IR 05000424/1998003
| ML20247K767 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 05/15/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20247K728 | List: |
| References | |
| 50-424-98-03, 50-424-98-3, 50-425-98-03, 50-425-98-3, NUDOCS 9805220201 | |
| Download: ML20247K767 (33) | |
Text
_ _ _ _ _ _ _ _ - _ - _ _ _ _ _
_ _ _ _ _ _ _ _
_
_
U. S. NUCLEAR REGULATORY COMMISSION (NRC)
REGION 11 Docket Nos.
50-424 and 50-425 License Nos.
50-424/98-03. 50-425/98-03 Licensee:
Southern Nuclear Operating Company, Inc.
i Facility:
Vogtle Electric Generating Plant (VEGP) Units 1 and 2 Location:
7821 River Road Waynesboro, GA 30830 Dates:
March 8, 1998 through April 18, 1998 Inspectors:
J. Zeiler, Senior Resident Inspector M. Widmann, Resident Inspector i
K. O'Donohue. Resident Inspector i
T. Fredette. Resident Inspector. Hatch (Sections 02.3, E8.5.
and F8.1)
W. Kleinsorge, Regional Inspector (Section M1.4)
G. Kuzo, Regional Inspector (Sections R1 and R8)
N. Merriweather, Regional Inspector (Sections El, E3.1.
E8.1. E8.2, E8.3, and E8.4)
L. Straton, Regional Inspector (Section S1.1)
Approved by:
P. Skinner, Chief Reactor Projects Branch 2 Division of Reactor Projects e
Enclosure 2 9905220201 980515 PDR ADOCK 05000424 G
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ ~
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _
i EXECUTIVE SUMMARY Vogtle Electric Generating Plant Units 1 and 2 NRC Inspection Report 50-424/98-03, 50-425/98-03 This integrated inspection included aspects of licensee operations, engineering, maintenance, and ]lant support.
The report covers a 6-week period of resident and region Jased inspection.
Doerations I
Management's direct involvement in contingency plan walkdowns and the
.
heightened awareness of the operating staff to the contingency plans during periods of increased plant risk during Unit 2 refueling outage (2R6) was excellent.
The overall conduct of 2R6 was a strength
)
(Sections 01.1 and 01.3).
j The inspectors determined that the decision to trip the unit was
.
justified based on the observed rod position discrepancies between Digital Rod Position Indication and the demand position step counters.
The licensee adequately determined the cause of the rod control malfunction and corrected the equipment-related degradation prior to unit startup (Section 01.2).
The inspectors concluded that refueling activities were performed in a
.
controlled manner and in accordance with specified procedures.
A thorough event review was performed for gripping a fuel assembly by its hold-down s3 rings during core unload.
The licensee's corrective actions were comprehensive and should prevent recurrence (Section 02.1).
Several discrepancies were identified in the licensee's implementation
.
of foreign material exclusion (FME) controls around the refueling cavity during Unit 2 core offload activities.
The licensee adequately resolved these discrepancies and improvement in FME implementation was noted during core reload (Section 02.1).
Unit 2 containment closecut following the refueling outage was adequate;
.
however, some debris was identified by the ins)ectors following the licensee's containment closeout inspection.
T1e amount of debris found would not have challenged containment sump performance (Section 02.2).
A Non-Cited Violation (NCV) was identified for failure to follow l
.
.
procedures during the Unit 2 Engineered Safety Features Actuation System l
(ESFAS) testing.
The operations test coordinator performed a step out l
of sequence that resulted in the inadvertent resetting of the required safety injection actuation signal (Section 04.1).
Enclosure 2
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
___
l
Maintenance Overall, 2R6 outage maintenance and surveillance activities were
.
3erformed satisfactorily.
Maintenance and surveillance personnel were knowledgeable of their assigned tasks.
Procedures provided adequate detail and guidance and were present at the work location and being followed.
The licensee established good coordination ar.d detailed 3re-briefings for the complex surveillance activities associated with t1e ESFAS and Emergency Core Cooling System flow balance testing (Sections M1.1 and M1.2).
An NCV was identified for the incorrect landing of a jumper which
.
resulted in the failure of the Auxiliary Component Cooling Water (ACCW)
pump No. 1 to start during ESFAS testing (Section M1.3).
A violation was identified for failure to adequately perform liquid
.
penetrant (PT) pre-examination cleaning for 10-inch safety injection system weld No. 21204-122-6 (Section M1.4).
Except for the violation regarding the failure to implement the
.
precleaning requirements of the PT procedure, inservice inspection (ISI)
activities observed / reviewed were conducted in accordance with procedures and regulatory requirements (Section M1.4).
i Enoineerino The electrical modifications on Unit 2 involving the 125-VDC ground
.
detection system. Emergency Diesel Generator Fuse protection, and Class IE Switchgear were being performed in accordance with the design change control program.
The material condition of the completed work was good.
Overall the design change packages and the 10 CFR 50.59 safety evaluations were adequate (Section E1.1).
The calculation performed to determine the battery ground maximum fault
.
resistance and to determine the new ground fault relay setpoint was adequate (Section E3.1).
An NCV was identified for an inadequate temporary test procedure to
.
,
retest secuencer actuation of ACCW pump No.1, which failed to start as
'
required curing ESFAS testing.
The procedure was inadequate, in that.
'
as written. ACCW pump No. 1 did not receive the correct start signal and other components actuated unexpectedly (Section E3.2).
Enclosure 2
_ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _
- _ _ _ _ _ _ _ _ _ _
l
Plant SupocIl Radiation and contamination controls for routine and 2R6 outage e
activities were appropriate and met applicable Updated Final Safety Analysis Report. Technical Specification. and 10 CFR Part 20 requirements (Section R1.1).
Two isolatea examples of poor radiological survey practices were
.
properly addressed by licensee management (Section R1.1).
,
Health Physics staff knowledge of survey documents required for the
.
unconditional release of bulk material from radiologically controlled areas was inconsistent with 10 CFR Part 20 (Section R1.1).
,
Worker doses from 2R6 outage activities were properly evaluated by the
.
licensee and were within 10 CFR 20.1201 limits (Section R1.2).
The reactor coolant cleanup and 2R6 chemistry programs were
.
appropriately managed and contributed to reduced dose rates
)
(Section R1.3).
A violation was identified involving a contractor failure to follow
.
requirements for maintaining proper control and visual contact of an escorted individual 'Section S3.1).
Enclosure 2
__
__
_ _ _ _ _ _ _ _
_
Reoort Details Summary of Plant Status Unit 1 The unit operated at full power throughout the entire inspection period.
Unit 2 The unit began the inspection period entering Mode 2 ia preparation for the sixth refueling outage (2R6).
Dt,rieg control rod insertion after entering Mode 3 a manual reactor trip was
- tiated following discrepancies between the demand position step counters and the digital rod position indicators.
The unit entered Mode 6 on March 12 with fuel offload completed on March 18.
Fuel reload was completed April 11.
At the end of the inspection period, the unit was in Mode 3 with preparations for Mode 2 in progress.
I.
Operations
Conduct of Operations 01.1 General Conduct of Unit 2 Sixth Refuelino Outaae (2R6) (Insoection Procedure (IP) 71707)
The inspectors reviewed the overall conduct of the refueling outage from a safety perspective.
