' item -
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involved concerns relative to the adequacy _of licensee control _s since several objects were found in the refueling cavity and vessel during refueling.
The inspector reviewed documentation of licensee actions
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relative to this concern.
The licensee communicated the need.for
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increased emphasis on the control of. foreign objects entering the
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refueling canal area to all personnel via a Station Manager Memorandum dated March 25, 1988.
Improvements were made to procedure MP/0/A/7150/12; Refueling Canal Cleanliness Watch and in tool controls such as committing additional personnel to monitor tool controls, providing better storage containers, trash containers, and personnel training. Licensee actions appear thorough and appropriate to the circumstances.. Therefore, this item.is closed.
d.
The inspector reviewed licensee responses and_ implementation relative to selected Three Mile Island Action Items as'follows:
L II.F.2.3.B. Instrumentation for Detection of Inadequate Core Cooling.
The licensee has implemented instrumentation as described in the FSAR, Table 1.8-1, page 17 which includes a i
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subcooling monitor, vessel level instrumentation and core exit thermocouples.
Two trains are provided and displayed. in the control room, h
II.K.3.1.B. Automatic Power. Operated Relief Valve Isolation
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System.
Review and analysis showed that this system was not
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necessary and is documented in the FSAR, Table 1.8-1,.page 25.
II.K 3.5.B. Automatic Trip of Reactor Coolant Pumps During Loss-of Coolant Accident. -(LOCA).
Licensee response is documented in the FSAR, Table l.8-1, page 25. The licensee has, l
implemented the most recent Westinghouse guidelines for tripping pumps which are manual trips only upon loss of subcooling and i
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Trips are minimized for Steam Generator Tube rupture and non-LOCA events.
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II.K.3.10. Proposed Anticipatory Trips.
Licensee response to-l
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this item is contained in the FSAR. Table 1.8-1. page 25-27.
The licensee has reinstituted the reactor trip on turbine trip at reactor power greater than 48%.
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These items are considered closed, e.
(0 pen) Violation 369.370/89-14-01:
Inadequate Surveillance Procedurcs for.MSIV's.
In response to Information Notice No. 88-51 the licensee committed to testing the closure of the Main Steam Line Isolation Valves by spring force alone, with no air assist, as part of a JC0 which stated that the spring was sufficient.to close the-
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MSiVs without air assist.
On September 8, 1989, the licensee tested the MSIVs using the spring closure only.
The times, without air
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assist. Varied between 4.2 seconds and 13.8 seconds, with only three of.the eight valves meeting the Technical Specification requirement of less than five second closure.
After consultation with~ the Design Engineering group, the licensee decided' that a determination of the root cause of the slow stroke times and a qualification of the non-safety-related portions of the air ' supply system was needed.
Tests were run to determine the minimum capabilities of the. air receiver tank and the integrity of the two series check valves on the-receiver air supply lir.e.
The licensee retested the valves with' air'in its normal line-up and with air just from the receiver tanks. with instrument air valved out.
In both cases, the valves stroked well within the Technical Specification requirements, comparable to' the stroke times received i
during the 1988 testing.
Further investigation revealed that the air supply port to the over-piston area of the control cylinder isolated during the testing which removed the air supply.
As the cylinder stroked down, then, a vacuum was pulled on the over-piston area. The (
higher than expected differential pressure across the cylinder slowed
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the cylinder travel time.
When a vent path was provided for the
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cylinder, the stroke times returned to values comparable' to with-air times.
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Work continued through September 11 to qualify the MSIV air supply
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systems on both units, to allow Unit I continued operations and to allow Unit 2 startup from the refueling outage.
A modification is i
being developed to provide a vent path for the cylinder when air'is
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isolated to it.
An operability statement for Unit 1 has been produced which states that with the newly qualified air supply, the valve stroke times will meet Technical Specification requirements.
Since no modifications were required to qualify the system, the MSIVs were considered to be operable during the period between the original JC0 and the time when the air system was qualified.
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This lack of ability to meet stroke time requirements without the use of air assist is considered an aspect of violation 369. 370/89-14-01
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and the licensee was requested to fully document these additional actions to NRC.
j No violations or deviations were identified.
8.
Evaluation of Licensee Self-Assessment Capability (40500)
The inspector reviewed the licensee programs for self assessment.
This l
paragraph also serves to document previous reviews in.this area..
The licensee has implemented various programs for self assessment.
