ML20129F556
| ML20129F556 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 10/24/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20129F533 | List: |
| References | |
| 50-382-96-20, NUDOCS 9610290157 | |
| Download: ML20129F556 (20) | |
See also: IR 05000382/1996020
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ENCLOSURE
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket No.:
50-382
License No.:
Report No.:
50-382/96-20
Licensee:
Entergy Operations, Inc.
Facility:
Waterford Steam Electric Station, Unit 3
Location:
Hwy.18
Killona, Louisiana
Dates:
August 27-29 with in-office review until October 3,1996
Inspectors:
L. E. Ellershaw, Reactor Inspector
J. Kudrick, Senior Reactor Systems Engineer
G. E. Werner, Project Engineer
Approved By:
Dr. Dale A. Powers, Chief, Maintenance Branch
Division of Reactor Safety
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ATTACHMENTS:
Attachment 1:
Partial List of Persons Contacted
List of Inspection Procedures Used
List of items Opened, Closed, and Discussed
Attachment 2:
List of Documents Reviewed
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9610290157 961024
ADOCK 05000382
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' EXECUTIVE SUMMARY
Waterford Steam Electric Station, Unit 3
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NRC Inspection Report 50-382/96-20
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This inspection was performed using the guidance of NRC Inspection Procedures 73756,
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" Inservice Testing Of Pumps And Valves," dated July 27,1995; 92902, " Followup -
Maintenance " dated March 14,1994; and 92903," Followup - Engineering," dated March .
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14,1994, to determine whether the inservice testing program had been properly
established and implemented in accordance with Technical Specifications, the applicable
ASME Code, correspondence between NRC and the licensee concerning requests for relief
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and requirements imposed by NRC/ industry initiatives.
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Maintenance
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Pump ;.est requirements had been established in accordance with the 1980 Edition
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of the ASME Code through the Winter 1981 Addenda (applicable to Waterford 3),
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and were being satisfactorily implemented (Section M1.1).
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An inspection followup item was identified to review and evaluate data used to
support the licensee's position that safety-related pumps had not degraded to the
point where they would not meet system design requirements, and that there were
no operability concerns (Section M1.1).
The failure to shut Containment Spray Valve CS-118A following completion of Train
A containment spray pump testing, as required by Step 7.1.23 in Procedure
OP-903-035, was an apparent violation of Technical Specification 6.8.1.a (Section
M 8.1 ) .
The disposition (over a five month period) of a condition report, which led to the
identification of a 10 CFR 50.72(b)(2)(iii) required notification to NRC, was untimely
(Section M8.1).
Enaineerina
The f ailure to perform the required written safety evaluation to provide the bases for
determining that the difference between the facility design configuration and the
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Final Safety Analysis Report did not involve an unreviewed safety question was an
apparent violation of 10 CFR 50.59 (Section E8.2).
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Reoort Details
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Summarv of Plant Status
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Waterford Steam Electric Station, Unit 3, was at full power operation during this
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inspection.
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ll Maintenance
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Conduct of Maintenance
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M 1.1 ASME Code Section XI inservice Testina
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a. Insoection Scope (73756)
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The inspectors reviewed the most recent ASME Code Section XI inservice test data
regarding operability verification for selected Train B pumps given in the following
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procedures:
" Emergency Diesel Generator Fuel Oil Transfer Pump," Surveillance
Procedure OP-903-117, Revision 1;
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" Component Cooling Water Pump," Surveillance Procedure OP-903-050,
Revision 11;
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" Boric Acid Pump," Surveillance Procedure OP-903-004, Revision 9;
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" Auxiliary Component Cooling Water Pump," Surveillance
Procedure OP-903-050, Revision 11;
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" Charging Pump," Surveillance Procedure OP-903-003, Revision 8;
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" Containment Spray Pump," Surveillance Procedure OP-903-035, Revision 8;
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" Safety injection Pump," Surveillance Procedure OP-903-030, Revision 10;
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" Chilled Water Pump", Surveillance Procedure OP-903-063, Revision 9.
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b.
Observations and Findinas
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The inspectors determined from their review of the licensee's records that pump
test parameters and acceptance criteria had been established in accordance with
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ASME Code requirements, and that the performance of all pumps met the
acceptance criteria.
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During their review of the surveillance test package for Component Cooling Water
Pump B, the inspectors identified four change notices applicable to the surveillance
procedure. One change notice reduced the test flow rate from 6000 to 5600 gpm
(Change Notice 1 dated September 18,1995), and another change notice reduced
the test flow rate from 5600 to 4800 gpm (Change Notice 2 dated February 7,
1996). The change from 6000 to 5600 gpm was initiated to provide consistent
test conditions, since the surveillance procedure for Component Cooling Water
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Pump AB specified a flow rate of 6000 gpm, while the surveillance procedures for
pumps A and B specified 5600 gpm. The change from 5600 to 4800 gpm was
recommended by system engineering to preclude the diversion of component
cooling water to the reactor coolant pumps which could cause low flow alarms and
increased reactor coolant pump cooler temperatures.