The inspectors concluded from many observations and inspections conducted during the outage that plant safety and its enhancement were considered priorities by the licensee.
This was evident in several areas, includirig:
1) outage planning, which included thorough risk assessment of numerous electrical supply configurations:
2) scheduling 2R6 activities to avoid fueled midloop evolutions: 3)
management attention given to the use of contingency plans during periods of increased risk: and 4) the comp h ion of several modifications to enhance plant safety and reduce plant risk.
The overall conduct of the outage was identified as a strength.
01.2 Manual Reactor Trio durina Unit 2 Shutdown a.
Insoection Scope (71707 and 40500)
The inspectors reviewed the circumstances associated with the manual reactor trip from a hot standby condition initiated during the Unit 2 shutdown to begin refueling outage 2R6.
l Enclosure 2
_ _ _ _ _.
_
____-
_
____-
--
-
b.
Observations and Findinos On March 8. 1998. Unit 2 was being shutdown to begin refueling outage 2R6.
After insertion of the control banks and shutdown banks A and B
and during the insertion of shutdown banks E and D, the operators noted discrepancies between the demand position step counter and the digital rod position indicator (DRPI).
After confirming that DRPI was indicating properly. DRPI indication was used to fully insert the rods in these banks.
During attempts to insert bank C. there was no rod movement indicated by DRPI.
After discussions with management, a decision was made to manually trip the reactor due to the unusual circumstance. All remaining rods inserted into the core as required.
Subsequent troubleshooting of the rod control system during 2R6 identified a degraded integrated circuit board.
Westinghouse Corporation performed a failure analysis of the board which revealed a degraded transistor.
This condition caused continuous full current being supplied to the shutdown bank stationary coils and resulted in the erratic rod control system movement.
DRPI was checked out and confirmed to have been functioning properly.
Prior to startup, the degraded circuit board was replaced and the rod control system tested to verify that the system was functioning as designed.
The inspectors reviewed licensee Event Report 2-98-001, which addressed the investigation of the trip.
The resort was thorough and included a detailed review of the root cause of t1e rod control failure, assessment of operator response, and provided extensive corrective actions to address enhancements in the rod control system preventive maintenance and operations abnormal operating procedures governing rod control system malfunctions.
c.
Conclusions The inspectors determined that the decision to trip the unit was justified based on the observed rod position discrepancies between DRPI and the demand position step ccunters.
The licensee determined the cause of the rod control malfunction and corrected the equipment related degradation prior to unit startup.
01.3 Unit 2 Shutdown Risk Walkdowns and Contingency Plans a.
Insoection Scone (71707)
The inspectors reviewed the licensee's contingency plans, implemented during scheduled 2R6 periods of increased risk.
.
Enclosure 2 i
. _ _ _ _.
- _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _
__ -__ _
_
_- __
b.
Observations and Findinas During 2R6 the licensee established contingency plans and controls for periods of high ' shutdown risk.
Licensee personnel periodically walked down critical systems and instrumentation lineups to verify availability.
The inspectors also performed selected reviews of these
activities.
!
!
Licensee management performed evolution briefings in support of draining I
the reactor coolant system to 192 feet for vessel disassembly.
The inspectors observed portions of those briefings and concluded that they were appropriate.
On March 18. following full core offload, the Train A Nuclear Service Cooling Water (NSCW). Component Cooling Water (CCW), and Spent Fuel Pool (SFP) Cooling, systems were removed from service, leaving only the SFP Train B available for heat removal. The inspectors verified that SFP cooling was maintained operational, and that a feed and bleed
)ath, as l
described in Abnormal Operating Procedure 18030. " Loss of SFP _evel or
. Cooling." Revision (Rev.) 10. was available.
On April 2. following return of the Train A systems to service, the licensee removed Emergency Diesel Generator (EDG) 2B from service.
In addition.-Train B of the NSCW and CCW systems were removed from service, leaving the Train A SFP cooling to maintain temperature control of the pool during this period. Train B SFP cooling was available as the feed and bleed path during this period of work. The inspectors verified that Train A SFP cooling was the protected train and was fully operational.
c.
Conclusions Licensee management was directly involved in contingency plan walkdowns
,
and maintained a heightened awareness of the contingency plans during periods of increased risk.
Operational Status of Facilities and Equipment l
02.1 Core Off-load and Reload Activities i
a.
Insnection Scone (71707)
The inspectors observed portions of the reactor disassembly, core off-l l
load activities, and subsequent core reload and verification activities.
l b.
Observations and Findinas The inspectors observed portions of reactor disassembly, core off-load and reload activities from the control room, fuel handling bridge, and spent fuel pool area.
Core off-load was initiated on March 16 and completed March 18.
Core reload was initiated April 8 and completed Enclosure 2 l
- _ - - _____- _._
I
April 11.
The inspectors noted good pre)arations, coordination, and control of activities.
The inspectors o] served core verification performed immediately after the completion of core loading.
Additionally, the inspectors reviewed portions of the site reactor l
engineering core verification tape. All assemblies were identified as
!
loaded in their designated core reload location per the reshuffle plan.
l The inspectors determined that the licensee reloaded the core in accordance with its reshuffle plan.
During core off-load on March 17, with ninety-six fuel assemblies removed, fuel assembly ST11 was latched, removed from the core. and lowered into the upender basket of the fuel transfer system. At this time. it was observed that the assembly was not indicating in the full down position.
Camera inspections of the assembly revealed that the gripper was latched to the hold down springs instead of the lip of the top nozzle.
The assembly was unlatched and inspected for damage to the hold down springs.
No damage was identified.
The assembly was properly relatched to the top nozzle and transferred to the spent fuel Further visual inspections revealed no damage to the assembly. pool.
Following thorough testing of the fuel handling equipment, fuel off-load was recommenced.
The inspectors reviewed Event Report 2-98-002, which investigated the root cause of the incident. The licensee determined that the refueling bridge mast most likely was raised before the fuel assembly latching process had completed.
Indication that latching has been completed is from the illumination of a lamp light.
Mast movement is allowed following illumination of this light.
Corrective actions for this incident included revising the fuel handling procedure to require verification of proper weight indication during the first inch of upward movement of an assembly.
In addition, interlocks were added to the refueling machine that prevent upward movement if the gripper is in the process of latching or unlatching, and stopping upward movement if proper weight is not sensed in the first inch of movement.
Prior to core reload, the inspectors verified that these corrective actions were implemented.
The inspectors determined that the licensee's investigation and resolution of this incident were thorough and should prevent recurrence.
During the inspections on two different occasions, the inspectcrs noted problems in the licensee's implementation of Foreign Material Exclusion (FME) controls around the refueling cavity.
For the most part. these involved an oversight on the workers in the Zone II area to present materials taken into the area to the FME monitor for inclusion into the
,
i program log.
These items were resolved promptly after identification by the inspectors.
The inspectors noted a general improvement in the implementation of these controls in subsequent inspections conducted during refueling and reactor vessel activities.
i Enclosure 2 l
l l
_ _ _ _ _ _ _ _ _ _ _
.
.
- _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _
l c.
Conclusions
The inspectors concluded that refueling activities were performed in a controlled manner and in accordance with specified procedures.
A thorough event review was performed for gripping a fuel assembly by its hold-down s) rings during core unload. The licensee's corrective actions were compre1ensive.