The licensee is not connitted to an onsite review connittee. however, a McGuire Safety Review Group (MSRG) is established.. This group functions
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to observe plant activities through independent surveillances, investigate events and develop Licensee Event Reports including corrective actions. A corporate Nuc1 car Safety Review Board (NSRB) is also established.
This-group provides independent review and audit of various. activities.
Activities of this group are described in Technical Specification 6.5.2.
Both groups are made up of at least five dedicated individuals with various technic 31 backgrounds.
The MSRG and NSRB report to the Manager. Nuclear Safety Assurance (NSA).
i Also reporting to the Manager NSA is Operational Nuclear Safety.
This group coordinates the Operating Event Program (followup industry events.
NRC Notices, etc), performance indicator reporting. Nuclear Plant Reliability Data System. trending. (incident reports. NRC findings.
operating experience, etc.), component failure monitoring, and INP0 information and this group also reviews event safety analysis.
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I The Quality Assurance Department conducts corporate audits of plant activities as well as numerous surveillances with.the onsite surveillance
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group.
The licensee has e site performance indicator program which includes goals for Station Generation. Radiation Exposure. Outstanding Work Requests.
Maximum Outage Days, and Quality of Operation (includes reactor trips.
contaminated floor space personnel errors. personnel contaminations.
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waste generation, control room indication problems, injuries and thermal l
performance).
The licensee aiso has implemented a problem investigation process via Station Directive 2.8.1.
Problem Investigation Reports are issued to document problems and assure adequate specific and preventive corrective actions are implemented.
The inspector attended selected portions of an NSRB meeting to evaluate the deptn of their reviews.
The board was thoroughly briefed by plant personnel in multiple areas. in order to provide independent review and audit of designated activities.
The inspector noted that the members of
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the board were receiving technical input and were asking technically sound questions and making valid observations on the issues.
Comprehensive feedback was offered to plant management personnel.
Whenever an issue
could not be resolved at the meeting, action items were being issued. with J
expected response dates to the responsible organizations.
The board reviewed minutes of previous meetings and dealt with items that remained open and with questions posed by various members of the board.
The inspector noted no weaknesses or adverse findings during his observations of the NSRB meeting.
The inspectors have routinely reviewed all PIR's. LERs. MSRG surveillance reports, performance indicators. INP0 results and QA surveillance summaries.
A detailed inspection of QA audits and surveillances was
accomplished and documented in NRC Report 369.370/89-18.. Problems in
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these areas have been documented as these are identified.
Previous reports have documented that outstanding work requests ar.c control room indication problems have not been appreciably reduced.during-the year.
Occasionally root cause. analysis of events is weak and therefore preventive actions are incomplete.
Examples were previously cited (See
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violation 88-30-02 relating to'LER 370/88-09; violation 88-33-08 relating to LER 369/88-40 and violation 369'370/89-16-02 relating to LER
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369/89-06).
Also timeliness of Design Engineering review of problems.has been a recurring problem (See Reports 369.370/89-14 paragraph 4.d.;
369,370/89-16. paragraph 6.a. and 369.370/89-24, paragraph 9).
These concerns will be addressed as part of the ongoing routine inspection program.
Generally the licensee appears to have a broad based program for
self assessment which is generally aggressive.
While all goals have'not
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been met management stresses goals and performance.
No violations or deviations were identified.
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9.
Ventilation Systems Review (37700, 61726)
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Annulus Ventilation System On September 19, 1988 NRC issued Information Notice No. 88-76: Recent Discovery of a Phenomenon Not Previously Considered in the Design of Secondary Containment Pressure Control.
The Notice described a situation whereby another licensee had not considered adequately the j
temperature-induced difference in the pressure gradients inside. and outside secondary containment as these vary with elevation. Whenever
outside temperature is lower than the temperature maintained in the secondary containment, the vertical pressure decrease at the higher elevation outside the secondary containment is greater than the
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pressure decrease inside the secondary containment because of the j
higher density of the colder air.
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McGuire Station has a steel containment :with an Annulus area
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(secondary containment) between the containment and a concrete
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Reactor Building.
The Annulus Ventilation (VE) System is the system t
which performs essentially the same function.as the system described i
in the Notice.
The Design Basis described in the Final Safety
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Analysis Report (FSAR) 6.2.3.1 includes " Produce and maintain a negative pressure in the annulus following a LOCA".
The system is
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designed to cycle between recirculation and discharge modes of operation to accomplish this purpose.
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I transmitter causes a switch to recirculation at negative 3.5-inches I
water gauge pressure.
At negative.5-inches the transmitter would cause.the system to go back to discharge, and so forth.
