ML20134L146
| ML20134L146 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 02/10/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20134L104 | List: |
| References | |
| 50-369-96-11, 50-370-96-11, NUDOCS 9702180302 | |
| Download: ML20134L146 (50) | |
See also: IR 05000369/1996011
Text
U.S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos:
50-369. 50-370
License Nos:
Report No:
50-369/96-11, 50-370/96-11
Licensee:
Duke Power Company
Facility:
McGuire Generating Station Units 1 & 2
Location:
12700 Hagers Ferry Rd.
Huntersville, NC 28078
Dates:
December 1. 1996 - January 11, 1997
Inspectors:
S. Shaeffer Senior Resident Inspector
M. Sykes, Resident Inspector
G. Harris Resident Inspector
N. Economos. Regional Inspector, (M1.2)
D. Forbes, Regional Inspector. (R.1. R.2, R.4, R.7)
E. Testa, Regional Inspector, (R.1, R.2, R.4, R.7)
E. Girard, Regional Inspector (E.1)
R. Moore, Regional Inspector, (E.4. E.7)
R. Hall. NRR. (E.1)
T. Scarbrough. NRR. (E.1)
S. Tingen. NRR, (E.1)
Accompanying Personnel:
M. Holbrook, Consultant. INEL
Approved by:
C. Casto, Chief Projects Branch 1
Division of Reactor Projects
ENCLOSURE 2
9702180302 970210
ADOCK 05000369
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EXECUTIVE SUMMARY
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McGuire Generating Station. Units 1 & 2
NRC Inspection Report 50-369/96-11. 50-370/96-11
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This integrated inspection included aspects of licensee operations, engineer-
ing, maintenance, and plant support. The report covers a 6-week period of
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resident inspection.
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Ooerations
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O erator aerformance in recognizing a feedwater valve hydraulic fluid
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1 ak at tie valve controller was good. Operators placement of the unit
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in a favorable operating condition to preclude a more serious transient
was timely. The subsequent return to rated power was conducted with
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good attention to plant parameters and personnel safety (paragraph
02.1).
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Review of the monitoring, root cause determination, and prevention of
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component mispositionings concluded that the licensee's program was
properly focused and receiving good management attention.
However.
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while the number of significant mispositioning events (as defined by the
licensee's program) has decreased, the overall number of mispositioned
components was considered high indicating further improvements could be
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made with continued management focus (paragraph 07.1).
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Maintenance
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Initial use of automatic welding equipment to weld secondary pipe welds
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was unsuccessful and the licensee stopped work until the problem was
identified and corrected.
Welding of dissimilar metal welds manually
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was satisfactory.
Radiography, material control, and personnel training
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within the areas inspected was adequate (paragraph M1.2).
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Licensee response to the failure of 1RN171 during testing was good.
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Proper compensatory actions were taken to maintain system operability
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until an appropriate temporary modification was developed, evaluated.
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and implemented (paragraph M2.2).
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Preparation and coordination of Boraflex testing was good.
Previous
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test results coupled with current testing indicated no gross boraflex
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degradation had occurred.
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Enaineerina
The licensee met the intent of GL 89-10 in verifying the design basis
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capabilities of their MOVs.
Several weaknesses were identified.
Examples included the limited quantity or quality of data that was used
to establish the capabilities of several groups of MOVs and the
licensee's application of PRA in establishing MOV operability. The
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licensee initiated actions to correct the more significant weaknesses
and IFI 50-369, 370/96-11-01 was identified to track their completion.
ENCLOSURE 2
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Strengths were identified which inc'uded strong management and personnel
support, application of state of the art technology, leadership in
addressing industry problems, and the detailed packages of data and
evaluations that were developed for each valve group (paragraph E1.1)
A related NCV was identified involving improper deferrals of periodic
MOV lubrications (NCV 50-369. 370/96-11-04. Section E8.3).
Based on the NRC staff's review of the McGuire GL 89-10 program and its
implementation, and the actions established by the licensee in PIP 0-
M96-3542, the NRC staff is closing its review of the GL 89-10 program at
McGuire.
The completion of these licensee actions will be assessed as
part of the NRC staff's monitoring of the licensee's long-term MOV
program (paragraph E1.1).
A Violation was identified (VIO 50-369/96-11-02) concerning the
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installation of a temporary security fence to restrict access to the
Un't 1 exterior valve vault.
Installation of the security fence was not
conducted in accordance with established station procedures (paragraph
E2.1).
The commercial grade dedication (CGD) process was effectively
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implemented at McGuire by the Procurement Engineering organization and
was consistent with applicable regulatory requirements.
Resolution of
identified procurement problems was adequate. Self assessments provided
adequate monitoring of station performance in CGD activities.
An URI
was identified (URI 50-369. 370/96-11-03) for further NRC review of the
environmental qualification of Grinell hydraulic pipe supports
(paragraph E4.1).
Plant Suncort
The inspectors determined that the licensee effectively implemented a
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)rogram for shipping radicactive materials required by the NRC and
Jepartment Of Transportation regulations (paragraph R1.1).
Radiological facility conditions and housekeeping in radioactive waste
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storage areas were good.
Material was labeled appropriately, and areas
were properly posted.
All exposures were below regulatory limits and
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the licensee was continuing to maintain exposures As low As Reasonably
Achievable (paragraph RI.2).
A Violation was identified (VIO 50-369/96-11-05) for the failure to
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conduct a 10 CFR 50.59 written safety evaluation to 3rovide the bases
for the determination that a test not described in t1e FSAR did not
involve an unreviewed safety question. This test involved the lowering
of hydrazine levels in Unit 1 secondary systems which could have
potentially impacted reactor power from the effects of the change on
feedwater venturi fouling (paragraph R1.3).
ENCLOSURE 2
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The licensee had maintained an overall high level of operability for
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radiation monitors in 1996 and was effectively tracking monitor
performance. An Inspector Followup Item was identified (IFI 50-369/96-
11-06) to track the closecut actions on the Problem Investigative
Process associated with the solubilization of the Cobalt-58 (paragraph
R2.1).
The licensee had continued to maintain effective capabilities to perform
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environmental samples (paragraph R2.2).
Radiation Protection technicians and Chemistry technicians were
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receiving an appropriate level of refresher training to enhance work
activities (paragraph R4.0).
The licensee was effectively conducting formal RP and Chemistry audits
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as required by Technical Specifications and completing corrective
actions in a timely manner (paragraph R7.0).
Review of an annual fire drill conducted in December 1996 concluded that
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the drill was conducted in a realistic and professional manner.
The
licensee's critique of the drill was candid and identified several
issues, which, once resolved, shculd improve the plant and offsite
agency response to a fire emerger.c. (paragraph FS).
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ENCLOSURE 2
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Report Details
Summary of Plant Status
Unit 1 began the ins)ection period at approximately 28% power while the
licensee repaired a lydraulic leak at feedwater isolation valve ICF26.
Following repairs, the unit was returned to 100 percent power on December 3.
and operated at power throughout the remainder of the reporting period.
Unit 2 operated at 100 percent power throughout the reporting period.
Review of UFSAR Commitments
While performing inspections discussed in this report the inspectors reviewed
the applicable portions of the UFSAR that were related to the areas inspected.
The inspectors verified that the UFSAR wording was consistent with the
observed plant practices, procedures, and/or parameters.
I. Operations
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Conduct of Operations
01.1 General Comments (71707)
Using Inspection Procedure 71707. the inspectors conducted frequent
reviews of ongoing plant operations. In general, the conduct of
operations was professional and safety-conscious; specific events and
noteworthy observations are detailed in the sections below.
Operator
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monitoring of plant parameters for abnormal conditions was good. TS LCO
compliance was maintained, and special attention was given to specific
challenge areas such as freeze protection monitoring.
02
Operational Status of Facilities and Equipment (71707)
02.1 Hydraulic Fluid Leak at Feedwater Isolation Valve ICF26
a.
Insoection Scope
At the beginning of the inspection period. Unit 1 operated at
approximately 25 percent power. Operators had reduced Unit 1 power in
order to realign main feedwater flow from the D steam generator main
feedwater nozzle to the upper feedwater (auxiliary feedwater) nozzle.
The licensee had previously performed a rapid downpower in accordance
with station abnormal procedures and had realigned feedwater flow to
reduce the effects of an inadvertent closure of feedwater isolation
Valve ICF26 due to a previously identified hydraulic fluid leak at the
valve controller.
Valve ICF26 is located in the Feedwater System flowpath to the D steam
generator main nozzle.
Maintenance technicians were dispatched and
identified a failed 0-ring at the oil side accumulator fill valve.
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ENCLOSURE 2
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Valve ICF26-is a' safety-related hydraulic isolation valve. The valve
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receives a signal to close on a Safety Injection. Low Tavg coincident
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with Reactor Trip. HI-HI doghouse water level. or HI-HI steam generator
level.
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b.
Observations and Findinas
With the unit stabilized at approximately 25% power, the licensee began
repairs to correct the failed 0-ring and re-establish normal feedwater
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flow to the D steam generator.
During the troubleshooting and repair
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efforts, maintenance technicians found the accumulator fill valve loose
enough to remove without the use of a wrench indicating insufficient
The licensee could not determine whether the insufficient
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' torque was due to maintenance on the fill valve or a failure to verify
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torque following installation of the manifold block assembly during
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1E0C10.
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During the repair effort the licensee repaired or replaced several other
components to enhance valve performance.
During functional testing, the
licensee determined that the associated hydraulic pump performance did
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not meet station acceptance criteria.
The pum) was replaced and tested
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satisfactorily.
The licensee also contacted t1e equipment vendor prior
to completion of the repairs to ensure that the repairs were adequate to
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prevent recurrence. The valve was stroke tested satisfactorily and
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returned to service.
Feedwater flow was realigned to the main feedwater
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nozzle and the Unit I was returned to 100 percent power.
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c.
Conclusion
Although the licensee identified evidence of a loss of torque at the
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fill valve, no root cause was identified. The inspectors also concluded
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that Maintenance and Engineering response following identification of
the hydraulic fluid leak was prompt and detailed.
Operator performance
in recognizing the controller malfunction and placing the unit in a
favorable operating condition was also good.
The subsequent return to
rated power was conducted with good attention to plant and personnel
safety.
07
Quality Assurance in Operations (40500)
07.1 Review and Control of Comoonent Mispositioninas
a.
InsDeCtion Scoce
During the inspection period, the inspector reviewed the licensee's
process for identifying, tracking, and improving the number of component
mispositioning events.
ENCLOSURE 2
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b.
Observations and Findinas
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The inspector reviewed implementation of the licensee's SM process
(McGuire Monthly Managers Meeting on Mispositionings) used to identify.
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track, and improve the stations' performance in the area of com)onent
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mispostionings. The process has included a monthly meeting wit 1 multi-
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disciplined managers, to discuss the most recent mispositioning problems
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and events. The issues are then classified by component. type of.
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activity, responsible group, and potential causes.
From the monthly
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meetings-, annual assessment reviews were performed, which provided
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recommendations from the SM members to incorporate changes to improve
the plant performance in the configuration control area. The inspector
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attended portions of 5M meetings, reviewed annual assessments of
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mispositioned components, and discussed the process with involved
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personnel.
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At the December 1996 SM review, recent mispositioning problems were
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presented to the members by the res)onsible managers. The inspector
considered that each individual pro)lem was discussed in good detail and
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allowed members to appropriately classify the severity of the problem
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and agree or disagree on the assigned root cause.
Managers presenting
issues were well prepared and the SM members exhibited a good
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questioning attitude. The inspector also reviewed several meeting
minutes from other 5M meetings.
The documents were detailed and
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contained specific action items incorporating process improvements made
by the SM members.
The licensee's process incorporated detailed
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trending of the identified mispositionings. such that corrective actions
could be taken for adverse trends in specific areas.
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Mis)ositioning issues at McGuire are classified as mispositioning
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pro)lems or events.
Hispositioning problems which have impacted plant
operations prior to discovery are classified as " events".
Based on the
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available data for several years, the number of " events" has decreased
from 11 in 1994 to 0 for 1996. Over this time, the number of
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mispositioning problems being reviewed has increased due to a lower
threshold for reporting misposition type problems.
This further
indicated to the inspector that the overall " events" have been reduced.
However, based on the overall number of mispositioning problems (68 for
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1996), the number of mispositioned components was considered high.
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Conclusions
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Review of the monitoring, root cause determination, and prevention of
component mispositionings concluded that the licensee's program was
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properly focused and receiving good management attention.
However.
while the number of significant mispositioning events (as defined by the
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licensee's program) has decreased, the overall number of mispositioned
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componerts was considered high indicating further improvements could be
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made.
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ENCLOSURE 2
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Miscellaneous Operations Issues (92700)
08.1
(CLOSED) LER 50-369/96-02:
Inadvertent Manual Initiation of a Unit 1
Feedwater Isolation due to an Inappropriate Action.
The event occurred
when an operator was preparing to close the reactor trip breakers, the
operator inadvertently depressed the train B feedwater isolation
initiate push button.