02.2 Containment Closecut a.
Insoection Scone (71707)
On April 16, 1998, the inspectors conducted a walkdown of the Unit 2 containment to assess material condition prior to startup. At the time of the walkdown, the licensee had completed its containment exit ins)ection, containment integrity had been established, and the unit was in iode 4.
The inspectors reviewed Procedure 14900-C. ' Containment Exit Inspection." Rev. 3. used to perform the Unit 2 containment closecut.
b.
Observations and Findinas The material condition within containment was satisfactory however, the inspectors identified minor pieces of debris and several tools within readily accessible areas of containment. The inspectors estimated the total amount of debris and tools removed to be one cubic foot. The inspectors also identified several minor material deficiencies which were identified to the licensee for resolution. The inspectors determined that the debris and material deficiencies did not represent a substantial challenge to containment sump performance.
The inspectors discussed these findings with licensee management.
c.
Conclusions The inspectors concluded that the licensee adequately performed the Unit 2 containment exit inspection.
The debris identified would not have challenged containment sump performance.
The closeout of containment process conducted by licensee personnel was identified as a negative observation.
02.3 Reactor Trio Switchaear Seismic issues a.
Insoection Scone (71707)
,
The inspectors reviewed the licensee's actions as a result of a discovery that Reactor Trip Bypass circuit breakers had been left in the
!
" racked-out" versus " disconnected" position in the Unit 1 and 2 reactor protection system switchgear.
I Enclosure 2 t
i l
l
-
_ _ _ _
_ - _ - _ - _ - _.
.
.
.
._
-
l b.
Observations and Findinas
!
On April 2,1998, during performance of Trip Actuating Device Operational Tests (TADOT) on Unit 2 Reactor Trip switchgear. operations and maintenance personnel discovered the Train A and B Reactor Trip Bypass circuit breakers to be in the " racked-out" position versus the required " disconnect" position.
In the racked-out position, the breakers are not properly secured and present a potential seismic hazard to other equipment in the switchgear.
Subsequently, the Unit 1 reactor trip bypass breakers were examined and found to be in the same racked-out Josition.
Licensee personnel immediately racked all the bypass breacers to the disconnect (secured) position. The inspectors reviewed Deficiency Card (DC) 119980134 which documented this condition and immediate corrective actions. The DC assigned engineering to evaluate the Unit 1 seismic impact of the condition for re)ortability considerations.
The licensee initiated a formal Root Cause and Corrective Action (RCCA) investigation to determine if any generic problems existed with circuit breaker rack-in/ rack-out activities or switchgear configuration control.
c.
Conclusions Licensee actions to correct a mispositioned reactor trip bypass breaker condition and assess the immediate impact were prompt and technically a)propriate.
Pending review of the licensee's seismic evaluation and tie RCCA report findings, this issue is identified as Inspector Follow-up Item (IFI) 50-424, 425/98-03-01. " Reactor Trip Switchgear Seismic
)
Determination."
Operator Knowledge and Performance 04.1 00erations Performance Durina ESFAS Test a.
Insoection Scooe (71707)
'
The inspectors reviewed the circumstances surrounding the resetting'of a Safety Injection (SI) signal during the performance of the 18-month
Engineered Safety Features Actuation System (ESFAS) surveillance test.
i b.
Observations and Findinas l
On April 8. 1998 during the inspector's observation of the performance
of surveillance Procedure 14667-2. " Train B Diesel Generator and ESFAS Test." Rev. 5. Section 5.4. Diesel Generator Start on SI Signal, the-SI signal was reset prior to the required procedure step at the direction
,
L of the Test Coordinator.
Enclosure 2
_ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _
_ --
.
_
_
_
--
-- ---_ -_ -_ _-_ - _ - _ - _ -
_-
After discussions between the Shift Superintendent. Unit Supervisor, and the Test Coordinator, the Test Coordinator then re-performed the steps necessary to re-initiate the SI signal and the test was successfully completed without further deviations.
In subsequent discussions between the ins)ector and the Test Coordinator, he stated that he convinced limself that the procedure had omitted the step to reset and, therefore, decided to reset the signal.
However, the reset was addressed in the restoration section of the procedure.
He also stated that he convinced himself that it was acce) table to reset the signal because of his experience in performing mont11y slave relay surveillance tests.
During those surveillance, the procedure directs the operator to reset the signal.
Because he was working within that same cabinet, he believed that it was normal to
. reset the signal even though it was not specifically stated in that section of procedure 14667-2.
The inspectors concluded that the Test Coordinator failed to follow the written, approved procedure, and as such, did not meet the requirements of procedure 00054-C for aerforming steps in a stated sequence. Consistent with Section VII of t1e NRC Enforcement Policy, this was identified as Non-Cited Violation (NCV) 50-425/98-03-02. " Failure to Follow Procedure During Performance of ESFAS Test "
As a result of this issue the licensee implemented several corrective actions.
The surveillance procedure was revised to include a note that explicitly states that the signal is not to be reset during performance of that section of the procedure.
In addition, the Test Coordinator was counseled on the importance of following rules for the proper performance of procedures.
l c.
Conclusions The inspectors concluded that the licensee failed to follow procedure during the performance of a surveillance test, in that, the Test
!
Coordinator performed ~a step out of sequence, resulting in the inadvertent resetting of the required SI signal. The failure to properly perform the written, approved procedure as stated was contrary to procedural requirements and was identified as an NCV.
II.
Maintenance M1 Conduct of Maintenance
"
M1.1 Maintenance Work Order Observations (62707 and 92902)
i The inspectors observed portions of maintenance activities involving
,
several maintenance work orders (MW0s).
!
,
Enclosure 2
- _ _ _ - _ - _ _ _ - - _ - _ _ _ _ - _ _ _ - _
_ _ _ _ _
_ _ _ _.
._
.
l
Overall, maintenance personnel were knowledgeable of their assigned
)
tasks.
Procedures were present at the work location and being followed.
]
The procedures provided sufficient detail and guidance for the intended
maintenance activities.
1.
Emergency Diesel Generator Overhauls:
During the refueling outage, the inspectors observed various activities associated with the disassembly and reassembly of Train A and B Emergency Diesel Generators (EDGs).
Contractor and licensee pemonnel were sufficiently skilled to perform the assigned tasks, and licensee personnel coordinated with contractor representatives and licensee management to resolve identified issues.
In addition, the problems identified during post-maintenance testing were appropriately addressed and corrected.
2.
Split Pin Replacement:
The ins)ectors observed activities performed under MWO 29803100.
"Salit 31n Replacement." During replacement of the upper guide
)
tu)es, the upper guide tube at location D-2 did not sit properly on the lower guide tube. There was a gap between the upper and
'
lower guide tubes when the upper guide tube was installed.
Engineering personnel stated that the gap may have been there since original installation.
Design Change Package (DCP)
.'
98-V2N-0024. which included the placement of two 0.075-inch thick intermediate flanges, was developed and installed between the upper and lower guide tube to remove the gap. Observation and
'
con rection of the gap problem was an example of good attention to detail. The inspectors also noted that individuals performing the activities consistently used three-way communications.
The split
i
ain replacement work was performed under water for As-Low-As-
'
l Reasonable-Achievable (ALARA) considerations, which required the
-
l refueling cavity to remain filled after core unload.
'
M1.2 Surveillance Observation a.