Section IWP-3110 in Section XI of the ASME Code allows a change to reference
values, as long as an inservice test at the conditions of an existing set of reference
values is performed and the results analyzed. When these are verified as being
satisfactory, then a second test is to be performed at the new reference conditions,
and these results are then to be used to establish the new set of reference values.
The licensee performed these actions and documented the results in accordance
with the requirements of Section IWP-3112 in Section XI of the ASME Code.
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Component Cooling Water Pumps A, B, and AB were each tested using the initial
reference values, followed by a test using the new reference conditions. The tests
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were documented as follows: Component Cooling Water Pump A, Work
Authorization 01144519, performed on February 24,1996; Component Cooling
Water Pump B, Work Authorization 01144513, performed on March 15,1996; and
Component Cooling Water Pump AB, Work Authorization 01143621, performed on
February 8 and 9,1996.
The inspectors also learned that licensee personnel had initiated Condition Report
CR-96-0414 on March 20,1996, to address a concern regarding component
cooling water pump inservice test acceptance criteria. Licensee personnel had
identified that the inservice test acceptance criteria permitted component cooling
water pump performance to degrade below the design basis accident performance
point without requiring action to restore pump performance. The design flow rate
for the component cooling water system was 6554 gpm. The certified pump curve
showed that the pump should develop 151 feet of head at 6554 gpm, or 148 feet
of head at 6800 gpm. This difference translated to 2 percent of head margin or
3.8 percent of flow margin. However, the inservice test acceptance criteria, which
was established in accordance with ASME Code requirements, allowed for (a)
7 percent head degradation and 6 percent pump flow degradation below the
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reference point, respectively, before the alert range limit was reached, and (b)
10 percent head degradation and 10 percent pump flow degradation before the
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action range limit was reached. Consequently, this amount of pump degradation
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would result in the system performing below the design basis flow rate before the
inservice test trend data indicated any problem. The licensee's review of the
condition report that discussed this susceptibility resulted in a number of corrective
actions, some of which were still open at the end of the onsite inspection.
The licensee performed a review of all other safety-related pumps for similar
conditions. Subsequent to the onsite inspection, the inspectors were informed that
the review determined that inservice test acceptance criteria, with the exception of
the component cooling water pumps and Auxiliary Component Cooling Water Pump
A, were set sufficiently high such that pump degradation would be detected before
the design basis accident required flow rate was encroached on.
An associated root-cause analysis report dated May 14,1996, stated that the
component cooling water pumps had been, or were, operating in a condition where
the pumps would not have provided 148 feet of head at 6554 gpm. It also stated
that a previous analysis associated with another condition report (CR 95-0955)
determined that even with flow as low as 6000 gpm, the design basis heat load
could still be removed by the component cooling water system. In addition,
previous special flow balance tests had been performed using Special Test
Procedure 01150154,"CCW System Flow Balance Test." The results from those
special flow balance tests showed that the total flow rate was approximately 6300
gpm, which was below the design basis accident flow rate of 6554 gpm.
Because of the lower-than-expected flow rates, engineering performed a 10 CFR 50.59 evaluation to assess the impact of the lower flow rates on heat transfer
during a design basis accident. The evaluation, approved on October 20,1995,
assumed a 6000 gpm component cooling water system flow rate, which bounded
the design flow rate of 6554 gpm and the actual measured flow rate of 6300 gpm.
New analyses were performed for the design basis accidents using the assumed
lower flow rates. These analyses demonstrated that the component cooling water
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system heat removal performance with the lower total flow rate was greater than
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that assumed in the accident analyses. The evaluation concluded that, even with
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flow rates as low as 6000 gpm, the component cooling water system would still
perform its safety function and all applicable acceptance limits would be met.
During their review of the root-cause analysis report, the inspectors noted that the
issue of pumps degrading below the design basis accident flow rate before the
inservice test program would detect the problem, had been identified at Arkansas
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Nuclear One as early as 1990. It appeared that in February 1994, Waterford
personnel became aware of the issue, and on March 20,1996, the condition report
was issued on this subject. While all corrective action tasks have not been
completed, licensee personnel informed the inspectors that their review of all
safety-related pump test data for the past year showed that the pumps exceeded
the design basis accident flow rates established for their respective systems.
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The inspectors did not review and evaluate the data which was used to support the
licensee's position that the pumps had not degraded to the point where they would
not meet system design requirements, and that there were no operability concerns.
This matter will be reviewed as an inspection followup item during a future
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inspection (50-382/9620-01).