Unit 1 operators promptly recognized the
inadvertent CF isolation condition and initiated actions to reset the
function.
No plant transient resulted from the event.
Subsequent
corrective actions were taken emphasizing the use of self-checking. The
inspectors reviewed additional corrective actions taken to eliminate
human factor contributors to the issue including the addition of covers
to prevent inadvertent actuation of these push buttons. This LER is
closed.
II. Maintenance
M1
Conduct of Maintenance
M1.1 (ie. neral Comments (61726 and 62707)
The inspectors witnessed selected surveillance tests to verify that
approved procedures were available and in use. test equipment in use was
calibrated, test prerequisites were met, system restoration was
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completed, and acceptance criteria were met.
In addition. resident
inspectors reviewed and/or witnessed routine maintenance activities to
verify. where applicable, that approved procedures were available and in
use, prerequisites were met, equipment restoration was completed, and
maintenance results were adequate.
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a.
D oection Scope
The inspectors observed all or portions of the following work
activities:
Procedure / Work Order
Title
IP/0/A/3090/30
Installation and Removal of Temporary
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Modifications
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PT/0/A/4550/36
Controlling Procedure for SFP Storage Rack
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Boraflex Examination
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Installation of Temporary Modification
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b.
Observation and Findinas
The inspectors found the work performed under these activities to be
professional and thorough. All work observed was performed with the
work package present and in active use.
Technicians were experienced
ENCLOSURE 2
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and knowledgeable of their assigned tasks.
The inspectors frequently
observed supervisors and system engineers monitoring job 3rogress, and
When applicable, personnel were present whenever required
)y procedure.
quality control
appropriate radiation control measures were in place.
In addition, see the s)ecific discussions of maintenance observed under
M1.2. M2.1. and M2.3.
)elow.
M1.2 Steam Generator Reolacement (50001)
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a.
Insoection Scope
To evaluate the licensee's Steam Generator Replacement Program (SGRP)
for McGuire Unit 1. by observation of selected work activities including
welding. material storage and handling, nondestructive testing,
machining and welder training.
b.
Observations and Findinos
Apolicable Codes and Standards
By review of the applicable sections of the McGuire FSAR. Steam
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Generator (SG) Replacement Manual and various scope documents, the
inspectors ascertained that the following ASME Code Sections and
Editions were applicable to the SGRP.
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SG replacement. ASME Code Section XI. 1989 Edition. Article
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Weld process, inspection and leak testing. Corporate Manual
Section NSD-400 and Code Case N-416-1
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Welding procedures for secondary pipe butt welds, were qualified
to the latest ASME Code Section IX in effect at the time of the
qualification. The original construction code of record for
McGuire Units 1 and 2 is the ASME Code Section II, 1971 Edition.
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Main Feedwater Pioe Weldina
The Main Feedwater (CF) piping is being rerouted to accommodate the
location of the CF nozzle on the replacement Steam Generators. The
replacement piping is made from SA-335 (P11) material which was produced
from chrome-moly steel. This material was selected on the basis of its
demonstrated good resistance to flow assisted erosion corrosion attack.
The licensee is replacing all the feedwater piping from the CF nozzle
back to the crane wall including the bypass lines around the check valve
ENCLOSURE 2
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adjacent to the CF nozzle.
This rerouting added approximately 40 feet
of oiping from the crane wall, rising up to the 788 foot elevation.
The
Ecn;e of work on the CF system was addressed in modification package MG
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19420, "CF Piping Reroute Due to S/G Nozzle Relocation," February 9,
1996.
In an effort to minimize weld fabrication and pipe assembly inside
containment, the licensee is prefabricating CF pipe spools on-site at
the fab-shop. The method of weld fabrication has been changed from
manual to machine welding. This change was in part due to lessons
learned from the Catawba SGRP and for improvement in weld quality, ease
of welding and a corresponding increase in production.
The automatic welding machines used were Dimetric's Gold Track IV/DSP
models.
To gain the necessary proficiency in their operation, the
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licensee selected 32 welders. These welders were divided into four
teams and sent to the vendor's facility for training.
Each team trained
for a period of three weeks which consisted of classroom instruction,
hands-on machine familiarization and testing on pipe coupons.
On
January 7, 1997, the inspector accompanied the SGRP Weld Lead Engineer,
on a visit to the vendor's facility located in Davidson. North Carolina.
The inspector toured the training facility, observed the hands-on
training activities and discussed the program with the weld lab manager.
The training provided appeared adequate and should achieve its
objectives.
In the fab-shop, the inspectors observed fabrication of CF pipe spools
using the gas tungsten arc (TIG) welding machines discussed earlier in
the 3revious paragraph.
By review of Work Process Control Sheets (WPCS)
and rield Welds Data Sheets (FWDS) the inspector ascertained the
following:
CF pipe groove welds were being fabricated using a TIG welding machine
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procedure qualified for welding chrome-moly material.
The parameters of
the qualification were documented on Procedure Qualification Record
(POR) L-1400, Rev. O. dated November 31, 1996.
Production welds were
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fabricated using FWDS L-222A Rev. 1 generated to provide detailed
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information of the POR above, for use on production welds. As such, the
inspectors verified that machine settings were consistent with the
qualification parameters and the essential variables of ASME Code,
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Section IX.
ENCLOSURE 2
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Welding in progress at the time of this inspection involved the
following weld joints.
Weldina in Proaress
Weld #
lile
Description
Condition
CFIFW27-15
16"x.844"
Elbow to Pipe
Welding Out
S/G-10
CFIFW27-23
16"x.844"
Elbow to Pipe
Welding Out
S/G-1D
The inspectors also checked completed welds for weld reinforcement.
workmanship, weld and welder identification, cleanliness and component
identification for traceability. This work effort was performed on the
following weldments.
Completed Welds
Weld #
Size
Descriotion
Condition
CF1FW24-27
16"x.844"
Pipe to Elbow
Comaleted.
S/G-1A
RT Rejected
CFIFW24-27
16"x.844"
Pipe to Elbow
Completed.
S/G-1A
RT Rejected
CF1FW27-6
18"x.938"
S/G-ID
Completed.
RT Accepted
CF1FW27-20
16"x.844"
Pipe to Elbow
Completed.
S/G-1D
RT Rejected
As a followup to this field inspection, the inspector reviewed selected
Weld Process Control Sheets for completeness and accuracy including
signoffs for cleanliness, fitup, code inspector's hold points. preheat,
interpass temperature checks and final visual inspections as applicable.
In addition, the inspectors selected for review. qualification records
of welders who participated in the fabrication of these welds. A total
of seven welders were selected for a check of performance qualifications
and updates.
These qualification records were found to be in order.
Material Traceability and Control
Feedwater Piping - Material used for replacement of the feedwater line
was purchased to Specification DPS 1206.00-02 001. Rev. 7. July 11,
1995. and to the recuirements of ASME Code Section II and III.1989
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Edition.1989 Addenca Articles NC-2000 and NCA-3800 for SA 335. P11
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Class 2 Materials.
Associated elbows and reducers were purchased to
ENCLOSURE 2
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5)ecification DPS 1206.00-02-0003. Rev. 6 and to the requirements of the
a?ove-mentioned codes for SA 234. WP11 Class 1 material.
Material
identification and QA traceability numbers were as follows:
Material
Heat No.
0A No.
16" sch/80
942558
MC41918
Seamless Pipe
18" sch/80
194931
MC44334
Seamless Pipe
18" sch/80
195097
MC44336
Seamless Pipe
16" sch/80
9016A
MC41617
45* Elbow
Material
Heat No.
0A No.
16" sch/80
9017A
MC41618
90 Elbow
18" sch/80
1G4B2U1H9
MC45066
90 Elbow
16" sch/80
1G4B2U1I9
MC45068
45 Elbow
By review of certificates of conformance, the inspectors verified that
chemical analysis, mechanical tests, hardness, charpy V-Notch impact
tests and hydrostatic testing had been performed and that the results
were within code allowable limits.
Filler Metal - Material for fabricating feedwater pipe welds was
purchased to the licensee's Specification DPS-1206.00-02-0005. Rev. 003
and to the requirements of ASME Code Section 11. Part C and Section III
(95). NB-2400 for Class 1 material.
Filler metal used for this
application was as follows:
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Material Tvoe
Size
Heat / Lot No.
0A No.
ER80S-B2
0.045"
219389
897241
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ER80S-B2
0.035"
219389
897242
ENCLOSURE 2
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Material TvDe
Sjlg
Heat / Lot No.
0A No.
1/8" dia.
DN6209
855070
ERN1CR-3
3/32" dia.
CN6830
854652
ERN1CR-3
1/8" dia.
DM6577
854651
8y review of certified material test reports, the inspectors verified
that chemical, mechanical and weld tests performed, produced results
that were within code allowable limits.
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Nondestructive Examinations (NDE) Radicaraohv
Main feedwater pipe welds, fabricated onsite, were radiographed as
required by the applicable code. The licensee's code implementing
procedure for this examination was NDE-10 Rev.19. General Radiographic
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Procedure, which referenced ASME Code, Sections V and XI, 1989 Edition.
The inspector reviewed radiographs of completed main feedwater welds to
verify proper penetrameter type, size, placement, and sensitivity as
well as film density identification, quality and weld coverage. Welds
selected for this work effort were as follows:
Weld No.
S_izg
Descriotion
Status
CF1FW24-24
16"x.844"
Pipe to Elbow
Accepted: 12/9/96
CFIFW24-23
16"x.844"
Pipe to Elbow
Rejected: 12/12/96
Lack of fusion
(LOF) and porosity
CF1FW24-19
16"x.844"
Pipe to Elbow
Rejected: 12/12/96
and 12/17/96 LOF
and porosity
CF1FW24-27
16"x.844"
Pipe to Elbow
Rejected: 12/05/96
porosity
Accepted 12/10/96
.
CFIFW24-14
18"x.938"
Pipe to Reducer
Rejected: 12/10/96
porosity
Accepted: 12/11/96
CFIFW24-10
18"x.938"
Pipe to Reducer
Accepted: 12/05/96
CFIFW24-09
18"x.938"
Elbow to Pipe
Accepted: 12/11/96
ENCLOSURE 2
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Weld No.
Sjle
Descriotion
Status
CFIFW27-05
18"x.938"
Elbow to Pipe
Rejected: 01/07/97
LOF
CFIFW27-06
18"x.938"
Pipe to Elbow
Accepted: 12/17/96
CFIFW27-10*
18"x.938"
Reducer to Elbow Accepted: 12/19/96
CF1FW27-20
16"x.844"
Pipe to Elbow
Rejected: 01/07/97
LOF
- Records on file indicated that this weld had been postweld heat
treated (PWHT).
A review of associated records revealed that PWHT
parameters were consistent with code and procedural requirements.
Review of Other Secondary Pioina System Weld Radicarach5
Completed welds in the Auxiliary Feedwater (CA) system, the Blowdown
Recycle (BB) system and the Wet Lay-up Recirculation system were
radiographed as required by code to determine acceptability for service.
These welds were primarily dissimilar metal, nozzle to pipe welds.
Welding a short transition piece of stainless steel pipe to the nozzles
allowed for the fabrication of the dissimilar metal welds to be made
while the replacement S/Gs were still in the onsite manufacturing
facilities.
Making those welds at this time eliminated the need to
fabricate them inside containment where accessibility contributed to
welding problems at Catawba.
Weld No.
Size
Description
Stats
BB1F-45
3"x.438"
Pipe to Nozzle
Ac cepted: 12/12/96
BB1F-22-59
3"x.438"
Pipe to Nozzle
Accepted: 12/10/96
BBW1FW4-23
3"x.438"
Pipe to Nozzle
Accepted: 12/13/96
BBW1FW4-24
3"x.438"
Pipe to Nozzla
/.ccepted: 12/13/96
CAW 1FW18-1
6"x.719"
Pipe to Nozzle
Accepted: 12/11/96
ENCLOSURE 2
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11
I
Weld No.
SiZg
Descriotion
Status
CA1FW15-1
6"x.719"
Pipe to Nozzle
Accepted: 11/20/96
BB1F-267
3"x.438"
Pipe to Nozzle
Accepted: 01/07/97
BB1F-277
3"x.438"
Pipe to Nozzle
Accepted: 12/18/96
BB1FW5-25
3"x.438"
Pipe to Nozzle
Accepted: 11/21/96
Through this review the inspector ascertained that the radiographs
i
showed the welds were properly evaluated. That radiographic density, and
penetrameter sensitivity were sufficient to display the penetrameter
image and the specified hole which are essential indications of the
radiograph's image quality.
In addition, the inspector noted that
i
penetrameter location met the intent of code requirements that the films
were free of artifacts, chemical streaks and equipment related problems -
4
observed on previous inspections.
This review also disclosed that use of Dimetric TIG machines used to
)
'
fabricate the CF welds did not meet the licensee's expectations of
increased weld production with a corresponding reduction of rejections.