Insoection Scoce (61726)
The inspectors observed the erformance or reviewed several of the following surveillance and lant procedures:
14666-2:
Train A Diesel Generator ESFAS Test. Rev. 6 14667-2:
Train B Diesel Generator ESFAS Test. Rev. 5 14721-2:
Emergency Core Cooling System (ECCS) Subsystem Flow Balance, Rev. 17 Enclosure 2
l b.
Observations and Findinas (
l During performance of ESFAS surveillance testing in accordance with l
procedures 14666-2 and 14667-2, the inspectors reviewed the documented
!
test results and compared them to the acceptance criteria for each section.
The ins]ectors also reviewed the exception reports end verified that eac1 was adequately dispositioned.
The pre-job briefings were observed Briefing summaries for each test section are included in the surveillance.
These briefing summaries are detailed and each action addressed. The inspectors noted that personnel responsible for plant actions during the test sections received copies of the briefing notes.
In addition, the surveillance tests were identified as infrequent activities, and as such, required management pre-job briefings.
The management pre-job briefings were conducted appropriately.
The pre-job briefings were conducted in the Unit 2 control room.
The ins)ectors observed that over twenty individuals were present for the pre-jo) briefings and expressed concern that the control room personnel could be distracted from their responsibilities.
Operations management stated that the briefings were held in the control room to ensure that the operator assigned to the EDG control board would be included.
Operations management further stated that alternative locations would be considered.
Both surveillance were satisfactorily completed and acceptance criteria met.
Overall, the test activities observed were well-controlled.
However, one NCV was identified for the test coordinator's failure to follow tha procedure as discussed in Section 04.1 of this report.
c.
Conclusions The observed surveillance activities were generally completed by personnel knowledgeable of their assigned tasks.
Procedures provided sufficient detail and guidance for the intended surveillance activities.
The licensee established good coordination and detailed pre-briefings for complex surveillance activities, including ESFAS and ECCS flow testing.
M1.3 Incorrectly Landed Electrical Jumoer a.
Insoection Scooe (62707)
The inspectors reviewed the circumstances surrounding an incorrectly landed electrical jumper in the Train A Sequencer Panel, identified by the licensee on April 15, 1998.
The jumper was identified by the licensee during troubleshooting for the failure of Auxiliary Component Cooling Water (ACCW) pump No.1 to start during ESFAS retesting.
Enclosure 2
_
_ _ _ _ _ _ _.
- - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ _ _ _ _ _ _
_
b.
Observations and Findinos During troubleshooting under MWO 29702891. Instrumentation and Control (I&C) personnel identified an electrical jumper that was incorrectly landed.
The incorrectly landed jumper resulted in the failure of ACCW pump No. 1 to start during ESFAS testing. The safety significance of the pump not starting on the sequencer start signal was minor in that the pump was able to start on low ACCW header pressure after the sequencer timed out.
The jumper was lifted and correctly landed.
The licensee determined that the last work performed in the Train A sequencer was surveillance test procedure 24613-2. " Safety Features Sequencer Train A Channel Operation Test and Channel Calibration." Rev.
14. on March 20, 1998.
Procedure 24613-2 did not address the jumper and did not provide steps for the disconnecting and landing of the jumper.
The individuals performing the work replaced the jumper where they thought it was previously connected.
Because the work was performed under an MWO. an available optim for documenting the lifting and landing of the jumper was use of Procedure 20249-C. "Short Term Documentation of Temporary Jumpers and Lifted Wires." However, no documentation for the lifteg and landing of the jumper was completed.
The licensee stated that similar work was performed on the other three sequencers and that the Unit 2 Train A Sequencer was the last one.
I&C personnel indicated that the other jumpers were also removed during performance of the tests and were landed correctly.
Licensee management stated that it was their expectation for individuals to.stop and correct procedure steps that are not correct for the plant configuration or that cannot be performed as written.
Technical Specification 5.4.1 requires that written procedures be established, implemented, and maintained covering the activities recommended in Appendix A of Regulatory Guide (RG) 1.33. Rev. 2.
Appendix A includes procedures for the performance of testing.
The incorrect lifting and landing of the jumper was not implemented in I
accordance with the surveillance steps of procedure 24613-2. Consistent with Section VII of the enforcement Policy, this licensee-identified and corrected violation is identified as NCV 50-425/98-03-03. " Failure to Follow Procedures Resulting in Incorrectly Landed Electrical Jumper."
c.
Conclusions The inspectors identified a non-cited violation for the incorrect landing of a jumper which resulted in the failure of the Auxiliary Component Cooling Water pump No. I to start during ESFAS testing.
I Enclosure 2 I
l l
. _ __ -- _ ___ _ _______ _ __ _ _ ___ _ _ __ _ _ _ _
- - - - _ _
M1,4 Unit 2 Inservice Insoection Activities a.
Insoection Scone (73753)
To evaluate the licensee's inservice inspection (ISI) program and the program's im)1ementation, the inspectors reviewed numerous procedures, observed wor ( in )rogress, and reviewed selected records.
Observations were compared wit 1 ap)licable procedures, the Updated Final Safety Analysis Report (UFSAR). and American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code. Sections V and XI, 1989 Edition, No Addenda (89NA).
Specific areas examined included the following observations:
1) liquid penetrant (PT) examination of Item Nos. 21204-034-10, 21204-044-1.
21204-044-2. 21204-122-6 and 21204-123-1: 2) manual ultrasonic (UT)
examination of Item Nos. 21204-044-1, 21204-122-6, 21204-123-1, 21301-002-3, 21301-003-1. and 21301-003-3: 3) visual (VT-1) examination of Item Nos. 21201-P6-022-B25-B32, 21201-P6-022-B33-B48 and 21204U6149:
4) data acquisition activities associated with eddy current (ET)
examinations of steam generator (SG) tubing; and 5) video tape of the remote visual (VT) examination of the reactor vessel.
The inspectors also reviewed selected completed examination reports and the Repair and Replacement Program.
Procedures reviewed included:
UT-V-404, " Manual and/or Mechanized Ultrasonic Examination of Full Penetration Welds." Rev. 9: MT-V-505.
l
" Magnetic Particle Examination " Rev. 4: PT-V-605, " Liquid Penetrant l
Examination Procedure," Rev. 3: VT-V-735. " Visual Examination (VT-3),"
Rev. 3: " Reactor Vessel Examinations 1998 Program Plan." Rev. 1: GBE-DP-
'
01. "DynaPulser Linearity Verification Procedure For Vogtle Unit 2:"
GBE-ISI-254. " Remote Inservice Examination of Reactor Vessel for the
,
Vogtle Nuclear Power Plant. Unit 2." Rev. 1: GBE-ISI-10. " Qualification of Ultrasonic Manual Equipment for Vogtle Unit #2," Rev. 1: GBE-ISI-54,
'
" Manual Ultrasonic Examination For Upper Shell To Flange weld For Vogtle Unit 2:" and GBE-ISI-88. " Underwater Remote Visual Examination For Vogtle Unit 2 Nuclear Power Plant," Rev. O.
The inspectors performed an independent evaluation of indications to confirm the licensee's ISI examiners' evaluations.
The inspectors also examined ET data for all indications that were 35% through-wall and greater.
The inspectors reviewed records for the nondestructive examination (NDE)
3ersonnel and equipment utilized to perform ISI examinations, including:
1DE equipment calibration and materials certification: and records attesting to NDE examiner qualification, certification, and visual acuity.