M8
Miscellaneous Maintenance issues (92902)
M8.1 (Closed) License Event Reoort 50-382/96-012: Containment Spray Valve CS-118A
Found Partially Open.
a.
Inspection Scope
The inspectors reviewed Licensee Event Notification 30902 dated August 21,1996,
and Licensee Event Report 50-382/96-012 dated September 20,1996, which
informed the NRC that Containment Spray Valve CS-118A was found partially open
(approximately one and one-half turns) on November 11,1995, subsequent to
testing performed on the Train A containment spray pump. Valve CS-118A serves
as a flow test line isolation valve. The licensee event notification also identified that
during a design basis loss-of-coolant accident while in the recirculation mode,
leakage through this valve back to the refueling water storage pool could have
exceeded the 8 gpm limit established in Updated Final Safety Analysis Report
Table 15.6-19 and, subsequently, could have overexposed the control room
operators.
b.
Observations and Findinas
Following performance of Surveillance Procedure OP-903-094,"ESFAS Subgroup
Relay Testing- Operating," Revision 8, licensee personnel observed that containment
spray riser level had dropped approximately 55 ft. Investigation revealed that Valve
CS-118A, an Anchor Darling,4 in, normally locked closed, manual gate valve, was
not fully closed and was found to be one and one-half handwheel turns open.
Condition Report CR 95-1165,which was initiated on November 11,1995,
concluded that containment spray pump operability was not affected since
backleakage to the refueling water storage pool was quantified by engineering
judgement to be less than the acceptance criterion of 60 gpm.
The system engineering superintendent reviewed Condition Report CR 95-1165 and
identified that the condition report f ailed to address Valve CS-118A backleakage
radiological consequences. Consequently, licensing personnel initiated Condition
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Report CR 96-0287 (March 2,1996) to address the new issue.
The licensee identified that Valve CS-118A had last been operated on
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September 19,1995, during performance of Surveillance Procedure OP-903-035,
" Containment Spray Pump Operability Check," Revision 8, Section 7.1,
" Containment Spray Pump A." During the surveillance, Valve CS-118A was opened
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and then subsequently locked closed; however, discussions with operations
personnel revealed that the valve apparently had not been completely closed. This
was determined during a subsequent check (as a result of the containment spray
riser level drop) on November 11,1995, when personnel used a " valve operator" (a
procedurally allowed leverage bar), and closed the valve an additional one and
one-half handwheel turns. Coridition Report 951165 identified the difficulty in
operating the valve as a cause for the valve not being fully closed.
Based upon the above dates, Valve CS-118A was partially open for'approximately
53 days; however, the plant was in a refueling outage during this period for 44 days
(September 22 to November 5,1995). In accordance with Technical Specification 3.6.2.1, the containment spray system was not required to be
operable in Mode 4 (below 400 psia), and in Modes 5 and 6. The inspectors review
of control room logs showed that the plant reached Mode 4 and 350 psia at 0600
on September 23,1995. Subsequently, during power ascension, the plant reached
Mode 4 and 375 psia at 0600 on October 31,1995. Therefore, the valve was
partially open when it was required to be closed between September 19 and 23,
and between October 31 and November 11,1995.
Technical Specification 6.8.1.a, requires that written procedures are to be
implemented for those activities referenced in Appendix A of Regulatory Guide 1.33, one of which is surveillances. Surveillance Procedure OP-903-035, used to
satisfy that requirement, contained Step 7.1.23, which required that Containment
Spray Valve CS-118A be closed and locked following completion of Train A
Containment Spray Pump testing. The performance and verification of this step
was documented by initials on Step 23 in Attachment 10.1 to the procedure on
September 19,1995. The failure to shut Valve CS-118A as required by
Surveillance Procedure OP-903-035 was an apparent violation of Technical Specification 6.8.1.a (50-382/9620-02).
Valve CS-118A was one of the 13 valves previously identified by the licensee as
not having been included in the inservice test program and, therefore, not tested
(see NRC Inspection Report 50-382/9609). The licensee had placed the valve in the
inservice test program, and on March 2,1996,the valve was tested for leakage.
Test results showed no leakage (i.e.,0.0 gpm).
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However, from March to June 1996, the licensee attempted to bound the potential
leakage problem that existed during September through November 1995, through
design calculations, bench testing of a similar valve, and performing a special test
through partially opened Valve CS-118A. However, none of the licensee's efforts
were successful on quantifying the leakage.
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On August 21,1996, engineering personnel determined they could not quantify
refueling water storage pool backleakage at low flow rates. Based upon this
information, licensing personnel, in accordance with 10 CFR 50.72(b)(2)(iii), made a
4-hour notification on August 21,1996. The inspectors reviewed the reportability
aspects of this issue and found them acceptable.