As the radiographs indicated, a large number of welds exhibited various
amounts of porosity which in many cases exceeded code allowable and some
lack of fusion.
In an effort to alleviate this problem the licensee
made several adjustments to the welding technique without success.
Production welding was stopped. and technical experts were contacted for
assistance.
Weld samples were fabricated on test coupons from the same material and
filler metal. Some of the changes implemented at this time included:
1) a switch in the technique used to deposit weld metal
i.e.. weave to
stringer bead. 2) change in the position of the tungsten electrode with
respect to the weld groove. 3) a change in the travel speed and an
incremental increase in amperage.
Following the close of this
inspection on January 10, 1997, the licensee's cognizant engineer
informed the inspectors by telephone. that one of the test samples which
has been welded with Argon gas only instead of the Argon-Helium mixture
was relatively free of porosity indications. Also at this time, the
inspectors were informed that test results indicated the porosity
indications were identified as calcium inclusions whose origin was still
under investigation. The licensee indicated that additional test
samples would be welded using Argon gas only.
If successful, this would
verify that the problem was not technique or welder related and as such.
production would begin again using the single gas procedure.
Repairs on
'
completed welds were being made using the manual TIG process.
ENCLOSURE 2
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12
c.
Conclusions
The low rejection rate of dissimilar metal welds on the secondary )ipe
systems, i .e. , BB, BW and CA, attained during this SGRP suggests tie
licensee has made significant programmatic improvements in this area
which would include preparation. allocation of technical resources and
training.
However, the relatively high rejection rate of machine
fabricated CF welds suggests that preparation for machine welding on a
production basis was inadequate.
Stopping of weld production to
investigate and correct the problem was ap]ropriate and had good support
from management.
However, if deletion of -lelium gas from the procedure
resolves the porosity problem, it would be reasonable to conclude that
this problem should have been identified and corrected before moving
into the production phase of the operation.
The quality of radiographs in terms of setup preparation and film
development exhibited a significant improvement over that observed on
previous inspections at other Duke nuclear stations.
M2
Maintenance and Material Condition of Facilities and Equipment
M2.1 Boraflex Testina
a.
Insoection Scoce (IP 62703)
The inspectors witnessed portions of the licensee activities to assess
the condition of the boraflex neutron absorber material in the Unit 2
S)ent Fuel Pool Storage Racks.
The testing was performed to evaluate
tie potential for gamma radiation-induced shrinkage of the absorber
material and the long term performance effects as a result of gamma
exposure and the wet pool environment. The McGuire Unit 1 and Unit 2
fuel storage racks were provided by Westinghouse.
b.
Observations and Findinas
The inspectors conducted observations of )ortions of the storage cell
testing.
The inspectors noted that fuel ]uilding ventilation was in
operation with operable radiation monitoring equipment.
Control room
indications of fuel pool level and fuel pool boron concentration met TS
requirements. The test equipment was calibrated onsite using a special
-
calibration / test cell provided by the vendor.
In the licensee's res)onse to NRC Generic Letter 96-04, Boraflex
Degradation in Spent ruel Pool Storage Racks, the licensee committed to
perform in-situ testing of the fuel storage racks. The licensee
consulted Northeast Technologies to perform the testing. The
examination was conducted in accordance with PT/0/A/4550/36. Controlling
Procedure for SFP Storage Rack Boraflex Examination, using the Boron
Areal Density Gage for Evaluating Racks (BADGER) test equipment. This
equipment was developed by Northeast Technologies under contract with
ENCLOSURE 2
.
.
.
. -
13
the Electric Power Research Institute. The test also provided
additional data to validate the BADGER device for in-situ boron-10
determination in borated fuel storage pools.
The licensee selected several high dose storage cells to quantify Boron-
10 areal density, gaps, thinning, and absorber end elevations. The
a
testing was performed by Northeast Technologies under direct supervision
of a McGuire site sponsor.
Test equipment performance was adequate.
l
The licensee identified some gaps and potential thinning but none was
indicative of gross degradation.
Qualitative analysis of the test data
will be performed by Northeast Technologies and a final report is
'
expected to be issued in approximately 8 weeks. At that time. Duke is
expected to compare the information with current criticality
calculations and make any necessary revisions to ensure storage rack
operability.
,
c.
Conclusion
The inspectors concluded that the preparation and coordination of the
testing was good.
Previous attenuation test results reviewed by the
inspector coupled with the current data including spent fuel pool silica
1
concentration provided good preliminary evidence that no gross boraflex
absorber degradation has occurred.
Final test results are expected in
six to eight weeks.
M2.2 1RN171B Service Water to Diesel Coolina Water Heat Exchanaer Sucoly
Valve Failure
a.
Insoection Scone (IP 61726)
During VOTES testing of the Service Water to Diesel Engine Cooling Water
Heat Exchanger Supply Valve. 1RN171B. the licensee noted abnormally high
running loads and the valve failed to stroke as expected.
The safety-
i
related valve is designed to open to allow essential cooling water to
the Diesel Generator Engine Cooling Water System heat exchanger. The
valve was being tested because of indications of rapid degradation based
on motor current analysis and thrust data. According to licensee data,
the valve failed to operate due to an abnormally high gearbox load.
Based on discussions with the licensee. no preventative maintenance
procedures or activities were recommended for this type gearbox.
.
-
b.
Observations and Findinas
The licensee responded to the valve failure by placing the valve in the
open position and removing power to allow continuous flow through the
heat exchanger to ensure adequate cooling would be available if
necessary during an event.
The licensee evaluated the effects of
continued service water flow on the heat exchanger and determined that
the continuous flow would not cause any immediate operational problems.
Nonetheless, the licensee developed a temporary modification to reduce
ENCLOSURE 2
.
.
. .
.
i
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14
the likelihood of accelerated heat exchanger fouling due to continuous
]
cooling water flow.
,
!
The temporary modification removed the automatic functions from 1RN171
,
and transferred these signals to heat exchanger outlet flow control
,
valve. 1RN174.
Valve IRN174 was then placed in the closed position.
!
This allowed cooling water flow to be shut off when the diesel generator
was not in operation. .The normal inlet valve was secured open with
power removed.
Testing was performed to verify component performance by
simulating a safety injection, emergency diesel generator and auxiliary
feedwater starts in accordance with the temporary modification test
i
package.
Service Water System flow balancing was not necessary since
overall system flows were not affected. The applicable standards
continued to be met since both valves were butterfly valves with
electric motor operators supplied with safety power, were aeriodically
i
tested to meet a 60 second stroke time, and there was no 31ysical
'
changes to the valves. The temporary modification was scleduled to be
[
removed during the Unit 1 E0C11 outage,
t
The inspectors also sampled affected procedures and verified that
j
necessary procedure revisions had been completed.
!
c.
Conclusions
i
The inspectors noted good response by station Maintenance and
l
Engineering to implement necessary temporary modifications to return the
system to operable status without compensatory measures.
No Unreviewed
'
Safety Question associated with this modification was identified.
,
M8
Miscellaneous Maintenance Issues (92902)
M8.1
(CLOSED) VIO 50-369.370/96-08-02:
Inadequate Containment Annulus
Surveillance Procedure.
This violation addressed the licensee's failure
to ensure the requirements of a TS surveillance test were met.
Specifically, procedures used by the licensee to conduct surveillance
testing on containment annulus ventilation were not adequate to ensure
that heaters remained operable for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. The inspector
verified that applicable procedures were corrected to ensure compliance
with TS requirements. The inspector also noted that the licensee
planned to conduct further study of the broader issue under a
comprehensive plan to review the implementation of TS surveillance
testing at the station. The inspector considered the immediate
corrective actions taken for the violation adequate and the scope of the
surveillance testing review to identify other problems appropriate.
This violation is closed.
ENCLOSURE 2
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. . . _ .
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15
III. Enaineerina
El
Conduct of Engineering
t
E1.1 Generic Letter (GL) 89-10 Proaram Imolementation
a.
Insoection Scope (TI 2515/109)
!
!
This inspection provided an assessment of the licensee's implementation
i
of GL 89-10. " Safety-Related-Motor-0perational Valve Testing and
Surveillance". The licensee had notified the NRC that implementation
was complete in a letter dated December 28. 1995.
_
l
The inspection included evaluations of: the scope of MOVs included, the
calculations of the design basis differential pre::sure, the
determinations of MOV settings and verifications of MOV capabilities.
!
the periodic verification of MOV capabilities. the MOV post maintenance
i
and post modification testing requirements, assuring against pressure
'
locking and thermal binding, and trending of MOV performance.
The inspectors conducted their assessment through a review of the
j
licensee's GL 89-10 implementing documentation and through interviews
}
with licensee personnel.
The documents reviewed included: "NRC Generic
Letter 89-10 Program Plan." Rev. 4: " Guideline for Performing Motor
]
03erated Valve Reviews and Calculations". DPS-1205.19-00-0002. Rev. 0:
'
" Engineering Support Program. Generic Letter 89-10. Motor Operated
Valves." Rev. 0: " Evaluation of Rate-of-loading Effects". DPC-1205.19-
00-0002. Rev. 0: DPC-1205.19-00-0001. Rev. 1. " Evaluation of Stem Factor
.
and Stem C.O.F. Assumptions:" a summary matrix of margins available for
all MOVs in the GL 89-10 program: and the additional procedures,
calculations, test records etc.. referred to in the following
paragraphs..
i
b.
Observations and Findinas
1.
Scone of MOVs Included in the Proaram
j
The scope of valves originally in the licensee's GL 89-10 program
!
was reviewed previously by the NRC during Inspection 92-11. This
!
scope would be considered acceptable based on current NRC
-
positions.
During the current inspection the inspectors evaluated
,
the subsequent deletions of valves from the program to determine
,
if they had been adequately justified.
The deletions were
,
. identified by comparing the valves in the original program with
those in the current program.
The inspectors then assessed the
acceptability of the deletions through a review of justifications
l
given in MCC-1205.19.00-0011 and design and functional information
i
described in the FSAR and other licensee documents (including
l
piping and instrumentation diagrams).
The inspectors found that
the licensee had a satisfactory basis for each HOV deletion.
l
,
ENCLOSURE 2
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16
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A total of 121 valves had been removed from the scope of the
program that was previously reviewed by the NRC during Inspection
.
92-11. The inspectors found the basis for the removal of these
valves was satisfactory. They had been removed because they were
identified not to have a safety function, the power had been
removed from the valves, or because they were excluded by
Supplement 7 to GL 89-10. " Valve Hispositioning in Pressurized
i
Water Reactors". Some valves were also removed from the program
,
'
because they were determined not to be gate, globe, or butterfly
i
valves. The MOV program remained very large with a total of 425
i
MOVs (consisting of 189 gate.136 globe, and 100 butterfly
valves).
.
r
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2.
Revised Desian-Basis Differential Pressure Calculations
,
!
{
In inspection 95-06. NRC inspectors noted that the licensee was
revising the calculations of MOV maximum differential pressure and
<
stated that a further NRC review of the calculations would be
,
performed.
That further review was performed during the current
i
inspection. The inspectors reviewed the differential pressure
i
,
!
calculations (such as MCC-1223.42-00-0026 for the auxiliary
feedwater svstem) the calculation reference documentation, and
'
the applicable system flow drawings. The inspectors found that
'
satisfactory calculations had been completed for all systems.
4
4
!
Inspection 95-06 particularly noted that the maximum differential
pressure calculation for valves NI-184B and NI-185A did not
i
include pressure locking effects. The inspectors found that these
I'
<
valves had been modified with a bonnet equalization line (drawing
'
MCFD-1562-03.01) to preclude pressure locking. The inspectors
found that the calculation (MCC-1223.12-00-0017. Rev. 4) had been
appro)riately revised after inspection 95-06 and that it accounted
3
for t1e system configuration applicable to use of these valves.
i
1
.
3.
Determinations of Settinas and Verifications of Cao$ilities for
Gate Valves
'
,
,
The inspectors selected and reviewed calculations test data and
,
1
evaluations for the following sample of gate valves, in order to
assess the licensee's validation of calculation assumptions and
j
,
their determinations of MOV settings and capabilities:
1CA0042
Auxiliary Feedwater System Pum) 1B Isolation Valve
1
1KC0018
Reactor Building Isolation of ion-Essential Header
1NC0031/3/5 PORV Block Valves
1ND0019
ND Pump Suction Isolation Valve
1NI0009
NC Cold Leg Injection from NV
.
1NIO100
Suction Valve from RWST to NI Pumps
1NV0244/5
Containment Isolation for the NV Charging Header
4
2KC0018
Reactor Building Isolation of Non-Essential Header
ENCLOSURE 2
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1
17
2NC0031
PORV Block Valve
2NIO100
Suction Valve from RWST to NI Pumps
2NIO333
Safety Injection Pump Suction Crossover from NV
2NV0095
Containment Isolation for the Reactor Coolant Pump
Seal Return
)
The inspectors findings were as follows:
l
MOV Sizina and Switch Settinas
i
McGuire gate valve thrust calculations typically utilized the
!
standard industry equations. Mean seat diameter was used to
calculate valve seat area.