Enclosure 2
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _
- _ _ _ _ _ _ _ _ _ _ - _ - _
_ _ _
_ _ _ _
. _ _ _
-
l b.
Observations and Findinas
i Unit 2 SGs 2 and 3 tubing was subjected to ET examination. These
'
examination included:
100% full length dual bobbin: 40% dual + point rotating pancake (RPC) to) of tube sheet 3 inches; and 40% single t
row 1 and 2 U-bend RPC.
10 tubes exceeded the plugging limit of 40%
through-wall.
The licensee identified 42 indications in the reactor vessel flange to shell weld (Weld 03).
Using tip diffraction methods, the licensee sized the indications and determined that the indications were within the allowable limits of ASME B&PV Code paragraph IWB-3510-1.
These indications were located a) proximately 3.86 inches from the inside diameter (ID) surface of t1e wssel.
The indications were not identified during the baselin. examination conducted approximately ten years ago.
The longest indication (20 inches) produced the largest through-wall size by tip diffraction (0.23 inches): all other through-wall indication sizes, determined by the same method, were approximately 0.15 inches.
The licensee concluded that the indication's elevation, depth from the ID surface, and the practically identical location around the weld suggested a thin irregularly shaped embedded slag stringer.
The inspectors reviewed the licensee's evaluation and concurred with the licensee's conclusion.
The inspectors noted during the contractor-performed PT examination of 10-inch safety injection system weld No. 21204-122-6, that pre-examination cleaning was inadeqate.
The ins tenaciouslyadheringwhitedepositapproximatelykectorsnoteda x k inch, at the toe of the weld after precleaning and prior to the application of the penetrant.
This deposit was of a size that could mask relevant rejectable indications.
This improper precleaning was contrary to procedure PT-V-605. " Liquid Penetrant Examination Procedure " Rev. 3.
paragraph 9.2, which states in part that, "The examination area and the adjacent areas within at least 1-inch shall be dry and free of any dirt, grease, lint, water, scale, welding flux, weld spatter, oil or other extraneous matter that could interfere with the examination." Failure to implement ISI Procedure PT-V-605 is a violation of Technical Specification 5.4.1.a. which requires written procedures be implemented for the activities identified in Appendix A of Regulatory Guide (RG) 1.33. Revision 2. February 1978.
Appendix A requires that procedures for maintenance activities be implemented.
Immediately following the examinations, the licensee properly recleaned and satisfactorily PT examined weld No. 21204-122-6.
The licensee subsequently documented this issue in DC 219980276. This violation is identified as 50-425/98-03-04, " Failure of Contractor Examiner to Implement a PT Procedure Requirement."
Except as noted above. ISI examinations observed / reviewed were conducted in accordance with approved procedures, by qualified and certified examiners using certified / calibrated equipment and materials.
Enclosure 2
_ _ _ _ _ _ _ _ _ _ - _ _ - _ _
_ _ - _ _
_ - _ _ _ _ -
To evaluate the effectiveness of the licensee's Repair and Replacement Program, the inspectors reviewed records associated with MWO 29800310.
described as " Modified chemistry sample point down stream of valve 2-1609-U4-012 by welding in an elbow and tubing adapter per DCP 94-VAN 0057." The welding was accomplished by properly qualified welders using certified welding filler materials. The MWO was in compliance with ASME B&PV Code Section XI.
c.
Conclusions Exce)t for the violation related to failure to implement a requirement of t1e PT procedure. ISI activities observed / reviewed were conducted in accordance with procedures and regulatory requirements.
III.
Enaineerina El Conduct of Engineering El.1 unit 2 Modification Activities a.
Insoection Scope (37550)
The inspectors assessed the material condition and adequacy of the electrical modifications that had been partially completed or completed during 2R6 as listed below:
DCP 95-VAN 0038, 125 Vdc Ground Detection Modification (Completed)
DCP 95-V2N0064. D/G Fuse Protection (Completed)
DCP 96-VAN 0031. Class 1E Switchgear (Partially implemented).
Ap)licable regulatory requirements included 10 CFR 50. Appendix B: 10 CFR 50.59: and the licensee's design control and modification procedures, b.
Observations and Findinas The inspectors, accompanied by a plant modifications representative, conducted an inspection on portions of those electrical modifications
listed above that had been completed on Unit 2.
The inspectors observed i
that the modifications had been satisfactorily installed and that they were consistent with the design change details. Also, the inspectors noted that the quality and material condition of the modification work was good. The inspectors verified that )ost-modification testing was required to be performed or that it had 3een satisfactorily com)leted.
The inspectors reviewed in detail DCP 95-VAN 0038. Rev. 0, and t1e associated ground detection relay setpoint calculation. X3CF13. Rev. 1.
and found both to be adequate.
The 10 CFR 50.59 safety evaluations for Enclosure 2
_ _ _ _ _ _ _ -
._
_ - _ _ _ _ _ _.
__
_ _ _
__
_
_
the modifications were adequate and no unreviewed safety questions were
-
identified.
c.
Conclusions The electrical modifications on Unit 2 involving the 125-VDC ground detection system. Emergency Diesel Generator Fuse protection, and Class 1E Switchgear were being performed in accordance with the design change control program.
The material condition of the completed work was good.
Overall, the design change packages and the 10 CFR 50.59 safety evaluations were adequate.
E3 Engineering Procedures and Documentation
'
E3.1 Review of Calculations Sucoortina Electrical Modification a.
Insoection Scone (37550)
The inspectors assessed the adequacy of a calculation that established a new setpoint for the 125-VDC ground detection relay.
Applicable
,
regulatory requirements were delineated in 10 CFR 50 Appendix B and ANSI l
N45.2.11.
'
b.
Observations and Findinos
The inspectors reviewed Electrical Calculation X3CF13. Rev. 1, which recommended that the ground detection scheme be revised to provide only one detector per battery and that the ground fault relay setpoint be modified to allow for the detection and alarming of ground faults of approximately 155 kilo-ohms.
This new relay setpoint was needed to provide protection from the mis-operation of the most sensitive relays connected to the battery.
The inspectors noted that the assumptions and conclusions were well supported and technically adequate. The calculation was reviewed in accordance with the licensee's Quality
)
Assurance (OA) program. The recommendations noted in this calculation were implemented by DCP 95-VAN 0038 Rev. O discussed in paragraph El j
above.
c.
Conclusions The calculation )erformed to determine the battery ground maximum fault resistance and tie determination of the new ground fault relay setpoint was adequate.
Enclosure 2
_ _ _ _ _ _ _
__-
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _
15 E3.2 Inadeauate Temocrary Procedure for ESFAS Testina
i l
a.
Insoection Scooe (37551)
j i
l The inspectors reviewed the circumstances surrounding problems l
associated with the performance of engineering temporary test procedure T-ENG-98-05 " Sequencer UV Test For ACCW pump No. 1 (2-1217-P4-001)."
<
Rev. O.
j b.
Observations and Findinas l
!
During the performance of surveillance procedure 14666-2 " Train B l
Diesel Generator and ESFAS Test." Rev. 5. Section 5.3. the ACCW pump No.
l 1 did not auto-start within the required time following a loss of l
Offsite Power (LOSP) actuation signal.
A test exception report (TER) to 14666-2 was written.