Subsequent to the onsite inspection, the licensee submitted Licensee Event
Report 50-382/96-012 dated September 20,1996. In the report, the licensee
discussed a special test that was performed on September 7,1996,in another
attempt to quantify the leak rate through Valve CS-118A. While there were no
consequences associated with the actual event, the test determined that the
maximum leak rate during a design basis loss-of-coolant accident would have been
approximately 11.8 gpm, which exceeded the 8 gpm backleakage limit specified in
Table 15.6-19 in the Updated Final Safety Analysis Report.
The backleakeage limit was established to maintain the control room thyroid dose
limits specified in Criterion 19 of Appendix A to 10 CFR Part 50. This number
(11.8 gpm) was arrived at by using actual flow rates from three different valve
settings, and by perforrning valve leakage calculations using expected values during
an accident for elevation head, containment spray pump head, valve and line losses,
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and differential pressure across Valve CS-118A. The test also determined that the
backleakage would not have caused off-site dose limits (10 CFR Part 100) to be
exceeded. The licensee determined that a leak rate greater than 12.5 gpm during
the design basis loss-of-coolant accident would be required to exceed the off-site
dose limits. The licensee performed additional control room dose limit calculations.
In addition to the excessive backleakage through Valve CS-118A, the licensee
considered the number of hours personnel would be in the control room during an
accident, and included other possible leakage paths. The Licensee Event Report
concluded that the dose attributed to these conditions would not exceed the thyroid
dose limit for the control room staff.
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c.
Conclusions
Pump test requirements had been established and were being implemented in
accordance with the 1980 Edition of the ASME Code through the Winter 1981
Addenda (applicable to Waterford 3). An apparent violation of Technical Specification 6.8.1.a occurred when licensee personnel f ailed to shut Valve
CS-118A following a containment spray pump test. The licensee's disposition (over
a five month period) of Condition Report 96-0287 (March 2,1996), which led to
the identification of a required report (August 21,1996) to NRC, was untimely.
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E2
Engineering Support of Facilities and Equipment
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E2.2 Review of Updated Final Safety Analysis Report Commitments
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A recent discovery of a licensee operating their facility in a manner contrary to the
Updated Final Safety Analysis Report description highlighted the need for a special
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focused review that compares plant practices, procedures, and/or parameters to the
Updated Final Safety Analysis Report description. While performing the inspections
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discussed in this report, the inspectors reviewed the applicable sections of the
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Updated Final Safety Analysis Report that related to the areas inspected.
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in addition, the inspectors reviewed the licensee's 10 CFR 50.59 Safety Evaluation
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Report, "LDCR 97-0047, Location of Cabinets C-3A(B) and C-4 Outside the CVAS
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Boundary, Rev.1," dated July 28,1996. This report was initiated on July 22,
1996, because the licensee became aware that their justification, in response to
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Final Safety Analysis Report Question 480.36, for not conducting 10 CFR Part 50,
Appendix J, Type C leak tests on the valves in the instrument lines through
Containment Penetrations 53 and 65, was not correct (Appendix J, Type C leak
tests are pneumatic tests used to measure containment isolation valve leakage
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rates). This resulted in errors between the wording of the Updated Final Safety
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Analysis Report and the current plant configuration regarding the containment
vacuum relief lines.
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The 10 CFR 50.59 report identified the following Updated Final Safety Analysis
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sections and tables requiring changes: Section 1.9.37; Section 7.1.2.7;
Table 3.9-9; Table 6.2-32, and Table 6.2-43. In addition, the following tables in the
Technical Requirements Manual were also identified as requiring changes:
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Tables 3.6-1 and 3.6 2. Licensee personnel stated that the Updated Final Safety
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Analysis Report would be changed to reflect the actual plant configuration, and to
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revise leak rate test requirements.
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Additional details and information pertaining to the potential consequences of these
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discrepancies are located below in Sections E8.1 and E8.2.
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E8
Miscellaneous Engineering issues (92903)
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E8.1
(Closed) Unresolved item 50-382/9609-04: the determination of whether there
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were active safety functions for Excess Flow Check Valves CVR-302A(B),
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" Containment / Annulus Differential Pressure Sample Line Check Valve."
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Containment Penetrations 53 and 65 each contain two containment vacuum relief
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instrument lines. One instrument line senses differential pressure across the
containment vessel and provides a signal to actuate the containment vacuum relief
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system. The other instrurnent line monitors the same differential pressure and
provides an input to the plant computer.
Excess Flow Check Valves CVR-302A(B) were installed in the containment vacuum
relief sensing instrument lines to meet the guidance specified in Regulatory
Guide 1.11. This was committed to in Section 6.2.4.2.2 of the Updated Final
Safety Analysis Report. Regulatory Guide 1.11 states, in part, "In the event of a
rupture downstream of the valve, the valve should close automatically or be capable
of being closed during normal reactor operation and under accident conditions."