Valve factors were based on in-plant
test results or results from other industry sources. The licensee
used statistical methods to evaluate the effect of MOV performance
l
uncertainties on available margin.
i
Valve Factors and Grouoina
i
i
McGuire dynamically tested 53 of their 189 gate valves. Measured
'
valve factors were used for the gate valves that ure dynamically
'
tested. The licensee gathered in-plant and other industry valve
i
factor data, and formed 17 gate valve groups to justify the
applied valve factors for non-dynamically tested gate valves. The
inspectors found that the valve factor justifications for each
valve group and the current setup of the MOVs was adequate for
!
i
design-basis capability.
However, they noted that the bases for
,
,
4
the valve factors established for a few groups was weak, due to
l
the limited quantity or quality of the su) porting data. These
i
4
weaknesses were of greater concern when tie valve groups contained
valves with marginal calculated design basis capabilities. The
l
l
licensee provided actions to address the cases which the
'
i
inspectors considered more significant through Problem
!
Investigation Process (PIP) 0-M96-3542.
Discussions of the
j
weaknesses noted and the licensee's actions are as follows:
i
'
a
Group B consisted of 24 Aloyco split-wedge gate valves of
!
4
different sizes. The inspectors noted that the four 12-inch
valves in this group were marginal and that none had been
dynamically tested.
Based on the lack of dynamic test data and
t
i
.
.
the marginal capabilities for these valves. the inspectors
,
considered that the justification for the assumed valve factor was
i
weak. The licensee established an action item in PIP 0-M96-3542
to initiate work requests to improve the margin for these valves
and to obtain additional test data to support the group valve
'
factor.
!
Group D consisted of eleven 4-inch Aloyco split-wedge gate valves.
,
The licensee initially based the group valve factor on a single
,
.
'
dynamic test. The inspectors observed that this was inconsistent
"
i
ENCLOSURE 2
<
!
D
,,a
n.n
--,
,
.+.
- , , .
-w
,
-e,
- - -
-e-
-,
, , , , - - -
_ _ _ _
!
18
'
i
with GL M-10 Sup)lement 6. which indicates that group valve
l
factors
Fculd be ]ased on tests of nominally 30%. but no less
-
than 2 v Wes from the group. The licensee obtained additional
i
test datu from similar valves tested at another plant which
satisfactorily supported the valve factor used for the group.
Group F consisted of eight 6-inch Borg Warner gate valves. The
valve factor for this group was based on test results from two
similar valves at another of the licensee's plants. The
-
inspectors considered this weak, as GL 89-10. Supplement 6.
'
recommended testing at least three valves from groups of this
si ze.~
Further, the inspectors found that the capability of MOV
!
1CF0129 in this group was marginal. The licensee established an
action item in PIP 0-M96-3542 to upgrade this MOV at the next
refueling outage.
i
Group H consisted of eight 4-inch Borg Warner gate valves whose
safety function is to close to isolate a faulted steam generator.
c
The inspectors found that the valve factors used in determining
i
the closing torque switch settings for these valves were
nonconservative, considering the licensee's limited in-plant test
data.
However, the torcue switches were bypassed for 95% of the
i
closing valve stroke anc the ins)ectors found that the valves were
capable of completing at least tlat much of the closing stroke.
,
They considered this marginally sufficient isolation capability.
!
The licensee's Improvement Plans indicated these MOVs were to be
i
'
replaced.
Group K consisted of six 3-inch Borg Warner gate valves which
j
functioned as the PORV block valves. The group valve factor was
based on test results obtained from similar valves at the
licensee's Catawba plant. This data was not directly applicable
to the McGuire valves. as it was obtained under pumped flow
i
conditions whereas the McGuire block valves experience blowdown
flow.
Based on their available output thrust at the current
,
torque switch settings and on industry blowdown test data from a
similar valve, the inspectors did not have an immediate
l
operability concern regarding these valves.
However, they
-
considered the ca) abilities of these MOVs to be marginal.
The
licensee establisled action items in PIP 0-M96-3542 to uagrade
i
-
these MOVs at the next refueling outage, and to either o)tain
!
additional data from other sources or apply the EPRI MOV
Performance Prediction Methodology to strengthen support for the
group valve factor.
!
Another weakness noted by the inspectors was that the licensee,
sometimes included multiple test data points from a given valve in
statistically analyzing the test data from a group of valves.
This could bias an analysis. The licensee established an action
i
item in PIP 0-M96-3542 to revise Duke Power Specification DPS-
,
l
ENCLOSURE 2
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19
1205.19-00-0002 to require the number of data points per valve to
be equal to avoid biasing the group valve factor in favor of
valves with more than one data point.
Load Sensitive Behavior
Gate and globe MOVs that were dynamically tested used the measured
load sensitive behavior values.
DPC-1205.19-00-0002. " Evaluation
of Rate-of-Loading Effects," specified a bias margin of 6.96% and
a standard deviation of 13.8% to account for the effects of load
sensitive behavior for GL 89-10 MOVs that were not dynamically
tested.
However, the licensee had noted a continuing improvement
in load sensitive behavior performance across all of the Duke
nuclear facilities. Therefore. Duke was in the process of
revising the load sensitive behavior justification.
Duke
personnel determined that, based on data from all of the Duke
facilities over the last 3 years, a mean of 4% with a standard
deviation of 10.56% represented the reduction of thrust at torque
switch trip under dynamic conditions as a result of load sensitive
behavior.
In response to inspectors' questions, the licensee
3 resented information to demonstrate that the load sensitive
3ehavior of the MOVs at the McGuire facility was consistent with
the overall Duke assumption of the effects of load sensitive
behavior.
The licensee established an action item in PIP 0-M96-
3542 to incorporate the new plant-specific information in Duke
corporate document DPC-1205.19-00-0002.
Stem Friction Coefficient
l
McGuire's original calculations assumed a stem friction
coefficient value of 0.15 in determining actuator output thrust
capability.
This value was selected based on a review of in-plant
j
test data from all Duke facilities.
Based on an updated review.
'
l
the licensee determined that the original assumptions for McGuire
were somewhat nonconservative, and the thrust calculations were
revised to incorporate the use of a 0.20 stem friction
coefficient, except where specific test data supported a lower
value.
The inspectors found this to be accepta]le.
However, the
inspectors noted that certain MOV program documents still cited
the use of a 0.15 stem friction coefficient. The licensee
established action items in PIP 0-M96-3542 to revise McGuire
calculation MCC 1205.19-00-0003 and Duke corporate document DPS-
1205.19-00-0002 to incorporate the new information for gate and
The licensee also established an action item in PIP
0-M96-3542 to discuss with plant personnel that the results of its
analysis of stem friction coefficient may not be effective and to
re-enforce the importance of performing high quality stem
lubrication. The licensee stated future plans to obtain more
reliable torque data than was currently available from the spring
pack curves used for some globe valves.
ENCLOSURE 2
)
20
Diaanostic Eauioment Uncertainties
NRC inspection 95-06 determined that McGuire personnel were not
accounting for V0TES diagnostic equipment uncertainties in the
open direction when measurements were outside the sensor
calibration range. These errors can become very large if
measurements are significantly outside the calibration range. At
the time of that inspection. McGuire personnel initiated PIP 0-
G95-0295 to resolve the issue. The methods identified in this PIP
included a review of all diagnostic testing that existed at that
time.
Subsequent to the inspection. McGuire developed DPC-
1205.19-00-0003.'" Evaluation of MOV Open Direction Issues," Rev.
O, which: 1) established the basis for the methods used to
evaluate the issue, and 2) provided the screening of McGuire's
MOVs.
The licensee also revised their diagnostic procedures to
obtain tension in the calibration range (when needed) which
reduces the applicable diagnostic error in the open direction.
After review of the actions taken to address this concern, the
inspectors considered this issue closed for McGuire.
Desian-Basis Caoability
At the outset of the inspection, the licensee presented a method
for calculating the thrust required to operate non-dynamically
tested MOVs that relied, in part, on perceived risk contribution
(as established using PRA techniques).
In this method, the risk
assigned to a valve was used to select a confidence level which
was used to determine the number of standard deviations to be
applied to each uncertainty in MOV performance (i.e.
variation in
valve factors, load sensitive behavior, and diagnostic equipment
uncertainty).
In effect. this resulted in operability being
determined on the basis of FRA risk. The inspectors noted that
this was considered weak and was contrary to GL 91-18, which
indicates that PRA or risk should not be used in operability
decisions.
During the inspection, the licensee reconsidered their methodology
and re-evaluated the current setup of all its GL 89-10 MOVs.
a) plying a 95% confidence level to uncertainties.
Additionally,
t1ey established action items in PIP 0-M96-3542 to 1) revise
specification DPS-1205.19-00-0002 to recuire minimum thrust
calculations to be based on a 95% conficence level for
uncertainties. 2) revise their calculations to provide minimum
thrust requirements based on the 95% confidence level
3) include
guidance on consideration of )otential MOV aging and degradation
effects, and 4) ensure that t1e work documents for the upcoming
Unit 1 outage reflect the appropriate valve setup requirements.
The licensee did not apply the risk-based methodology to the
butterfly valves or Kerotest globe valves. With respect to the
ENCLOSURE 2
21
non-Kerotest globe valves, the licensee established an action item
in PIP 0-M96-3542 to revise the thrust calculations to be more
consistent with accepted deterministic methods.
To allow a comparison to deterministic methods, the licensee
calculated available valve factors that were based on the thrust
available in the closing and opening directions.
These
calculations were adjusted to account for diagnostic equipment
uncertainty. torque switch repeatability, and bounding load
sensitive behavior assumptions.
MOVs with less than a 0.50
available valve factor were then reviewed individually by the
inspectors. The inspectors did not identify any immediate
operability concerns. However, the inspectors considered the
following MOVs to be marginal:
INC0031
INC0033
1NC0035
1ND0019
1NIO100
INV0244
1NV0245
2NC0031
2NC0033
2NC0035
2NV0095
The licensee agreed and noted that several of these MOVs were
already scheduled for margin improvements in the near future. The
licensee established an action item in PIP 0-M96-3542 to ensure
that these MOVs are upgraded at the next refueling outage.
l
4.
Determinations of Settinos and Verifications of Caoabilities for
The GL 89-10 ]rogram at McGuire included 134 small globe valves
i
manufactured )y Kerotest and two globe valves manufactured by
Walworth-Aloyco.
The licensee was not able to obtain reliable
diagnostic data from dynamic testing of the Kerotest globe valves
at McGuire.
Instead, they conducted a testing program at their
flow loop researc.h facility to evaluate the ]erformance of the
Kerotest globe valves under pumped flow and ) lowdown conditions.
The inspectors reviewed the results and evaluation of the Kerotest
globe valve test program, as documented in DPC-1205.01-00-0001.
Rev. 1. " Evaluation of Flow Loop Tests of Kerotest Valves." and
DPC-1205.01-00-0002. " Evaluation of Kerotest Valves in Steam
-
Blowdown Conditions." The inspectors noted that the vendor's
method for predicting thrust recuirements for the 2-inch soft-seat
design Kerotest globe valves hac been found nonconservative, based
on full flow differential pressure test results.
The licensee had
recognized this nonconservatism and had included additional margin
for these particular valves in their calculations depending on
service application. The licensee established an action item in
PIP 0-M96-3542 to include specific guidance in DPC-1205.01-00-0001
for the additional margin and for consideration of service
application when applying the Kerotest thrust prediction method to
ENCLOSURE 2
_ __ . . ____ .- _ _ . _ _ _ _.. _ . _ _
_ ._ _ _
_ _ _ _ _ .
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_ _ _
!F-
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22
i
the 2-inch soft-seat valves in 150 psi or less water service
!
j
applications.
,
The licensee tested one of the two globe valvas manufactured by
i
Walworth-Aloyco under dynamic conditions. The results supported
i
the 1.] valve factor assumed for the other globe valve
!
manufactured by Walworth-Aloyco.
.
"
.
<
j
The inspectors considered the licensee's thrust requirements for
l
the GL 89-10 globe valves to be acceptable.
,
5. -
Determinations of Settinas and Verifications of Cao6h11ities for
2
-
i
Butterfly Valves
4
.
The McGuire GL 89-10 program included 100 butterfly valves that-
,
4
were separated into 14 groups. The licensee conducted dynam!c
j
'
tests on 52 butterfly valves and applied the test results, where
applicable. to the valves that were not dynamically tested.
l
i
<
'
The inspectors reviewed the licensee " validation calculations"
l
which evaluated test data to demonstrate the capabilities of
butterfly valves to perform their design basis functions. The
l
inspectors found that the test data and evaluations documented in
i
'
these calculations demonstrated generally satisfactory settings
l
]
and capabilities for the licensee's butterfly valves.
However.
!
the inspectors noted weaknesses for three groups (E. I, and K) of
i
'
j
nuclear service water system valves:
!-
Group E consisted of twelve (six per unit) 10-inch, class
i
.
i
150 rdel NMK 11. Henry Pratt butterfly valves.