Subsequent licensee troubleshooting determined that a sequencer relay. K107. may have failed to actuate and was the likely cause for the pump start failure.
Relay K107 was replaced.
Engineering personnel received a verbal request to develop a temporary test procedure for the retest of the ACCW pump No.1 on April 12, 1998.
Engineering personnel developed temporary test T-ENG-98-05. Rev. O. and it was reviewed and approved for use on April 14. 1998.
On April 15. 1998, during performance of T-ENG-98-05. ACCW aump No.1 i
again failed to start from the sequencer start signal and t1ree other valves repositioned unexpectedly.
The licensee determined that the repositioned valves constituted an Engineered Safety Features actuation and the incident was reported under 10 CFR 50.72.
In response t6 the incident, two actions were taken by the licensee:
1) review and revise
. procedure T-ENG-98-05, and 2) perform troubleshooting of the sequencer to determine why the pump failed to start.
While troubleshooting the sequencer under MWO 29702891, a misplaced
'
jum)er was identified.
The jumper was connected in a manner that prolibited the pump start from the sequencer signal. The jumper was returned to the correct alignment.
This is discussed in Section M1.3.
Additional review by the licensee of T-ENG-98-05 identified that there were two relays in series. (K106 and K107) which actuated on a sequencer start signal, and, which actuated on a loss of power signal.
Because the two relays are in series, both relays must be actuated concurrently or the start circuit will not be completed.
T-ENG-98-05.
Rev. 0. did not ensure that both signals were present concurrently.
Engineering also reviewed wiring diagram.s and identified seventeen additional relays that would receive a signal and reposition when the under-voltage test pushbutton was depressed.
Engineering personnel stated that the vendor manual contained inaccurate statements that led them to believe that there would be no automatic action from depressing Enclosure 2 l
!
- _ - _ - - - _ _ _ _ _ _ _ - - _ _ - _ _ _ _ _ -
_
the under-voltage pushbutton only.
However, using the wiring diagrams, it was determined that depressing the test pushbutton was sufficierit to initiate some plant components' response.
Technical Specification 5.4.1 requires that written procedures be established. implemented, and maintained covering the activities recommended in Appendix A of Regulatory Guide (RG) 1.33. Rev. 2.
Appendix A includes procedures for the performance of testing.
Implicit in this requirement is the stipulation that the procedures be adequate for the circumstances.
Failure to identify the seventeen plant components affected by the under-voltage test pushbutton signal and failure to identify the relay sequencing problem resulted in an inadequate procedure.
Consistent with Section VII of the enforcement Policy. this licensee-identified and corrected violation is identified as NCV 50-425/98-03-05 " Inadequate Engineering Temporary ESFAS Test Procedure."
T-ENG-98-05 was revi ed to include jumpering the under-voltage Relay K106 and to list the additional seventeen relays and plant components affected by the test. On April 16, 1998. T-ENG-98-05. Rev. I was performed successfully and ACCW pump No. 1 started within acceptance limits.
c.
Conclusions The inspectors concluded that the temporary ESFAS test procedure was inadequate. in that. ACCW pump No. I would not start from a sequencer start signal as written and the test did not identify seventeen additional plant components that received actuation signals from the use of the under-voltage test pushbutton.
An NCV was identified for the inadequate procedure.
E8 Miscellaneous Engineering Issues (92903)
E8.1 (Closed) Unresolved Item (URI) 50-424. 425/97-08-03 DtermineDesian Basis for Ambient Temperatures Used in Qualified Life Evaluation of Centrifugal Charoina Pumo (CCP) Motors. "
The licensee determined that an apparent inconsistency existed in the Updated Final Safety Analysis Re) ort (UFSAR) between Section 3.11.B and 3.11.N.3-1.
Section 3.11.B of tie UFSAR stated that the qualified life of environmentally qualified equipment was based on design temperatures.
l However. Section 3.11.N0-1 (Sheet 9 of 16) stated that the generic qualified life may be extended on a plant-specific basis by employing less conservative plant-specific assum3tions concerning the plant normal operating environmental conditions.
T1e latter position supports the use of the less conservative tem)erature monitoring data in the qualified life calculation for t1e CCP motors.
The licensee issued a revision to Section 3.11.B to add a statement that the qualified life of environmentally qualified equipment would be based on design or Enclosure 2
- ______ _
_ _ _ - _ _ _ _. -
_
._______-____-______ -____ _ ___ _ _ _ _ _
temperature monitoring data.
The inspectors reviewed the UFSAR and the proposed change and concluded that the calculation for the CCP motor qualified life evaluation was acceptable.
E8.2 (Closed) URI 50-424. 425/97-05-04: " Missile Protection for the Turbine Driven Auxiliary Feedwater (TDAFW) Pumo Exhaust Line."
After further review of the UFSAR for the TDAFW, the inspectors concluded that the as-built installation of the TDAFW pump exhaust piping was acceptable "as-is" without missile 3rotection.
If the TDAFW pump was lost due to a missile striking the ex1aust piping, the two motor driven pumps would not be impacted and one motor driven AFW pump, assuming a single failure, would provide adequate shutdown cooling.
E8.3 (Closed) Violation (VIO) 50-424. 425/97-01-04:
"Inadeauate Testina of 701 Governor."
(Closed) VIO 50-424. 425/97-01-05:
" Failure to Maintain Records of 701 DSC Setooints."
(Closed) VIO 50-424. 425/97-01-06:
"Unreviewed Safety Ouestion i
Determination for Woodward 701 Governor."
\\
The above violations were related to the modification that replaced the mechanical governor on both Units 1 and 2 EDGs with a microprocessor-based digital speed contro'ler. The inspectors verified by interviews and review of supporting documentation that the corrective actions described in the licensee's response dated May 9.1997. for each of the above violations were implemented and complete.
E8.4 (Ocen) IFI 50-425/97-10-05: "Samolina Proaram for the Unit 2 Eaton Cable Solices."
The licensee inspected 100% of those Eaton cable splices on Unit 2 that were accessible and found several splices that did not meet current environmental qualification (EQ) documentation in that the required transition splice of heat shrink material was not installed.
The Eaton cables were environmentally qualified with the cable jacket so a heat
'
shrink transition splice was required by design to replace the cable jacket at the termination or splice location.
The licensee documented the cable splice problems on multiple deficiency cards.
Some of the nonconforming cables were repaired in accordance with the design drawings and the rest were determined to be acceptable to "use-as-is."
This ~use-as-is" dis)osition was based on some new test data that had been obtained from t1e Eaton cable manufacturer's EQ files which demonstrated that Eaton cables without the cable jacket had successfully passed an EQ test.
This previous testing had not been used by Eaton to support qualification for the cables supplied to the licensee.
Enclosure 2
_ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _
_-
_
On March 24, 1998, the licensee audited the test report at the Eaton cable manufacturer's facility (now Furon Corporation in Aurora. Ohio)
and confirmed that the testing would support qualification of Vogtle's cables without the cable jacket being installed.
The licensee then drafted a revision to the Eaton Cable Equipment Qualification Data Package (EODP). X3AJ04 incorporating the new test data.
The inspectors reviewed the draft EODP and noted that it referenced a report that was located in the vendor's files in Aurora. Ohio.
The inspectors had a concern about referencing the new Eaton test data in the E0DP when the actual test report was not included in the licensee's EQ central files.
The licensee stated that the test report was being maintained by the vendor as a OA record.