Both modes draw air through charcoal filters and discharge is to.the Unit Vent.
Other design basis are:
" Minimize the release of radioisotopes following a LOCA by recirculating a large volume of annulus air
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relative to the volume discharge for ' pressure maintenance" and
" Provide long-term fission product removal capability by decay and.
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filtration".
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The licensee apparently received the Notice in -late September.
however, it was misplaced until late January 1989.- when another copy was obtained and forwarded to Design Engineering (DE) for review. DE r
apparently initially considered that they did not hr" a similar
problem but continued to review the issue through dm gathering and
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analysis. On August 24. 1989 DE determined that the upper regions of
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atmosphere under certain temperature profiles.
The pressure transmitter sensing point for the VE system are located in the lower
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part of the annulus.
The surveillance test' connection are also located near the transmitters.
Re-analysis indicated that the set
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points should be changed to' negative 1.2 and 4.2-inches to maintain
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the negative annulus pressure during all temperature profiles.
The transmitters was reset and the Unit 1 (Operating Unit) system was l
tested satisfactorily on August 25. 1989.
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Technical Specification (TS) 3.6.1.8 requires two VE Systems to be operable in Modes 1 through 4.
In addition 10 CFR 50. Appendix B.
Criterion III requires that " Measures shall be established to assure
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that applicable regulatory. requirements and the design basis, as defined in 50.2 and as specified in the license application for
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those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings procedures, and instructions"' and further requires that "The design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the l
performance of a suitable testing program".
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The inappropriate location of the transmitters sensing point and
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l surveillance test point appears to be a violation of the above requirements occurring from initial operation. until August 25. 1989.
f for Unit 1 and until July 5,1989 for Unit 2 when the unit entered a
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refueling outage.
This is Violation 369.370/89-24-03:
Failure to Implement Adequate Design Controls Leading to Inoperable Ventilation Systems Under Certain Conditions.
This apparent violation is being
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considered for escalated enforcement and therefore a Notice of.
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Violation is not being issued with this report.
Other problems have also been identif.ied with VE.
Environmental
qualification of wiring * in the filter preheater and operation of
train cross-connect valves were questioned in ' July (See Report 369.370/89-18 para. 3.d.). The wiring problem required compensatory.
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instructions to leave both trains of the system in operation once
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these are started.
Two other questions were. raised as a result of continued DE review in August.
A question was raised as to whether water in lower ducting in the annulus which had occurred in the past had affected operability of VE.
Evaluation showed that the system would have performed satisfactorily and the stainless steel ducting should not have been damaged.
On August.29. 1989. DE. identified a number of transmitter failures- (identified in PIR-0-M89-0219 Operability Evaluation File) which could cause a. train to continuously operate in the discharge mode and continue to' draw down pressure in the annulus.
An analysis showed that maximum negative
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pressure would be 22-inches water gauge equivalent to.8 psig vacuum..
Evaluation showed that components (seals, penetrations, ducting, etc.) would withstand the pressure without damage and dose' rates
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would be within design basis limits. This failure mode had not been considered in the original design.
These problems coupled with a problem concerning the Hydrogen Skimmer System described in Report 369.370/88-24 and the ' problem described
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systems.
This weakness will be evaluated during followup of the violation described herein.
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Control Area Ventilation l
The Control Room (CR) portion of Control Area Ventilation System (VC)
is'a two train shared system for a common CR designed as described in the FSAR 6.4 to " maintain a positive pressure.in the Control Room".
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TS 4.7.6.e.3 requires verifying that VC maintains at least.1/8-inch water gauge positive pressure " relative to outside atmosphere".
The
system is designed to take outdoor air suction from two locations (two intakes each) on opposite sides of the plant through charcoal'
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filters and into the CR.
The system design prevents iodine from l
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iodine dose rates to personnel in the CR. High radiation closes the L
intakes and operators are instructed per procedures to open the least contaminated set if all four close.
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On July 16. 1989 a portion of a CR door seal was found missing while
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While reviewing
this problem the licensee questioned the adequacy of the surveillance i
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The "outside" reference point being used was the Cable Spread Room (CSR) rather than outside atmosphere.
The CSR test point was l
part of the original design.
On July 22, 1989 a Problem Investigation Report (PIR) was issued to DE to evaluate the question.
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On August 18. 1989. DE determined that the VC surveillance test was inadequate and outside atmosphere should be the conservative reference point.
Initial testing on August 18 showed pressure to be neutral when referencing to outside atmosphere and using one set of intakes.