These
i
valves were addressed by validation Calculation MCC-1205.19-
l
00-0030. Rev. O. The test data used to establish the
j
i
settings and capabilities of these valves was considered
l
weak by the inspectors, as it was from static and dynamic
tests performed on much larger (16-inch) valves.
4
i
Additionally, relying on this data, the calculated
,
]
capabilities of several of these valves only marginally
,
)
exceeded the design basis requirements (by 1% or less).
l
Group I consisted of four (two per unit) 6-inch, class 150.
.
-
model 7620. Fisher Controls butterfly valves. The test data
!
'
1
used to establish the settings and capabilities of these
j
valves was considered weak, as it was not quantifiable by
!
'
the licensee's normal evaluation methods and the dynamic
j
i
testing had been performed at only about 60% of design basis
t
differential pressure. The calculation employed what was
!
!
referred to as a "non-typical validation" approach in
i
assuring the capabilities of these valves.
!
!
i
L
ENCLOSURE 2
,
.
.
-,....c
,w
,-
v~-,
im-
4-,,-a-
-t-a-
--- -
- -
- - - -
---
m-e--
,-, se
e
v
.-c+
w
4
m
v?
-~---4
_
.
_
!
!
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23
,
Group K consisted of four (two per unit) 8-inch, class 150.
l
.
l
model NMK11. Henry Pratt butterfly valves. The inspectors
i
J
found the data which the licensee had used to establish the
!
settings and capabilities of these valves was weak, as it
!
was based on tests performed on tests of much larger valves
!
(16- and 20-inch). The licensee had statically and
!
dynamically tested the group K valves but had determined
that the test results could not be relied upon fer
quantitative evaluations.
Further the licensee did not
,
i
qualitatively demonstrate the capabilities of the valves
!
^
through the group K valve dynamic tests, as they were
performed at only about 60% of design basis differential
i
pressure.
j
.
t
!
Note: Group K Valve IRN1718. a nuclear service water supply
1 solation valve to the diesel generator heat exchanger, was
l-
tested during this inspection and failed to perform
,
-
i
satisfactorily.
Refer to paragraph M.2.2 for details on
'
l
this failure.
i
To address the above weaknesses, the licensee established an
!
I
action item in PIP 0-M96-3542 to perform dynamic testing with
i
diagnostics on four Group E valves and all valves in Groups I and
!
K.
Additionally, the action item required raising the torque
'
!
switch settings for the marginal Group E valves to provide
"
increased assurance they would perform their design basis
,
j
functions.
4
!
'
6.
Periodic Verification
',
!
l
The licensee incorporated MOV periodic verification requirements
l
l
into the MOV preventive maintenance (PM) Program. The inspectors
j
reviewed this PM Program, dated March 26. 1996. The program was
!
'
computerized and specified stem lubrication, diagnostic test, and
(
4
actuator inspection intervals. The inspectors concluded that the
!
PM Program for GL 89-10 MOVs was well defined.
During the
inspection the licensee was in the process of establishing
i
criteria for determining when periodic MOV dynamic testing would
t
!
be performed.
'
,
.
.
The licensee's periodic verification actions were considered
!
1
adequate for closure of the GL 89-10. The NRC may re-assess the
l
<
!
!
licensee's long-term periodic verification program as part of its
review of GL 96-05. " Periodic Verification of Design-Basis
i
Capability of Safety-Related Motor-0perated Valves", dated
,
4
September 18, 1996.
j
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!
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t
ENCLOSURE 2
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1
!
I
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_
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.
24
7.
Post Maintenance Testino (PMT)
The inspectors reviewed the licensee's PMT requirements and
!
guidance, which were specified in their PMT Program, dated May 29,
i
1996, and their Work Process Manual. Section 501 dated August 21,
1995. Additionally, the inspectors reviewed the PMT recorded for
!
W0s 94076860 (packing leak repair). 95005056 (adjust packing).
!
94076858 (adjust packing), and 95015982 (adjust packing). The
i
inspectors found the licensee's PMT acceptable for GL 89-10
l
closure; however, weaknesses were noted and were addressed by the
licensee as described below.
The PMT program only required stroke testing following packing
i
adjustment if the packing torque was not increased above that
present in the last base line diagnostic test.
From their review
of WO PMTs the inspectors found that the MOV engineer assigned a
more rigorous PMT than specified by the PMT Program. A diagnostic
!
test or MPM (Motor Power Monitor) test was required following
!
packing adjustment in addition to stroke testing. The inspectors
1
concluded that the PMT requirements specified in the PMT Program
l
for packing adjustments were lacking in detail.
During the
!
inspection the licensee provided an action item in PIP 0-M96-3542
i
to resolve this. The item specified updating of the PMT Program
to provide details of the PMT to be performed and review WPM 501
for consistency.
The licensee sometimes performed an MPM test in lieu of a
diagnostic *rce measurement as PMT following packing adjustment
or replacement. The inspectors observed that this should be
supported by data which demonstrates that the MPM testing is able
to discern unsatisfactory increases in packing forces.
During the
j
inspection the licensee documented in PIP 0-M96-3542 the need to
provide justification to support packing adjustment and
i
replacement without performing diagnostic testing as a PMT as
!
when MPM is used in place of diagnostic testing.
l
8.
Post Modification Testina
The licensee assigned post modification test requirements for MOV
modifications on a case by case basis.
The inspectors reviewed
i
modifications MGMM-7097 (Change 2KC0003 Actuator Speed), dated
April 10. 1996: MGMM-7155 (Replace 1NIO332 Actuator) dated
January 3, 1996: and MGMM-7305 (Replace INS 0020 Actuator). dated
i
October 2, 1995. The inspectors found that the licensee had
i
implemented acceptable post modification testing.
l
!
9.
Pressure Lockina and Thermal Bindina
l
!
The inspectors reviewed the evaluation of gate valves susceptible
i
to pressure locking and/or thermal binding which the licensee had
!
t
ENCLOSURE 2
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t
!
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- -- -
--
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- . - - - - .
____
_ _ _
.. _ __ _ . _ __ _ - _
_
_ _ -
__ _
t
.
1
25
,
&
completed in response to GL 95-07. " Pressure Locking and Thermal
i
Binding of Safety-Related Power-Operated Gate Valves".
In its
letters dated February 13 and July 31, 1996, the licensee
identified valves that were susceptible to pressure locking and/or
thermal binding and corrective actions.
l
The licensee's GL 95-07 submittals stated that analytical methods
I
were utilized to demonstrate that the actuators on valves
1(2)FW0027A. 1(2)ND0058A. 1(2)NIO136B. 1(2)NS00018. 1(2)NS0018A.
1(2)NS00388 and 1(2)NS0043A could develop adequate thrust to
>
overcome pressure locking. The inspectors were informed that
analytical methods were used as short term corrective action and
that modifications were scheduled to be implemented during the
!
upcoming Unit 1 and 2 refueling outages to eliminate the potential
!
for pressure locking. The inspectors reviewed Calculation MCC-
l
1205.19-00-0052. "GL 95-07 Pressure Locking & Thermal Binding
l
Evaluation." Rev.14. and verified that analytical methods were
adequate to demonstrate operability for present plant conditions.
Additionally, the ins)ectors verified the planned modifications
i
were )rescribed in Pl? 0-M96-0460. The licensee planned to update
i
its G_ 95-07 response to indicate that these valves would be
modified to eliminate the potential for pressure locking. The
inspectors considered that short and long term corrective actions
i
to preclude pressure locking in these valves were acceptable.
l
The licensee *s GL 95-07 submittals stated that testing would be
utilized to provide reasonable assurance that valves 1(2)NS0012B,
1(2)NS0015B.1(2)NS0029A and 1(2)NS0032A would not pressure lock.
I
The inspectors were informed that this testing was complete but
!
was only used as short term corrective action and that procedures
were being revised to eliminate the potential for pressure
i
locking.
The inspectors verified the plans to revise the
procedures to preclude pressure locking were prescribed in PIP 0-
M96-0460.
The licensee planned to update its GL 95-07 response to
1
indicate that procedures would be modified to cycle these valves
following operation of the applicable containment spray pump to
,
eliminate pressure in the valves' bonnets. The inspectors
considered that short and long term corrective actions to preclude
pressure locking in these valves were acceptable.
'
The licensee's GL 95-07 submittals stated that an analytical
method was used to demonstrate that the actuators on valves
1(2)LD0108A and 1(2)LD0113B could overcome pressure locking.
The
inspectors reviewed the associated Calculation MCC-1205.19-00-0052
and noted that the margin between the thrust required to overcome
pressure locking and actuator capability for this type of valve
was minimal.
During the inspection the licensee reevaluated the
use of this analytical method and concluded that it would be
prudent to modify these valves to eliminate the potential for
pressure locking. The inspectors verified the plans to modify
ENCLOSURE 2
l
__ _ _ _ _ . _ _
.
!
s
l
26
!
these valves to preclude pressure locking were prescribed in PIP
f
0-M96-0460.
The licensee planned to update its GL 95-07 response
to indicate that these valves would be modified. The inspectors
considered that short and long term corrective actions to preclude
pressure locking in these valves were acceptable.
,
The licensee's GL 95-07 submittals stated that valves 1(2)NI0009A
and 1(2)NI0010B open with high head injection Jump discharge
!
pressure acting on one side of the valve whic1 prevents the
valves from pressure. locking.
The inspectors reviewed emergency
-
safeguards feature test results and noted that it took
j
approximately two seconds for high head injection pum s to develop
j
full discharge pressure after receiving.a start signa .
Valves
NI0009A and NI00108 are normally shut and receive an open signal
l
at the same time the high head injection pumps receive a start
i
signal. Therefore, valves NI0009A and N10010B could o)erate at
locked rotor conditions for several seconds.
During t1e
!
inspection the licensee reevaluated this analysis and concluded
that it would be prudent to modify these valves to eliminate the
potential for pressure locking.
The inspectors verified the plans
l
to modify these valves to preclude pressure locking were
prescribed in PIP 0-M96-0460. The licensee planned to update its
i
'
GL 95-07 response to indicate that these valves would be modified.
The inspectors concluded that operation at locked rotor conditions
for a very short period of time may not adversely effect valve
operability but it is not a preferred long term corrective action.
As short term corrective action, the PIP 0-M96-0460 indicated the
licensee would periodically review emergency safeguards feature
test results to ensure that high head injection pumps develop full
discharge 3ressure within several seconds after receiving a start
signal.
T1e inspectors considered that short and long term
corrective actions to preclude pressure locking in these valves
were acceptable.
The licensee's GL 95-07 submittals stated that an analytical
method was used to demonstrate that actuators for PORV block
valves 1(2)NC00318, 1(2)NC0033A and 1(2)NC0035B could overcome
pressure locking. Th3 inspectors reviewed calculation MCC-
1205.19-00-0052 and noted that the margin between the thrust
required to overcome pressure locking and actuator capability for
-
this type of valve was minimal.
During the inspection the
licensee decided to reevaluate the use of this analytical method
as long term corrective action to preclude pressure locking.
The
licensee planned to update its GL 95-07 response to report the
results of this evaluation.
The inspectors concluded that use of
the analytical method was acceptable for short term corrective
action.
The PORV block valves were also determined to be susceptible to
thermal binding.
In order to resolve a GL 89-10 concern, the
ENCLOSURE 2
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27
licensee was planning to increase torque switch settings on these
valves which would increase the valves' susceptibility to thermal
binding.
During the inspection the licensee decided to reevaluate
corrective actions to prevent thermal binding for these valves.
The inspectors verified that this reevaluation was prescribed in
PIP 0-M96-0460. The licensee planned to update its GL 95-07
response to report the results of the evaluation.
With the exception of the PORV block valves. valves identified by
the licensee as susceptible to pressure locking have been
modified, are planned to be modified or procedures are to be
revised to eliminate the potential for pressure 'ocking.
The
inspectors independently reviewed pressure locking and actuator
capability calculations for selected MOVs and verified that the
licensee correctly calculated the thrust required to overcome
,
pressure locking and actuator output capability.
The NRC staff is
continuing its evaluation of these valves and others within the
scope of GL 95-07 as part of its review of the licensee's response
to potential pressure locking and thermal binding.
10.
Trendina
The inspectors determined that a satisfactory description of
failure and performance trending requirements was described in the
licensee's Engineering Support Program, Rev. O, section 4.6.
They
verified the licensee's im'lementation of adequate trending by
p
reviewing Jortions of the trend database.
Further, the ins)ectors
verified tlat the licensee performed and documented acceata)le
periodic reviews of MOV fal,ures.
The reports examined ay the
inspectors were McGuire Failure Analysis and Trending Program
Review 1/1/95 through 6/30/96, Yearly Review of Rotork Motor
Operator Failure (dated March 11. 1996). and Limitorque Annual
.
Maintenance Review (dated May 13. 1996).
The inspectors
considered the trending implemented by the licensee satisfactory.
,
11.