In Condition Report 96-0272, dated March 1,1996, the licensee identified
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that Valves CVR-302A(B) had a closed safety function. They were added to the
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inservice test plan as ASME Code Category C valves, but with a cold shutdown test
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frequency justification'. The condition report further stated that these valves were
listed in Table 3.6-2 of the Technical Requirements Manual which identified them as
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containment isolation valves and subject to Technical Specification 3.6.3, and
10 CFR Part 50, Appendix J testing.
The licensee's representatives indicated that the valves had been tested during
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previous refueling outages in the closed direction and were capable of meeting their
closed safety function. However, these penetrations were only tested during the
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10 CFR Part 50, Appendix J, Type A tests. (Appendix J, Type A tests measure the
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containment system overall integrated leakage rate under conditions representing
design basis loss-of-coolant accident containment peak pressure). No individual
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testing of the valves was performed. Licensee personnelinformed the inspectors
that closed-system leak tight integrity tests had been cor. ducted on the closed
systems during refueling outages; however, this required that the valves be open in
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order to verify system integrity.
Subsequently, licensee personnel stated that, following additional reviews, they
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identified that the excess flow check valves did not have a closed safety function
and did not have to be tested in the closed direction. The licensee supported this
position by referencing Final Safety Analysis Report Question and Response 480.36
and Updated Final Safety Analysis Report, Table 6.2-32, which showed that
10 CFR Part 50, Appendix J, Type C containment isolation valve leak testing was
not required for these penetrations. In addition, Table 6.2-43, showed that
Penetrations 53 and 65 were tested during Appendix J, Type A tests and were not
required to have a separate Appendix J, Type C test. However, NRC acceptance of
that position, as stated in the Safety Evaluation Report, "Waterford Steam Electric
'They also were part of the subject of apparent violation 50-382/9609-02, regarding a
f ailure to include valves in the inservice test program that were recuired to be tested in
accordance with Section XI of the ASME Code.
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Station, Unit 3, NUREG-0787" dated July 1981, did not imply that no testing was
acceptable. It simply indicated that an Appendix J, Type C leak rate test was not
necessary.
On July 20,1996, while the plant was in Mode 5, Cold Shutdown, the licensee
performed an inservice stroke test on Valve CVR-302B, and the valve failed to
close. The valve was declared inoperable, and Condition Report CR-96-1103 was
initiated. On July 21,1996, a stroke test was performed on Valve CVii-302A, with
similar results. The f ailure of Valve CVR-302A was incorporated into the existing
condition report.
The f ailed valves were removed from the containment vacuum relief sensing lines.
The replacement valves were tested and found to be acceptable. The licensee's
evaluation concluded that had these valves f ailed during plant operations, there
would not have been a safety concern. This evaluation was based on the fact that
the tubing downstream of these valves was safety related (i.e., fabricated to ASME
Code, Class 3 requirements, and classified as Seismic Category 1), and constituted
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a closed system outside of containment which would be assumed to fulfillits
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intended function (i.e., maintain its integrity) under accident conditions. The
inspectors agreed with the licensee's conclusion.
During the assessment of Condition Report CR 96-1103,the NRC inspector
discussed the issue with licensee personnel, who indicated that the instrument lines
did not terminate in the controlled ventilation area system, and that a potential for
bypass leakage existed. The inspectors, on July 21,1996, questioned licensing
personnel about what appeared to be a discrepancy, in that part of the licensing
basis was predicated on the lines terminating in the controlled ventilation area
system. On July 22,1996, further inspector questioning caused licensee personnel
to conduct a review of their response to Final Safety Analysis Report Question
480.36 regarding justification for not conducting Appendix J, Type C leak tests on
the valves in the instrument lines through Containment Penetrations 53 and 65.
Through this review, the licensee became aware that an incorrect response had
been provided to the NRC. This caused the initiation of Condition Report
CR 96-1123 dated July 23,1996, which documented this discrepancy. On August
26,1996, Licensee Event Report 50-382/96-009,which is discussed below, was
issued to address the issues identified during the licensee's review. Subsequently,
an August 12,1996, predecisional enforcement conference was postponed pending
review of the issues that were raised by the licensee's report.
E8.2 (Closed) LER 50-382/96-009: failure of Containment Vacuum Relief
Valves-402A(B) (excess flow check valves). On July 22,1996, the licensee
became aware that the cabinets for the instrumentation associated with the
containment vacuum relief system (sensing and monitoring lines) were not within
the controlled ventilation area system, and that this discrepancy was inconsistent
with the licensing basis. Previously, on July 5,1996, the licensee determined that
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the required Technical Specification testing had not been performed on Valves
CVR-401 A(B) and CVR-402A(B) which were located in the containment vacuum
relief monitoring lines.