The inspectors acknowledged that the vendor may be maintaining this document as a OA record: however it was not being maintained as a record for the qualification of the Eaton cables supplied to Plant Vogtle. The licensee stated that it was aursuing purchasing the test report from the vendor and expected to 1 ave it on site for review by June 1998.
This IFI remains open, pending review of
'
i the EQ testing performed on Eaton single conductors without cable
'
jackets.
E8.5 (Closed) URI 50-424. 425/96-10-06:
"Soent Fuel Pool Boraflex Concerns."
Degradation of boraflex material in spent fuel pool storage racks had been identified as an industry-wide issue for commercial reactors.
Analyses conducted on boraflex coupons from the Unit 1 spent fuel racks
indicated that the coupons were more degraded than expected. The L
licensee addressed this issue through additional criticality analyses of i
the Unit 1 and Unit 2 spent fuel pools, promulgation of administrative
]
controls for spent fuel pool fuel management and re-analysis of the spent fuel pools to eliminate the dependence on boraflex material.
Subsequent inspector follow-up of this issue reviewed each of these licensee actions, as well as the effectiveness of the boraflex i
surveillance program and interim and long term corrective actions.
The degrading boraflex condition was not safety significant because the level of soluble boron (-2000 ppm) in the spent fuel pool was enough to ensure subcriticality below the subcriticality margin requirements.
Generic Letter 96-04. "Boraflex Degradation in Spent Fuel Pool Storage Racks." issued by the NRC June 26. 1996. requested licensees to submit a description of proposed actions to confirm the 5% subcriticality margin for the lifetime of spent fuel storage racks. The licensee responded by 3 resenting a revised criticality analysis with no dependence on Joraflex. and requested technical specification changes to credit soluble boron in the spent fuel pools, spent fuel burnup and fuel storage methods for maintaining subcriticality associated with spent fuel storage. These changes were approved and issued February 20, 1998.
Based on the approved revision to the technical specifications, which effectively eliminate-issues associated with degradation and deterioration of boraflex material. URI 50-424, 425/96-10-06 is closed.
I Enclosure 2 l
- _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _
IV.
Plant Support R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Unit 1 and Unit 2 RP&C Controls a.
Insoection Scope (83750)
Radiological controls associated with on-going routine Unit 1 (U1)
operations and Unit 2 (U2) refueling outage activities were reviewed and evaluated by the inspectors.
Reviewed program activities included area 30 stings, radioactive waste (radwaste) and material container labels, ligh and locked-high radiation area controls, and procedure and radiation work permit (RWP) adequacy and implementation.
The observed program guidance and implementation activities were compared against applicable sections of the UFSAR and against approved procedures, technical specifications, and 10 CFR Part 20 requirements.
b.
Observations and Findinas All high radiation area postings and container labels were determined to meet the requirements for the associated radiological conditions.
Calibrations for ~in use" direct radiation and air sam)ler instrumentation were current.
Radiation controls for ligh and locked high-radiation areas. and for an "at power ~ U1 were implemented appropriately.
Pre-job ALARA briefings met procedural requirements.
During the week of March 23. 1998, two examples of workers providing ALARA program suggestions directly to the Health Physics (HP) staff regarding dog and radwaste reduction initiatives were noted.
Excluding two isolated Mervat bos of poor personnel and material radiological survey practices, activities associated with established RCAs were conducted in accordance with requirements.
The observed poor radiological practices included the failure to have monitoring equipment available for use by a worker exiting a remote radwaste Radiologically Controlled Area (RCA) and the improper removal of material from the main power block RCA subsequent to its causing a Small Article Monitor (SAM)-
9 alarm.
Both issues were addressed by licensee management.
The recent unconditional release and associated survey documentation for material, e.g.. metal containers, from established RCAs was determined to be appropriate.
However, the inspectors noted that for the release of selected materials e.g., scrap lumber, HP staff interpretation of survey documentation requirements was not consistent with 10 CFR Part 20.
Licensee management stated that additional evaluation of the current guidance and staff's knowledge to enhance this program area
!
would be conducted and changes made as necessary.
Enclosure 2 l
___________________ a
_ _ _ _ _ _ _ _ _ _ _ _
-_. _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _ __ _ _ _ _ _ _ _ __
!
!
l
,
c.
Conclusions Radiation and contamination controls met applicable UFSAR. Technical Specifications, and 10 CFR Part 20 requirements.
Isolated examples of poor radiological practices were identified and addressed by management.
Staff knowledge of survey documents required for the unconditional release of bulk material from radiologically controlled areas was inconsistent with 10 CFR Part 20 requirements.
R1.2 RP&C Dose Assessments a.
Insoection Scooe (83750)
The inspectors reviewed and discussed worker doses associated with 2R6 activities. Specifically, doses attributed to contamination events documented in personnel contamination reports (PCRs) and positive internal whole body counter (WBC) analysis results were reviewed and discussed with responsible licensee representatives.
Dose assessment methods and assumptions were reviewed for technical adequacy with results compared to 10 CFR Part 20. Subpart C.
occupational dose limits.
Results of " hot particle" shallow dose assessments were reviewed against guidance documented in Information Notice 90-48. " Enforcement Policy for Hot Particle Ex)osures." and NUREG/CR-5569. " Health Physics Position (HPPOS) Data 3ase." Rev.1.
b.
Observations and Findinas As of March 27, 1998, the licensee had documented and evaluated approximately 33 personnel contaminations.
Of these, approximately 16 were classified as PCRs. contamination levels exceeding 1000 counts per minute (cpm) per probe area, and were evaluated for assessment of shallow-dose equivalent (SDE) to the skin or extremities.
Skin dose for two of the PCR's requiring an assessment exceeded 100 millirem (mrem)
'
and were to be included in the individuals' permanent dose records in accordance with licensee administrative policy.
Preliminary calculations of the SDEs for the two individuals were 243 and 1.671 mrem and were within 10 CFR Part 20 limits.
Licensee assumptions regarding location of radioactive contamination or particles, shielding, and exposure times were appropriate.
Licensee records of WBC analyses conducted as of March 27. 1998, documented nine positive whole-body counts, six identified during routine termination analyses and three attributed directly to investigative analyses associated with potential worker internal contamination events.
Licensee assumptions regarding time and method of
L potential intake were appropriate.
Only four of the nine positive WBC Enclosure 2
- _ _ _ _ _ _.
._.
_
J
. _ _ _ _ _ _ _ _ _ _ _ _ _
results were assigned a committed effective dose equivalent (CEDE)
greater than or equal to 10 mrem.
Based on industry guidelines adopted by the licensee, the four calculated CEDES, ranging from 11 to 63 mrem, were to be included in the individuals' dose records.
c.
Conclusions Worker SDE and CEDE results from contamination events and work activities during 2R6 outage activities were properly evaluated and were within 10 CFR 20.1201 limits.
R1.3 Primary System Chemistry a.
Insoection Scooe (84750)
Unit 2 primary coolant cleanup and shutdown chemistry initiatives for 2R6 were reviewed and discussed with responsible licensee representatives.
Licensee activities and results were reviewed against applicable Technical Specifications, procedural requirements, and industry standards, b.
Observations and Findinas Licensee coolant cleanup initiatives included continued use of sub-micron filtration.
For the six months prior to shutdown. the reactor coolant sub-micron filtration was reduced to 0.1 microns.