Strenaths
The inspector observed a number of strengths in the licensee's
implementation of GL 89-10.
Particular examples included:
.
The detailed packages of data and evaluations that were
.
developed for each valve group.
The strong management and personnel support that was
.
necessarily provided to complete a program encompassing the
,
number of MOVs present at McGuire.
The application of state of the art technology such as
.
torque test stands and motor power monitoring.
ENCLOSURE 2
28
Leadership in addressing industry problems such as ir1 creases
.
in actuator ratings.
c.
Conclusions
The licensee had met the intent of GL 89-10 in verifying the design
basis capabilities of their MOVs
Several weaknesses were identified in
data and justifications of assumptions used in the verifications.
The
licensee planned actions to resolve these weaknesses which the
inspectors determined to be more significant and they were appropriately
3rescribed in PIP 0-M96-3542.
The licensee was requested to notify the
iRC of the status of the PIP actions 60 days after the end of the
upcoming Unit I refueling outage (IEOC11) and at the completion of all
of the items by December 31, 1997.
The inspectors identified the
completion of these actions as Inspector Followup Item 50-369. 370/96-
11-01. Actions to Address MOV Weaknesses.
In addition to weaknesses, the ins)ectors noted several licensee
strengths. which are described in ).11 above.
One related NCV was identified, involving periodic MOV lubrication and
is described in paragraph E8.3.
Based on the NRC staff's review of the McGuire GL 89-10 program and its
implementation, and the actions established by the licensee in PIP 0-
M96-3542. the NRC staff is closing its review of the GL 89-10 program at
McGuire.
The completion of these licensee actions will be assessed as
part of the NRC staff's monitoring of the licensee's long-term MOV
program.
E2
Engineering Support of Facilities and Equipment
E2.1 Vital Area Access
a.
Inspection Scope
During routine tours of the protected area, the inspectors noted the
addition of a locked security fence at the Unit 1 exterior doghouse and
a security barrier near the Unit 1 Refueling Water Storage Tank.
The
'
inspectors recognized the barriers potential impact on operator
emergency response and initiated activities to evaluate operator
awareness of this change to a station structure.
b.
Observations and Findinas
To verify that necessary training and/or procedure revisions had been
completed the inspectors conducted interviews and reviews of the
station emergency procedures.
The inspectors noted that no formal
guidance or procedure revisions had been provided to Operations
personnel describing the physical change.
As a result of these initial
,
ENCLOSURE 2
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29
findings, the inspectors requested a co)y of the modification package to
implement the change and was informed tlat the installation of the
barrier had not been completed as a modification.
The inspectors reviewed the Duke Power Nuclear Station Directive 301 and
the McGuire Modification Manual and determined that the installation of
the temporary barrier met the temporary modification criteria.
Therefore the modification should have been processed in accordance with
the licensee approved temporary modification process as outlined in
Nuclear Station Directive (NSD) 301 which defined a physical change or
addition to a station's structures system or components for a finite
period as a temporary modification.
Since the change to the facility was not handled ir, accordance with the
Temporary Modification process. no formal training or )rocedure
revisions were provided to Operations personnel descri)ing the change.
As a result, some operator r( ponse times under certain accident
conditions were slightly affeu.ed.
The slight increase in operator
response times did not exceed UFSAR assumptions. The licensee
recognized the oversight and developed immediate Operations training
packages for shift operations personnel and a followup reading package
discussing the barrier and necessary actions to take in emergency
situations.
Unannounced drills were conducted and operator response
times were validated.
c.
Conclusions
Following additional reviews of the McGuire Modification Manual. the
inspectors concluded that the installation of the barrier located at the
,
'
Unit 1 exterior valve vault was a temporary modification and should have
been controlled and implemented in accordance with NSD 301.
The failure
to meet the requirements of the modification manual is considered a
Violation and will be documented as VIO 50-369/96-11-02: Failure to
follow procedure for temporary modification installation.
,
l
E4
Engineering Staff Knowledge and Performance
E4.1 Enaineerina Staff Knowledae and Performance
a.
Scope
-
The inspector reviewed the material upgrade processes to determine if
commercial grade purchased items were appropriately evaluated and tested
for use in safety related applications. Approximately 30 commercial
grade dedication (CGD) packages were reviewed. These included
mechanical and electrical items.
Applicable regulatory requirements
'
included 10 CFR 50 Appendix B. FSAR. and the following-
l
ANSI N45.2.13-1976. 0A Requirements for Control of Items and
Services for Nuclear Power Plants
ENCLOSURE 2
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30
RG 1.123. 0A Requirements for Control of Procurement of Items and
Services for Nuclear Power Plant
GL 91-05. Licensee Commercial Grade Procurement and Dedications
Programs
b.
Observations and Findinas
Commercial grade dedication evaluations were aerformed by the corporate
procurement engineering organization for all Juke Power Nuclear
Stations.
Station 3rocurement engineering resolved receipt inspection
inconsistencies.
T1e sample of CGD purchased items reviewed were
appropriately evaluated and tested for use in safety related systems.
Critical characteristics for the items were adequately identified and
verified.
Commercial grade vendor audits and surveys approariately
evaluated the critical characteristics referenced in the CG)
evaluations. Adequate documentation was maintained to validate the
items * CGD,
Traceability of parts and warehouse storage were adequate.
Critical characteristics were adequately incorporated into receipt
inspection requirements.
The use of standardized commercial grade
procurement acceptance procedures was a good practice and assured
consistent quality verification actions at receipt inspection
Deficiencies or inconsistencies in technical or quality requirements
identified in receipt inspection were adequately resolved by station
Procurement Engineering.
c.
Conclusion
The commercial grade dedication process was implemented effectively at
McGuire and was consistent with applicable regulatory requirements.
No
examples were identified in which unqualified material or components
were installed in safety related applications.
E4.2 Resolution of Procurement Problems
a.
Scope
The inspector reviewed station Procurement Engineering's resolution of
procurement related problems identified in the licensee's Problem
.
-
Investigation Program reports (PIPS).
These issues were not limited to
CGD purchases.
Applicable regulatory requirements included 10 CFR 50
Appendix B and 10 CFR 50 Appendix A.
b.
Observations and Findinas
The inspector noted one issue of concern related to the purchase of
Grinell hydraulic pipe supports (snubbers) for the Steam Generator
Replacement Project.
PIP 0-M96-2408, dated August 26, 1996, identified
that recently purchased Grinell snubbers were not qualified to the
ENCLOSURE 2
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31
original purchase specification's environmental parameters. The
parameters included a 350 F temperature and 2 X 10E8 rad value limit
which was not met by the manufacturer.
These specifications were also
not met for the Grinell snubbers originally installed in the plant,
approximately 150 snubbers. These snubbers were installed in safety
related systems including the Reactor Coolant (RCS), Chemical & Volume
Control. Safety Injection. and Component Cooling.
'
The purchase specification temperature parameter was of concern because
it was based on worse case environmental conditions in containment.
This condition occurred only during a Main Steam Line Break (MSLB). The
containment temperature was anticipated to reach a) proximately 326 F
after a MSLB. The original specification of 350
assured that the
snubbers would remain operable in this environment. The vendor informed
the licensee that the snubbers' polycarbonate hydraulic reservoir would
experience significant deyradation above 250
F.
This could result in
loss of seal and subsequent loss of hydraulic fluid which would make the
<
,
snubber unable to perform its design function.
'
A licensee test conducted in December 1980, evaluated the durability of
the snubbers subjected to high temperatures.
The test heated the
j
snubber in an oven in stages up to 400
F.
The snubber was disconnected
and not subject to any internal hydraulic pressure forces that may exist
on an installed snubber; therefore, the internal contribution to
reservoir deformation was not included.
Reservoir warpage and loss of
hydraulic fluid began at 285 F.
The test conclusion was that. "the
snubbers should be considered unsafe at any temperature near 285 F
since the reservoirs started to visibly disfigure (not leak) at that
point after brief exposure to elevated tem 3eratures." There was no
,
additional documentation which addressed t1e acceatability of the
snubber design or the motivation for perfcrming t1e test.
~
The operability evaluation in PIP 0-M96-2408 concluded that the design
of the snubbers was acceptable and that there was no operability concern
with the snubbers or associated piping.
The evaluation stated that
,
following the MSLB the design load case is over and the snubbers are no
,
longer required to be operable. That is, the snubbers are no longer
components important to safety. The licensee indicated that no dynamic
piaing loads were anticipated following the MSLB.
Although actuation of
a RCS V - Operated Relief Valve would induce a dynamic load, this
-
actuatie was not expected after a MSLB.
The Grinell snubber design appeared to conflict with 10 CFR 50 Appendix
A. General Design Criteria four (GDC 4) which required that structures,
systems and components important to safety be designed to accommodate
the effects of and to be compatible with the environmental conditions
associated with normal operation, testing and postulated accident
conditions.
Based on the vender information and the licensee's test,
the snubbers can not be assured of performing their design restraint
function after exposure to temperatures above 285 F which was
ENCLOSURE 2
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f
32
anticipated in the MSLB event.
It was unclear if the design or
. licensing basis required the snubbers to be operable after a design pipe
4
rupture event (MSLB) or if all )otential dynamic loads following a MSLB
shutdown had been evaluated.
T11s issue is identified as Unresolved
i
Item URI 50-369.370/96-11-03, Environmental Qualification of Safety
Related Piping Grinell Snubbers in Containment. The item remains open
pending further NRC review to determine if the snubbers meet or are
required to meet the GDC 4 requirements.
c.
Conclusion
Review of procurement related items in the licensee's problem
I
identification program did not identify problems with the commercial
i
grade procurement process. An issue was identified related to the
purchase of Grinell hydraulic pipe supports (snubbers) which did not
meet the purchase specifications for environmental qualification. This
item was identified as an unresolved item pending further NRC review.
E7
Quality Assurance in Engineering Activities
l
E7.1 Ouality Assurance in Enaineerina Activities - Procurement Enaineerina
a.
Scope
,
I
The inspector reviewed the licensee's self-assessment activities
associated with procurement engineering 3rocesses.
Applicable
'
regulatory guidance was provided by 10 C R 50 Appendix B.
b.
Observations and Findinas
,
t
A self assessment of McGuire CGD activities was performed in May 1996.
The scope was adequate to assess station performance in this area.
i
Findings were appropriately resolved. Additional self assessment
activity included a benchmarking survey of procurement engineering
activities at five other nuclear facilities to identify opportunities to
enhance Duke procurement engineering practices. An assessment of in-
service CGD part failures was completed in July 1996. This assessment
was to provide information for modification of receipt inspection
requirements or change of vendor suppliers to identify and address
equipment with repeated in-service failures.
.
c
Conclusion
The licensee performed appropriate self assessment to monitor
performance of the commercial grade procurement process.
.
ENCLOSURE 2
)
<
T
4_-.
_
. . . . . _ .
_ . ,
.
__,,m.--.
- _ - - . , _ . . _ _ _ _ _ _
. . . - . . . - .
1
33
E8
Miscellaneous Engineering Issues (92902 and TI 2515/109)
E8.1
(CLOSED) VIO 50-369. 370/95-06-01. Analysis and Calculation Errors.
This violation identified errors in MOV thrust calculations and analyses
of static diagnostic test data.
The inspectors verified the licensee's
overall records of identification and correction of these errors in
Problem Investigation Processes (PIPS) 1-M95-0622. -0636 and -0619.
These PIPS referenced other documents for specific corrective actions
and the inspectors verified a sample of these, including: correction of
the thrust calculation for MOV 1CA0050 in Calculation MCC-1205.19-00-
0003. Rev. 4; and evaluation of structural limitations for Valve ICA0038
in Calculation MCC-1205.19-00-0023. Rev.1.
The inspectors also verified
that the licensee had investigated the extent of condition and a
provided appropriate corrections for other MOVs. As an example, the
inspectors verified correction for MOV 2CA0050 in Calculation MCC-
1205.19-00-0003. Rev. 4.
E8.2 (CLOSED) VIO 50-369. 370/95-06-02. Errors in Entries of Test Data.
This violation involved entries of incorrect test data.
The inspectors
verified the licensee's overall records of identification and correction
of these errors in PIPS 0-M95-0627 and -0641.
These PIPS referenced
other documents for specific corrective actions and the inspectors
verified a sample of these, including: correction of the bench test
torques recorded for MOV 2RN137 (identified WR# 94021156002) to increase
the values by 5 ft-lbs; and correction of procedure IP/0/A/3066/02H
(changes 0 to 8) to assure the correct structural limit is recorded on
the test data sheet.
No additional problems were identified.
This item
is closed.
E8.3 (CLOSED) URI 50-369. 370/95-06-03. Inaporopriate Deferral of MOV Stem
Lubrication.
This item involved the licensee's inappropriate deferral of stem
lubrications specified for several GL 89-10 MOVs.
The stem lubrications
(specified as preventive maintenance) had been deferred from a refueling
outage without recognizing that they could only be com)leted while the
plant was shut down.
This resulted in the intended lu]rication
frequency being exceeded.