During initiallicensing of Waterford Steam Electric Station, the licensee responded
to Final Safety Analysis Report Question 480.36 (dealing with testing of certain
containment penetrations) by stating that the containment vacuum relief sensing
lines and the containment vacuum relief monitoring lines each formed a closed
system outside of containment, were seismically qualified, and terminated within an
area exhausted by the controlled ventilation area system (a filtered ventilation
system). Based on this rationale, the NRC staff determined that the design and
isolation provisions of Containment Penetrations 53 and 65 would be acceptable to
meet General Design Criterion 56 of Appendix A to 10 CFR Part 50. This
determination was included in Section 6.2.4 of Safety Evaluation Report,
"Waterford Steam Electric Station, Unit 3, NUREG-0787" dated July 1981.
Further licensee review of their response revealed that, not only did the instrument
lines not terminate in the controlled ventilation area system, but the monitoring lines
did not meet the design criteria for a closed system outside of containment and they
were not seismically qualified.
10 CFR 50.59(a)(1) allows the holder of a license to make changes to the facility as
described in the safety analysis report unless the proposed change involves an
unreviewed safety question. 10 CFR 50.59(b)(1) requires the licensee to maintain
records of changes in the facility, to the extent that these changes constitute
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changes in the facility as described in the safety analysis report. The records must
include a written safety evaluation which provides the bases for the determination
that the change does not involve an unreviewed safety question.
The actual design configuration of the containment vacuum relief system was
different from that described in the licensee's response to Final Safety Analysis
Report Question 480.36, and until July 28,1996, the licensee failed to perform the
required written safety evaluation to provide the bases for a determination that the
deviation from the response to Final Safety Analysis Report Question 480.36 did
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not involve an unreviewed safety question. This failure was an apparent violation of
10 CFR 50.59 (50-382/9602-03).
Criterion 56 of Appendix A to 10 CFR Part 50 requires each line that connects
directly to the containment atmosphere and penetrates primary reactor containment
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shall be provided with two containment isolation valves unless it can be
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demonstrated that the containment isolation provisions for a specific class of lines,
such as instrument lines, are acceptable on some other defined basis. In addition,
the use of a simple check valve as the automatic isolation valve outside
containment is not alloweo.
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Regulatory Guide 1.11, which the licensee committed to in Section 6.2.4.2.2 of the
Updated Final Safety Analysis Report, provided suitable bases for demonstrating
acceptable alternative containment isolation provisions. However, in order to use
the provisions of Regulatory Guide 1.11, the valving provided for each instrument
line penetrating containment must reflect the importance of two safety features: (1)
the function the line performs, and (2) the need to maintain containment leaktight
integrity.
With respect to the sensing lines, the importance of the safety function was
recognized and application of Regulatory Guide 1.11 was justified. The lack of a
safety function for the monitoring line, however, established that use of Regulatory
Guide 1.11 was not an appropriate alternative with respect to the isolation
provisions as described in General Design Criterion 56, and the use of an excess
flow check valve was not an acceptable isolation barrier. The failure to implement
General Design Criterion 56 requirements by not providing the specified
containment isolation barriers in the containment vacuum relief monitoring lines was
a result of the licensee failing to recognize that the actual design configuration was
different from their response to Final Safety Analysis Report Question 480.36.
The inspectors also noted that during the operating license review, the NRC staff
asked for additional justification for not including Containment Penetrations 53 and
65 in an Appendix J, Type C leak testing program as discussed in Question 480.36.
The licensee response reiterated the isolation barriers as described above, but also
indicated that Solenoid Operated Valves CVR-401 A(B), which would close on a
containment isolation signal, would be added to the containment vacuum relief
monitoring lines during the first refueling outage. There was no mention of any
testing requirements associated with those valves since the acceptance of the
penetrations was based on the premise that closed systems existed in conjunction
with the excess flow check valves. The licensee's representatives stated that the
valves would be open for each Appendix J, Type A test so that the integrity of the
closed system would be tested. In addition, licensee personnelinformed the
inspectors during this inspection, that a local leak tight integrity test had been
conducted on all the closed systems during those refueling outages when the
Appendix J, Type A test was not performed.
After identifying that the containment vacuum relief monitoring lines did not
constitute a closed system, the licensee isolated these lines by closing the upstream
Solenoid Isolation Valves CVR 401 A(B). On July 26,1996, Valves CVR-401 A(B)
and CVR-402A(B) were leakrate tested under Work Authorization 01149602 with
the following results, as given in Table 1.
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TABLE 1
VALVE
AS-FOUND
AS-LEFT
LEAKRATE (sccm)
LEAKRATE(sccm)
CVR-401 A
20.0
20.0
CVR 4018
21.6
21.6
CVR-402A
129,000
44500*
CVR-402B
242,500
44500*
- Assigned design leakage - actual measured leakage was less
than this value.