Continued testing on August 18-19 showed that the 1/8-inch pressure.could be' maintained by taping over five of the nine CR doors and opening all four intakes.
DE analysis showed that dose rates-would not be exceeded with all four intakes open.
The system was declared conditionally operable with instructions to operators to maintain the doors taped and to maintain all four intakes open.
TS 3.7.6 requires two independent Control Room. ventilation systems to be operable in Modes 1 through 4.
.10 CFR 50. Appendix. B. Criteria III requirements are described in paragraph 9.a. above.
The inappropriate system reference point appurs to be a violation of these criteria which occurred since init13.1 operation.
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this appears to be a second example of the.iolation described'above.
It is noted that a violation (88-24-03) has been previously issued involving inoperability of the Hydrogen Skimmer System due to inadequacies in design and testing.
The licensee. in a response to this violation dated February 15, 1989, committed to a task force review of systems (including the VC and VE systems) to verify-that
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testing was adequate to determine that the design basis was being met.
This appears to have been an opportunity to discover the above
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One violation was identified as described above.
10. Facility Modifications and Design Interface (37701)
f The inspector held discussions with Corporate and on-site Design Engineering (DE) personnel to discuss DE interface with station personnel and to discuss improvements which have been made to the modification
program.
Based on. these discussions, conversations with site station management and inspector observations, the on-site DE group has clearly been a benefit to improving DE/ Site interface.
This group, although not normally in the direct chain of responsibility for modifications, stays t
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cognizant of how well the process is working and conducts independent reviews.
The group has coordinated one modification and'has fed back an improved methodology for fixing a lighting problem.
This group is also actively involved in plant problems, providing ' guidance and providing coordination of interfacing with Corporate Design Engineering to address i
the problems.
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In late 1988. DE reorganized into a site project oriented structure.
This
appears to have improved the interface since specific-personnel responsibility is more clearly defined.
Also the inspector has witnessed-DE personnel from the new organization at the site on a number of occasions observing plant activities and actively involved in problem
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solving.
The licensee has had an improved modification process in place for approximately two years, called TOPFORM.
This program added
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additional steps in the process such as scope review, pre-design survey, client (station). review. integrated design review, test acceptance criteria sumary and post-implementation review.
Improvements were made in areas such as 10 CFR 50.59 reviews and other. safety reviews.
Further NRC reviews will be conducted regarding specific modifications.
No violations or deviations were identified.
11. Exit Interview (30703)
The inspection scope and findings identified below were sumarized on September 12, 1989, with those persons indicated in paragraph 1 above.
The following items were discussed in detail:
Violation 369.370/89-24-01:
Failure to Adequately Document Shift Staffing-MultipleExamples(paragraph 3).
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Unresolved Item 369.370/89-24-02:
Control Room Access Control Problem
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(paragraph 3.c.)
Violation 369.370/89-24-03:
Failure to Implement Adequate Design Controls
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Leading to Inoperable Ventilation Systems Under Certain Conditions (paragraph 9).
The licensee was informed that this issue is being considered for escalated enforcement and an enforcement conference would be held at the NRC/ Region'II office.
The inspector requested that the licensee document the most recent problems with Main Steam Isolation Valves (see paragraph 7.e.) in a supplemental response to the previously issued violation and/or an LER.
One of the ventilation problems associated with the above listed violation involved an NRC Information Notice which had been misplaced for approximately four months and then took another 7 months to evaluate. The licensee was asked to comit to assuring that other Information Notices are not outstanding an excessive Manager Operation Nuclear Safety) period of time. The licensee (S. T. Rose.
comitted to the following actions:
Audit of Information Notices 1988 to present date to verify no others have been misplaced (completed by September 30, 1989)
Enforce the existing escalation policy for outstanding) notices (This policy had been on hold during the DE re-organization and also to
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apply the policy retroactively to outstanding evaluations. (completed
by September 30,1989).
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Evaluate Operating Experience program personnel's. understanding of.
whom issues should be forwarded to and take appropriate corrective action such as training (completed by October 31,1989).
. Re-emphasize a timely initial review of problems to assure the group
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charged with the review is the appropriate group.
Letter may also emphasize initial review for significance.
Also program will be evaluated to determine whether procedure changes are needed - (letter to be issued by October 31, 1989; procedure changes and training-initiated by December 31, 1989).
The licensee representatives present offered no dissenting comments, nor
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did they identify as proprietary any of the information reviewed by the inspectors during the course of their inspection.
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