The licensee identified this improper
deferral and, recognizing that missed stem lubrications could degrade
MOV performance, they documented the improper deferrals for resolution
through PIP 1-M95-0576.
When NRC inspectors were informed of this
issue, they identified it as an unresolved item for further evaluation
of the safety significance and generic implications.
The inspectors assessed this item during the current inspection through
a review of evaluations and corrective actions documented in com)leted
PIP 1-M95-0576. They agreed with licensee determinations that t7e
missed lubrications did not result in inoperability and that the
ENCLOSURE 2
_. -
.--
.- .
. - - - . -
- - - - . - . - - . - -
-
-
l
i
34
l
inappropriate lubrication deferrals resulted from inadequate
l
procedures / instructions.
The inspectors reviewed work orders (e. g.,
j
work orders 94030172 and 940301870) which documented the licensee's
i
subsequent completion of the improperly deferred lubrications.
l
Additionally, they reviewed the licensee's database listing of deferred
GL 89-10 preventive maintenance as evidence that the licensee's
l
procedures / instructions were now properly controlling deferrals. This
'
licensee identified and corrected violation is being treated as a Non-
Cited Violation, identified as NCV 50-369, 370/96-11-04: Inapproariate
' Deferral of MOV Stem Lubrication. This violation will not be su) ject to
enforcement action because the licensee's efforts in identitying and
l
correcting the violation meet the criteria specified in Section VII.B of
the Enforcement Policy.
E8.4 (CLOSED) URI 50-369. 370/95-06-04 Adeauacy of Actions to Address
Pressure Lockina and Thermal Bindino.
Adequacy of the licensee's actions to address pressure locking and
,
thermal binding will be addressed during the closure of GL 95-07. The
licensee's response to GL 95-07 is discussed in section E1.lb.12.
i
E8.5 (OPEN) URI 50-369.370/96-10-01:
Failure to Ensure Installation of
Correct Heaters in FWST Enclosure
The licensee has installed the correct heaters in the enclosure and
'
added high temperature alarms to alert control room operators of a
failed thermostat that could result in excessive enclosure temperatures.
'
The inspectors are continuing to evaluate the effects of excessive
.
temperatures on the level and temperature transmitters. This item
remains open.
r
IV. Plant SuDDort
t
'
R1
Radiological Protection and Chemistry Controls
R1.1 Transoortation of Radioactive Materials
a.
Insoection Scooe (86750. TI.2515/133)
.
-
The inspectors evaluated the licensee's transportation of radioactive
materials programs for im)lementing the revised Department of
Transportation (DOT) and luclear Regulatory Commission (NRC)
transportation regulations for shipment of radioactive materials as
required by 10 Code of Federal Regulatium (CFR) 71.5 and 49 CFR Parts
100 through 177.
ENCLOSURE 2
.
.
..
_
_
.
1
35
b.
Observations and Findinas
The inspectors reviewed procedures and determined that they adequately
i
addressed the following: assuring that the receiver has a license to
receive the material being shipped; assigning the form, quantity type.
and proper shipping name of the material to be shipped: classifying
waste destined for burial; selecting the type of package required
assuring that the radiation and contamination limits are met; and
preparing shipping papers.
i
Licensee's records for the three shipments of radioactive material
'
performed since the last inspection of this area were reviewed and the
inspe' tors determined the shipping papers contained the required
inf;rmation. The inspectors also determined the licensee had maintained
re.ords of shipments of licensed material for a period of three years
af'er shipment as required by 10 CFR 71.91(a). In addition, the licensee
possessed a current certificate of approval (NRC Form 311) for their
" Quality Assurance Program Description for Radioactive Material Shipping
Packages Licensed Under 10 CFR 71"
i
c.
Conclusions
Based on the above reviews, the inspectors determined that the licensee
had effectively implemented a program for shipping radioactive materials
required by NRC and DOT regulations.
R1.2 Occuoational Radiation Exposure Control Proaram
a.
Insoection Scoce (837501
The inspectors reviewed implementation of selected elements of the
licensee's radiation protection program. The review included
observation of radiological protection activities including personnel
monitoring, radiological postings high radiation area controls, and
verification of posted radiation dose rates and contamination controls
within the Radiologically Controlled Area (RCA). The inspectors also
reviewed licensee records of personnel radiation exposure and discussed
ALARA program details, implementation and goals.
b.
Observations and Findinas
.
The inspectors toured Auxiliary Building facilities, Units 1 and 2
Turbine Buildings, and selected radioactive waste storage areas. At the
time of the inspection, radiological housekeeping was observed to be
good.
Records reviewed determined the licensee was tracking and
trending personnel contamination events (PCEs). The licensee had
tracked approximately 196 PCEs for 1996 which included skin and clothing
contaminations.
Radiologically controlled areas observed we e
appropriately posted and radioactive material observed was appropriately
stored and labeled.
ENCLOSURE 2
,
36
The inspectors reviewed Operational and Administrative controls for
entering the RCA and performing work. These controls included the use of
RWPs to be reviewed and understood by workers prior to entering the RCA.
The inspectors reviewed selected RWPs for adequacy of the radiation
3rotection requirements based on work scope, location, and conditions.
or the RWPs reviewed the inspector noted that appropriate protective
~
clothing, and dosimetry were required.
During tours of the plant the
inspectors observed the adherence of plant workers to the RWP
requirements. The inspectors also performed independent radiation and
contamination surveys of selected areas in the Auxiliary Building and
confirmed RWP information.
The inspectors reviewed Total Effective Dose Equivalent Exposures (TEDE)
for 1996 and 1997. All exposures were well below regulatory limits.
No
workers had received internal exposures at investigative limits in 1996
or 1997 at the time of the inspection.
The inspectors discussed ALARA goals and annual exposures with licensee
management and determined the organizational structure and
responsibilities for the ALARA staff were clearly defined in
organizational charts. Areas reviewed included source term reduction.
ALARA accomplishments and future ALARA plans.
A discussion with
licensee represeritatives and a review of pertinent records determined
the licensee had established an annual site exposure goal for 1996 or
approximately 241.5 person-rem.
Site exposure actually accrued in 1936
was approximately 237.1 person-rem. The site's actual 1996 exposure was
based on operational exposure,
completion of a Unit 1 refueling outages
that began in 1995, a Unit 2 refueling outage, and the addition of two
forced outages. The site's three year average through 1996 was
approximately 129 person-rem per Unit.
Some ALARA initiatives reviewed
included the permanent installation of 50 remote raciation monitors in
the Auxiliary Building, reactor coolant letdown filtration downsizing,
utilization of mockups, extended controlled bursts during shutdowns, and
post outage RHR system flushes.
The inspectors also observed facility preparations for the upcoming 1997
I
Units 1 and 2 Steam Generator (S/G) replacement outages and discussed
3 reparations with the Radiation Protection (RP) S/G management and As
_ow As Reasonably Achievable (ALARA) coordinator.
The inspectors toured
the recently constructed Retired Steam Generator Storage Facility and
reviewed the radiological aspects of facility design with RP management.
In the area of Radiation Protection, the licensee planned to use
basically the same management and work crews that Duke had used to
perform the recent 1996 Catawba Nuclear Station Unit 1 S/G replacement
outage.
The licensee had incorporated RP/ALARA lessons learned into a
management database to be used during the preplanning and execution of
work during the 1997 S/G replacement outages.
Unit 1 "B" S/G. scheduled
for replacement in March of 1997, had a primary to secondary leak which
the licensee was tracking. The licensee was tracking and trending the
leak rate following EPRI guideline methodolcgies.
ENCLOSURE 2
- .
--
.
i
37
'
c.
Conclusions
Radiological facility conditions and housekeeping in radioactive waste
storage areas were observed to be good, material was labeled
appropriately, and areas were properly posted. In addition. RP/ALARA
i
preplanning for the upcoming S/G replacement outages was progressing
i
satisfactorily.
Radiation worker inter nal and external doses were being
maintained well below regulatory limits and the licensee was continuing
to maintain exposures ALARA.
l
R1.3 Water Chemistry Controls
a.
Insoection Scooe (84750)
The inspectors reviewed implementation of selected elements of the
[
licensee's water chemistry control program for monitoring primary and
secondary water quality. The review included examination of program
guidance and implementing procedures and analytical results for selected
!
chemistry parameters.
b.
Observations and Findinas
The inspectors reviewed Technical Specifications (TSs). which described
the operational- and surveillance requirements for reactor coolant
activity and chemistry, and Final Safety Analysis Report (FSAR) Section
,
10.4.7.2.1.
The section indicated that guidelines for maintaining
reactor coolant and feedwater quality were derived from vendor
recommendations and the current revisions of the Electric Power Research
Institute (EPRI) Pressurized Water Reactor (PWR) Primary and Secondary
Water Chemistry Guidelines. The FSAR also indicated that detailed
operating specifications for the chemistry of those systems were
addressed in the Station Chemistry Section.
The inspector reviewed selected analytical results recorded for Units 1.
!
and 2 reactor coolant and secondary samples taken between July 1996, and
!
December 1996.
The selected parameters reviewed for primary chemistry
included dissolved oxygen, chloride, fluoride, and sulfate. The
selected parameters reviewed for secondary chemistry included hydrazine,
iron, and cop)er. Those primary parameters reviewed were maintained
well within t1e relevant TS limits and within the EPRI guidelines for
power operations and cold shutdown modes.
.
Those secondary parameters reviewed were maintained according to station
)rocedures with the exception of hydrazine.
During a review of
lydrazine levels for Unit 1 feedwater, the inspectors noted the licensee
had lowered the hydrazine levels in the final feedwater system. On
March 25,1996. the licensee lowered the hydrazine levels for a period
of several month.s to evaluate if the hydrazine levels currently
specified in the Station Chemistry Manual were in excess of what was
needed to provide optimum corrosion protection juring power operations.
ENCLOSURE 2
38
This temporary change in chemistry parameters was addressed as a trial
in a memo from chemistry to engineering describing the details of the
trial.
In addition, the licensee was monitoring the blowdown
demineralizers to validate any benefits in' extending demineralizer resin
capacity based on the expected lower ammonia concentrations resulting
from the lowec hydrazine levels. The inspectors determined the licensee
conducted this test or experiment, referred to by the licensee as a-
trial, without a written safety evaluation that provides the bases for
the determination that the test or experiment, not described in the
FSAR. did not involve an unreviewed safety question. Accordingly, this
. failure to perform a written safety evaluation is a Violation of 10 CFR 50.59 (b) (1). This issue is identified as VIO 50-369/96-11-05: Failure
to Perform a 10 CFR 50.59 Review Prior to Performing a Test or
Experiment Not Described in the FSAR.
The inspectors reviewed the Problem Investigation Report (PIP)
Number 2-96-3238 dated November 6. 1996. This PIP involved start-up
activities associated with the addition of hydrazine with the Chemical &
Volume Control System demineralizers bypassed. The preliminary results
of the investigation indicated that the demineralizers were placed back
in service before the hydrazine concentration had reached the planned
level resulting in the solubilization and release of Cobalt-58 removed
during the shutdown cleanup. The licensee estimated that between 300 and
400 Curies were released to the primary water system from the
demineralizer beds. These estimates were based on remote teledosimetry
measurements.
This issue is a weakness in the chemistry control
program. in that, the increase in primary plant source term
radioactivity resulting from the solubilization of cobalt 58 will result
in an increase in containment radiation levels. This issue will be
tracked to verify licensee corrective actions / root cause analysis to
prevent reoccurrence and is identified as Inspector Followup Item IFI
50-369/96-11-06 Followup on Licensee Closeout Actions on PIP Number 2-
M96-3238.
c.
Conclusions
Based on the above reviews, it was concluded that the licensee's water
chemistry control program for monitoring primary and secondary water
quality had been imp emented, for those parameters reviewed. in
accordance with the TS requirements, the Station Chemistry Manual, and
the EPRI guidelines for PWR water chemistry with the exception of
secondary water hydrazine addition parameter changes in Unit 1 as
described above. A Violation was identified for the failure to conduct
a 10 CFR 50.59 written safety evaluation prior to performing a chemistry
test or experiment not described in the FSAR. An Inspector Followup
Item was identified to track the closeout actions on the PIP associated
with the solubilization of the Cobalt-58.
ENCLOSURE 2
39
R2
Status of Radiation Protection (RP) Facilities and Equipment
R2.1 Process and Effluent Radiation Monitors
a.
Insoection Scooe (84750)
The inspectors reviewed selected licensee procedures and records for
required surveillances on process and effluent radiation monitors and
for radiation monitor availability.
b.
Observations and Findinas
The inspectors toured the facility to observe the abysical operation of
selected process and radiation monitors in use. Tae inspectors reviewed
selected radiation and process monitor surveillance procedures and
records for performance of channel checks. source checks. channel
calibrations, and channel operational tests.