Excess Flow Check Valves CVR-402A(B) failed to close during the initial leakrate
test and were removed. It appeared that these valves failed because of excessive
manufacturing roughness in the valve body bore. The replacement valve body
bores were smoothed prior to installation and were successfully tested as shown
above. Technical Specification 3.6.1.2 limited combined bypass leakage to a rate
of less than or equal to 0.06 L,. The Solenoid Isolation Valves CVR-401 A(B) were
left shut since the combined bypass leakage through these lines could exceed
0.06 L,if both solenoid valves failed to shut. This leakage was determined by
licensee calculation to be equal to 63,069 scem and was incorporated into
Surveillance Procedure STA-001-006," Leak Rate Testing," Revision 2, Attachment
10.9. As of July 26,1996, with Valves CVR-401 A(B) shut, the combined bypass
leakage was 14,924.7 scem.
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Technical Specification Surveillance Requirement 4.6.1.2.d required that bypass
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flow be measured at least once per 24 months by either performing a Type B (not
applicable to containment isolation valves) or Type C test on individual penetrations
that can be tested. The failure of the licensee to test Penetrations 53 and 65 prior
to July 26,1996,in accordance with Technical Specification Surveillance
Requirement 4.6.1.2.d, was another result of the licensee not recognizing the
discrepancy between the actual design configuration of the containment vacuum
relief system and their response to Final Safety Analysis Report Question 480.36.
The licensee determined that both control room doses and offsite doses could have
been exceeded had a design basis loss-of-coolant accident occurred assuming
rupture of the nonessential monitoring line and failure of either Valve CVR-401 A or
CVR-401B with Valves CVR-402A or CVR-4028 failed in the open position
(as-found condition). Licensee calculated control room and offsite doses are shown
in Table 2.
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TABLE 2
LOCATION
THYROID DOSE
WHOLE BODY
SKIN DOSE
(rem)
DOSE (rem)
(rem)
2 Hr Exclusion
405.82(300)'
13.43(25)"
N/A
Area Boundary
30 Day Low
162.85 (300)~
3.323(25)*
N/A
Population
Zone
30 Day Contro!
45.71(30)*
1.36 (5)*
31.51 (30)*
Room
,
- The values within parentheses are regulatory limits from the standard review plan.
1
On August 21,1996, the licensee submitted License Amendment Request W3F1-
96-01441," Discrepancy Regarding the Design and Testing of Instrument Sensing
Lines Penetrating the Primary Containment," to resolve the design and testing
discrepancies associated with both the containment vacuum relief sensing and
containment vacuum relief monitoring lines.
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The inspectors reviewed the Updated Final Safety Analysis Report for other
instrumentation lines penetrating containment and reviewed the construction of
those lines. The inspectors verified that no other containment penetration lines
utilized excess flow check valves. Also, the licensee searched their parts data base
and identified that the excess flow check valves were used exclusively on the
Containment Vacuum Relief System (Penetrations 53 and 65). Updated Final Safety
Analysis Report Section 6.2.4.2.2, " Instrument i.ines," identified only one additional
instrument line (Penetration 54) that penetrated containment.
This instrument line was used for the wide range containment pressure
instrumentation, and was depicted on Drawing B-430, Sheet P-77. The Updated
Final Safety Analysis Report described the containment wide range pressure
instrumentation as consisting of a sealed, liquid-filled system with a bellows,
constructed in accordance with ANSI N271-1976," Containment isolation
Provisions for Fluid Systems." ANSI N271-1976 was used to satisf'y the
requirement of 10 CFR Part 50, Appendix A, Criterion 56.
The inspectors performed a walkdown of the piping and instrumentation outside
containment for Penetration 54 and identified that the transmitter was not enclosed
in protective shielding (protection from missiles and water jets) as required by ANSI
N271-1976. Followup discussion with the licensee identified that the piping and
transmitter wl:'.; located in an area that had been analyzed for jet impingement and
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no jet impingement concerns were identified. In addition, other components located
in the same area were all seismically qualified; therefore, no missiles would be
associated with their failure. Based upon this information, the inspectors
determined that Penetration 54 hardware were adequate to provide the required
safety function.
V. Manaaement Meetinas
X1
Exit Meeting Summary
Subsequent to the conclusion of the inspection on October 3,1996, the inspector
telephonically presented the inspection results to members of licensee management on
,
October 9,1996. The licensee acknowledged the findings presented.
The inspector asked the licensee whether any materials examined during the inspection
should be considered proprietary. No proprietary information was identified.