Performance of those surveillances was required by the TSs and/or 1he
Offsite Dose Calculation Manual (00CM) to demonstrate that the
instrumentation was operable. Those records reviewed indicated that the
surveillances were current and had been performed in accordance with the
a)plicable procedures. The most recent system status report available.
w11ch covered the aeriod January through November 1996. _ indicated that
the overall availaaility for the Radiation Monitoring System remained at
greater than 95 percent operability. The inspectors reviewed and
discussed operability trending records for both safety related and non
safety related monitors with the radiation monitor system engineer and
engineering management.
c.
Conclusions
Based on the above reviews, it was concluded that the licensee had
effectively implemented procedures to track the availability of
radiation monitors and to demonstrate operability of process and
effluent radiation monitors by performance of surveillances at the
frequencies specified in the TSs and the ODCM.
Discussions with
cognizant licensee personnel and a review of performance records
determined the licensee had maintained an overall high level of
operability for radiation monitors in 1996 and was effectively tracking
-
-
monitor performance.
R2.2 Environmental Monitorina Proaram
a.
Insoection Scone (84750)
The inspectors reviewed selected licensee procedures and records for
required surveillances on environmental monitors and monitor
availability.
ENCLOSURE 2
.
..
.-
. . - - - . _ .
.
._._. .
-. .
._
.-
. .
.-
.
-
,
i.
40
!
,
!
b.
Observations and Findinos
!
The inspectors reviewed an ongoing licensee project. MG-95-0449.
l
1
Environmental Sampling Deviation Reduction Plan.
The plan.. discussed in
1
a previous NRC inspection report (IR 96-06). initiated actions to reduce
i
equipment malfunctions which included: surge protection installation:
heat tracing lines for freeze protection: movement water proofing, and
grounding of electrical outlets; air sampler housing physical
modifications to increase air flow: and the addition of two backup
.
i
portable water backup samplers and eight additivaal air samplers. Based
]
on corrective actions, the licensee had reduced the number of equipment
malfunctions in 1996 to 15 as of July 1996 and the inspectors determined
'
only one malfunction had occurred during the last six months of 1996.
l
)
The inspectors selected two environmental air sample locations and one
!
surface water sampling location. The inspectors found the equipment
!
,
o)erable and the material condition acceptable.
Environmental
.
tiermoluminescent dosimeters (TLD's) at sam)le locations 190. 152 and
153 were visited and the inspectors noted tlat they were in place:
,
however, their position (height) above the ground varied from about 18
inches to about seven feet. When the inspectors pointed this out, the
!
licensee noted that this could result in inconsistent background dose
contributions and agreed to evaluate placement of environmental TLDs at
a consistent height.
I
i
c.
Conclusions
I
Based on a review of this area the inspectors determined the licensee
had continued to maintain effective capabilities to perform
environmental samples.
b
R4
Staff Knowledge and Performance in Radiation Protection & Chemistry
a.
Insoection Scone (83750 and 84750)
1^
Training was reviewed to determine whether radiation protection
i
technicians and chemistry technicians were receiving appropriate
training to accomplish their work assignments.
'
l-
2
b.
Observations and Findinas
'
.
1
The inspectors reviewed training agendas and schedules for radiation
i
protection and chemistry technicians for 1996 and also discussed
tentative training plans for 1997 with cognizant training personnel.
1
i
i
f
I
ENCLOSURE 2
.
V
- - .
-
.-
.
-
.
,
-
.
.
- - - . . .
. __ -
- . . - .
- - . . -
-
- _
--
!
.
i
41
c.
Conclusions
Based on a review of training activities reviewed, the inspectors
-
r
determined radiation protection technicians and chemistry technicians
!
were receiving an appropriate level of refresher training to support
ongoing work activities.
l
R7
Quality Assurance.in Radiological Protection and Chemistry
a.
Insoection Scooe (83750 and 84750)
Licensee activities and self assessment programs were reviewed to
I
determine the adequacy of identification and corrective action programs
!
for deficiencies in the areas of RP and Chemistry.
b.
Observations and Findinas
I
i
Reviews by the inspectors determined that Quality Assurance audits and
i
Self Assessment efforts in the area of RP and Chemistry were
accomplished by reviewing RP procedures, observing work, reviewing
l
industry documentation, and performing plant walkdowns to include
surveillance of work areas by supervisors and technicians during normal
!
work coverage.
Documentation of problems by licensee representatives
i
was included in Quality Assurance Audits and Self Assessment Reports.
Corrective actions were included in the licensee's Problem Investigative
Process and were being completed in a timely mannar.
The inspectors reviewed the initial calibration and Quality Assurance of
the Wholebody Counting System. The third Quarter 1996 Interlaboratory
'
Cross Checks and the Control Chart Calculations were reviewed and found
acceptable.
!
c.
Conclusions
l
The inspectors determined the licensee was effectively conducting formal
]
RP and Chemistry audits as required by Technical Specifications and
j
completing corrective actions in a timely manner.
l
F5
Fire Protection Staff Training and Qualification (64704)
l
1
.
.
Annual Fire Drill
l
a.
Insoection Scooe
~
On December 11, 1996, the inspector witnessed plant personnel and off-
j
site emergency personnel response to an off-hour annual fire drill.
!
i
ENCLOSURE 2
.
e
+
..n.
,~
,
,r
. . . .
. . _
_ . ,
...,_,..,_.m.
- ,
-r-,
. . - . -
- .
-
42
b.
Observations and Findinas
The drill scenario involved a fuel delivery truck accident which
resulted in a significant postulated fire around the Unit 1 FWST area.
Offsite agency response to the drill was provided by the Gilead and
Cornelius Volunteer Fire De)artments, which. included 23 fire fighters.
Five members of the onsite icGuire team also responded, and with the
incident commander, acted as lead for the offsite departments
All of
the stated annual fire drill objectives were satisfactorily
accomplished. The inspector noted that the drill personnel
realistically participated in the drill scenario.
Communication between
the fire location and the control room appeared to be good, with the
control room operators monitoring the postulated threat to plant status
through the incident commander.
The inspector also reviewed the post drill critique and concluded that
the players and controllers provided good feedback on potential areas
for improvement.
Numerous issues were identified during the critique
l
including:
-
Several of the fire fighters that responded to support the drill
were less than 18 years of age. These persons were escorted into
the protected area: however, did not enter any radiation
controlled areas or zones while onsite. The licensee recognized
this problem and at the end of the inspection period were taking
actions with the local fire departments to prevent recurrence.
In previous drills, offsite agency responders had entered the
-
protected area via the North gate: however, during this drill, the
control room requested that security admit them at the South gate.
This request was not conveyed to the responders via the 911
operator, therefore all engines initially responded to the north
gate, delaying their entry.
c.
Conclusions
The inspectors concluded that the drill was conducted in a realistic and
professional manner.
The licensee's critique of the drill was candid
and identified several issues, which, once resolved, will improve the
plant and offsite agency , esponse to a fire emergency.
Operators
cognizance of the postulated fire threat during the drill was good.
ENCLOSURE 2
43
V. Management Meetinas
X1
Exit Meeting Summary
The inspectors ) resented the inspection results to members of licensee
management at t1e conclusion of the inspection on January 15, 1997.
The
licensee acknowledged the findings presented.
The inspectors asked the licensee whether any materials examined during the
inspection should be considered proprietary.
No proprietary information was
identi fied.
PARTIAL LIST OF PERSONS CONTACTED
Licensee
Barron, B,, Vice President, McGuire Nuclear Station
Boyle, J. , Civil / Electrical Systems Engineering
Byrum, W., Manager, Radiation Protection
Cline. T.. Senior Technical Specialist, General Office Support
Cross, R., Regulatory Compliance
Davison, Valve Supervisor
Dolan, B.,
Manager, Safety Assurance
Geddie. E., Manager, McGuire Nuclear Station
Herran, P. , Manager. Engineering
Jones, R., Superintendent, Operations
Karriker,
S., Valve Engineer (Site GL 89-10 Program Lead)
Lamb,
J,. Valve Engineer
Michael, R.
Chemistry Manager
Nazar, M.. Superintendent, Maintenance
Painter. D., Valve Engineer
Sample, M.. Manager, Steam Generator Maintenance Group
Setzer, F. , Valve Engineer
Snyder, J. , Manager, Regulatory Compliance
Thomas, K,, Superintendent, Work Control
Travis, B. , Manager, Mechanical / Nuclear Systems Engineering
Tuckman, M., Senior Vice President Nuclear Duke Power Company
NRC
.
S. Shaeffer. Senior Resident Inspector, McGuire
M. Sykes, Resident Inspector, McGuire
G. Harris. Resident Inspector, McGuire
ENCLOSURE 2
44
INSPECTION PROCEDURES USED
1
,
IP 93702:
Prompt Onsite Response Event
IP 71707:
Conduct of Operations
i
IP 71750:
Plant Support
IP 62703:
Maintenance Observations
IP 61726:
Surveillance Observations
IP 40500:
Self Assessment
.
IP 37551:
Onsite Engineering
IP 37550:
Engineering Staff Knowledge and Performance
IP 50001:
Steam Generator Replacement Project Inspection
IP 83750:
Occupational Radiation Exposure
IP 84750:
Radioactive Waste Treatment. AND Effluent AND Environmental
Monitoring
IP 86750:
Solid Radioactive Waste Management AND Transportation Of
Radioactive Materials
TI 2515/133:
Implementation of Revised 49 CFR Parts 100-177 AND 10 CFR Part 71
IP 64704:
ITEMS OPENED, CLOSED, AND DISCUSSED
OPENED
TELE
IFI 50-369, 370/96-11-01
Actions to Address MOV Weaknesses
(paragraph El.1)
VIO 50-369/96-11-02
Failure to Im)lement Temporary
Modification >rocess (paragraph E2.1)
!
URI 50-369, 370/96-11-03
Environmental Qualification of Safety
Related Piping Grinell Snubbers in
Containment (paragraph E4.3)
NCV 50-369, 370/96-11-04
Inappro)riate Deferral of MOV
Stem Luarication (paragraph E8.3)
VIO 50-369/96-11-05
Failure to Perform a 10 CFR 50.59 Review
Prior to Performing a Test or Experiment
not Described in the FSAR (paragraph R1.3)
-
IFI 50-369/96-11-06
Followu) on Licensee Closeout Actions on
PIP Num)er 2-M96-3238 (paragraph R1.3)
ENCLOSURE 2
45
CLOSED
TITLF
LER 50-369/96-02
Inadvertent Manual Initiation of a Unit 1
Feedwater Isolation due to an
Inappropriate Action (paragraph 08.1)
VIO 50-369.370/96-08-02
Inadequate Containment Annulus
Surveillance Procedure (paragraph M8.1)
NCV 50-369. 370/96 11-04
Inapproariate Deferral of MOV
Stem Luarication (paragraph E8.3)
VIO 50-369. 370/95-06-01
Analysis and Calculation Errors
(paragraph E8.1)
VIO 50-369. 370/95-06-02
Errors in Entries of Test Data (paragraph
E8.2)
URI 50-369. 370/95-06-03
Inappropriate Deferral of MOV Stem
Lubrication (paragraph E8.3)
URI 50-369. 370/95-06-04
Adequacy of Actions to Address Pressure
Locking and Thermal Binding (paragraph
E8.4)
DISCUSSED
TTTLE
URI 50-369.370/96-10-01
Failure to Ensure Installation of Correct
Heaters in FWST Enclosure (paragraph E8.5)
LIST OF ACRONYMS USED
-
ALARA -
As Low As Reasonably Achievable
BB
-
Blowdown Recycle
CA
-
Auxiliary Feedwater System
-
Main Feedwater
-
CFR
-
Code of Federal Regulations
DEV
-
Deviation
-
Department of Transportation
-
Electric Power Research Institute
FSAR -
Final Safety Analysis Report
rWDS -
Field Welds Data Sheets
FWST -
Feedwater Storage Tank
GL
-
Generic Letter
IFI
-
Ins]ector Followup Item
i.MFL
-
Idalo National Engineering Laboratory
ENCLOSURE 2
.
-
46
LER
-
Licensee Event Report
-
Motor-0perated Valve
Motor Power Monitor
MPM
-
MSLB -
Main Steam Line Break
-
Non Cited Violation
-
Nondestructive Examination Radiography
NRC
-
Nuclear Regulatory Commission
-
NRC Office of Nuclear Reactor Regulation
ODCM -
Offsite Dose Calculation Manual
-
Personnel Contamination Event
-
Problem Investigation Process
-
Preventative Maintenance
-
PORV -
Power Operated Relief Valve
POR
-
Procedures Qualification Record
1
-
GA
-
Quality Assurance
-
Radiologically Controlled Area
-
-
Radiation Protection
-
Resistance Thermal Detector
RWST -
Refueling Water Storage Tank
-
SGRP -
Steam Generator Replacement Program
-
Spent Fuel Pool
TI
-
Temporary Instruction
TIG
-
Tungsten Inert Gas
TS
-
Technical Specification
Unresolved Item
-
VOTES -
Valve Operation Test and Evaluation System
-
Violation
-
Work Order
WPCS -
Work Process Control Sheets
ENCLOSURE 2