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7
ATTACHMENT 1
PARTIAL LIST OF PERSONS CONTACTED
Licensee
P. Caropino, Licensing Coordinator
C. Dugger, General Manager, Plant Operations
J. Fisicaro, Director, Nuclear Safety
C. Fugate, Shift Supervisor, Operations
T. Gaudet, Manager, Licensing
P. Gropp, Supervisor, Mechanical Specialties, Design Engineering
J. Holman, Manager, Safety Analysis
J. Houghtaling, Technical Assistant, Design Engineering
J. Howard, Manager, Procurement / Programs Engineering
P. Melancon, inservice Test Engineer
G. Robin, Supervisor, Programs Engineering
L. Rushing, Manager, Mechanical / Civil Design Engineering
M. Selman, Vice President, Operations
P. Stanton, Design Engineer, Mechanical Systems
C. Thomas, Supervisor, Licensing
R. Thweatt, Supervisor, Design Engineering, Mechanical
D. Urciuoli, Licensing Engineer
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D. Vinci, Manager, Systems Engineering
K. Walsh, Lead Senior Engineer, Operations
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A. Wrape Ill, Director, Design Engineering
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NRC
K. Brockman, Acting Director, Division of Reactor Safety
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D. Powers, Chief, Maintenance Branch
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L. Keller, Senior Resident inspector
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T. Pruett, Resident inspector
LIST OF INSPECTION PROCEDURES USED
Inservice Testing of Pumps and Valves
Followup - Maintenance
'
Followup - Engineering
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. - . - - - . . - . . .
_ . . .
- - _ - - . . - - - . - .
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LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
50-382/9620-01
IFl
An inspection followup item was identified to review and
evaluate data used to support the licensee's position that
safety-related pumps had not degraded to the point where thc,'
would not meet system design requirements and that theie
were no operability concerns.
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50-382/9620-02
eel
The failure to shut Containment Spray Valve CS-118A
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following completion of Train A containment spray pump
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testing, as required by Step 7.1.23 in Procedure OP-903-035,
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was an apparent violation of Technical Specification 6.8.1.a.
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50-382/9620-03
eel
An apparent violation regarding the failure to perform the
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required written safety evaluation to_ provide the bases for a
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determination that the difference between the facility design
{
configuration and the Final Safety Analysis Report was not an
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unreviewed safety question. This resulted in a failure to meet
General Design Criterion 56 requirements with respect to
providing the specified conteir. ment isolation barriers in the
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containment vacuum relief system, and a failure to test
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Penetrations 53 and 65 pnor to July 26,1996,in accordance
with Technical Specif cation Surveillance Requirements.
Closed
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50-382/9609-04
This item pertained to determination of active safety functions
for excess flow check valves CVR-302 A(B).
.
50-382/96-009
LER This item dealt with failure of Containment Vacuum Relief
C
Valves-402A(B)(excess flow check valves).
[
50-382/96-012
LER This item dealt with the failure to close Containment Spray
Valve CS-118A following pump testing.
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A
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ATTACHMENT 2
LIST OF DOCUMENTS REVIEWED
Desian Documents
"Waterford Quality Assurance Program Manual," Revision 5
" Technical Specifications," Amendment 112
" Updated Final Safety Analysis Report," Revision 9
"Waterford 111 Pump and Valve inservice Test Plan," Revision 8, Change 1
"Waterford ill Pump and Valve Inservice Test Plan," Revision 7, Change 10
W3-DBD-024," Inservice Testing Basis Document," Revision 0 and
Revision 1
W3-DBD-014, " Safety-Rr.iated Air-Operated Valves," Revision O
W3-DBD-042, " Sampling Sy. item," Revision 0
Procedures
Surveillance Procedure OP-903-035," Containment Spray Pump Operability Check,"
Revision 8
Surveillance Procedure OP-903-117," Emergency Diesel Generator Fuel Oil Transfer
Pump," Revision 1
Surveillance Procedure OP-903-050," Component Cooling Water Pump," Revision 11
Surveillance Procedure OP-903-004," Boric Acid Pump," Revision 9
Surveillance Procedure OP-903-050," Auxiliary Component Cooling Water Pump,"
Revision 11
Surveillance Procedure OP-903-003, " Charging Pump," Revision 8
Surveillance Procedure OP-903-035," Containment Spray Pump," Revision 8
Surveillance Procedure OP-903-030," Safety injection Pump," Revision 10
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Surveillance Procedure OP-903-063," Chilled Water Pump," Revision 9
Surveillance Procedure STA-001-006," Leak Rate Testing," Revision 2
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Administrative Procedure UNT-006-010," Event Notification and Reporting," Revision 14
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Administrative Procedure, UNT-006-011," Condition Report," Revision 4
Administrative Procedure, UNT-006-021," Pump And Valve Inservice Testing," Revision 2
Site Directive, W2.302, "10 CFR 50.59 Safety and Environmental Impact Screening and
Evaluation"
Condition Reports
CR 95-1165
CR 96-0287
CR 96-0272
CR 96-0414
CR 96-1123
Work Authorizations
0114960
01144519
01144513
01143621
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