Licensee representatives presented data indicating continued reduction of contact dose rates on cold leg, hot leg. and intermediate loop piping.
Between 2R5 and 2R6, dose rate reductions ranged from 26 to 47 percent for selected piping.
Routine shutdown chemistry was employed during the 2R6 outage and no concerns were identified during cooldown and initial cleanup.
Licensee representatives stated that changes in the outage schedule did not have an observable effect on system cleanup.
Trend data indicated a continued reduction of radionuclides inventory available for removal from i
the system with approximately 410 curies (Ci) of Cobalt (Co)-58 removed from the system during the 2R6 cleanup.
c.
Conclusions The reactor coolant cleanup and 2R6 chemistry programs were appropriately managed and contributed to reduced dose rates.
Enclosure 2
_ _ _ _ _ - _ - _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _
_ _ _ _ _ _ _ _
_ _ _
R8 RP&C Miscellaneous Issues (83750, 84750, 86750)
R8.1 (Closed) Escalated Enforcement Item (EEI) 50-424. 425/97-02-02: " Failure l
to Meet 49 CFR 173 475 Package Dose Rate Limits."
This issue involved the licensee's shipment of a radioactive material package with dose rates which subsequent to receipt at a vendor facility exceeded Department of Transportation (DOT) limits specified in 49 CFR 173.475.
Licensee corrective actions presented during an April 18,1997 NRC Enforcement Conference, and also detailed in a May 20. 1997, letter from C. K. McCoy, Vice President. Vogtle Project.
Southern Nuclear Operating Company, were reviewed.
In addition, supplemental commitments detailed in a June 3.1997. letter from J. P.
Jaudon. Director. Division of Reactor Safety. NRC Region II to C. K.
McCoy, Vice President,,Vogtle Project. Southern Nuclear Operating Company, were verified. As a result of the Enforcement Conference. EEI 50-424, 425/97-02-02 was closed and Violation (VIO) 50-424. 425/97-129-
'
01014 was opened.
The inspectors reviewed revised vendor procedures, veritied training
]
guidance, and attended a pre-job briefing associated with preparation
>
and packaging of vendor fuel sipping and reconstitution equipment for subsequent shipment.
Based on completion of licensee commitments, this VIO is closed.
R8.2 (Closed) IFI 50-424. 425/97-06-04: " Review Licensee Evaluation of Radiological Environmental Monitorina Program (REMP) Offsite C osscheck Results."
This item identified poor oversight in reviewing gamma spectroscopy system performance data and evaluating REMP offsite laboratory quality control (QC) analysis results. The inspectors reviewed and discussed revisions to procedure 33037-C. " Daily Quality Control of the Gamma Spectroscopy System." Rev. 5: offsite REMP vendor laboratory responses to a Quality Control review of the 1996 Radiochemistry Results dated July 18. 1997: and, results of the 1997 Summary of the Environtrental Laboratory Evaluation Program for environmental and effluent samples.
Based on improvements noted for licensee overview of effluent and environmental sample OC activities, this item is closed.
R8.3 (Closed) URI 50-424. 425/97-09-05: " Review Licensee Clarification of U1 and U2 Containment Miniource UFSAR Desian and Test Criteria and Evaluate the Adeouacy of the Containment Miniource Pressure Boundary Initial Test Criteria and Associated Results."
This item involved a review of the design basis and original acceptance testing conducted for the U1 and U2 Containment Minipurge Ventilation Systems based on identified leakage through ductwork fan belt openings.
Based on review of dose assumptions, original ventilation system licensing commitments, and pre-operational ventilation test data, the Enclosure 2
_ _ _ _ _ _ _ _ _ _ _ _
_ _ _ _
_.
_
_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ - _ _ _ _ -
-
inspectors noted that the "as-constructed" and tested system met the established commitments and requirements.
Furthermore, the inspectors noted that the January 27, 1998. Configuration Control Board minutes
!
documented that DCPs Nos. 97-V1N0064 and 97-V2N0065 were approved for replacing the currently installed equipment with axial drive fans.
The design changes were identified as positive program enhancements to eliminate the identified leakage associated with the minipurge ventilation systems.
This item is closed.
R8.4 (Closed) IFI 50-424. 425/97-09-06: " Review the Effects of Containment
)
'
Miniourae Leakaae on Site Emeraency Preparedness Activities durina Accident Conditions."
This item reviewed the current effect of minipurge leakage on the U1 and U2 equipment rooms during accident conditions.
From review of UFSAR dose estimates specified for selected accident scenarios and discussions with cognizant licensee representatives, the ins)ectors verified that the subject equipment rooms are not required to )e accessed by workers under accident conditions.
This item is closed.
S1 Conduct of Security and Safeguards Activities S1.1 Fitness for Duty (FFD)
a.
Insoection Scone (815021 The inspectors reviewed the licensee's corrective action in response to violation 50-424, 425/97-208-01014 issued June 25, 1997.
b.
Observations and Findinas NRC Inspection Report 50-424, 50-425/97-03 identified that the licensee i
failed to limit an individual's 3reliminary drug test result to the Medical Review Officer. Fitness or Duty Program Manager, and the r
employee assistance program staff, under the provisions of 10 CFR 26.24(d)(1).
The inspectors reviewed FFD records for the period of January 1. 1997.
to present for those individuals who had a positive drug test. Of those 22 records, the inspectors randomly chose six records.
Upon interview of licensee representatives and a review of records, the inspectors determined that preliminary drug test results were appropriately limited to those individuals described in 10 CFR 26.24(d)(1).
In a response dated July 25, 1997, the licensee documented a corrective action to i
reaffirm by letter, the understanding of the provisions of 10 CFR
'
26.24(d)(1).
The inspectors reviewed the affirmation letter, which outlined the requirements documented in SH-FFD-016. " Training and Qualification of FFD Personnel." Rev. 6, Attachment V.
Responsibilities of staff members and licensee management were clearly outlined; i e.
follow applicable procedures: consider that random screening lists. drug Enclosure 2
__
_
_ _ _ _ _ - _ _ _ _ _ _ _ _
a
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
-test discussions, releases, personal medical information, and information reviewed via the FFD program. to be private and confidential. Additionally, the affirmation letter specified that drug screening results, except those specified by existing procedures, shall not be discussed or otherwise communicated to individuals not associated with the administration of the program. The inspectors verified that the members of the FFD staff and associated licensee management confirmed their understanding and each signed an affirmation letter.
c.
Conclusions l
Corrective actions for Violation 50-424. 425/97-208-01014 were appropriate. Through record review and discussions with licensee representatives, the inspectors determined that the licensee was in compliance with the provisions of 10 CFR 26.24(d)(1) at the time of the
,
inspection.
This issue is closed.
l S3 Security and Safeguards Procedures and Documentation S3.1 Escort Duties and Control of Visitors l
a.
Insoection Scooe (71750)
i During the inspection period the inspectors observed the performance of escort duties for onsite visitors to support various outage activities.
!
b.
Observations and Findinas On March 30 and 31, 1998. during observation of maintenance on the l
Unit 2 Train B EDG. the inspectors observed the escorting of a cameraman and maintenance technician.
These individuals were being escorted by contractor personnel responsible for the maintenance work on the 2B EDG.
J On two separate occasions, the escort responsible for the cameraman did not maintain control or visual contact with that person as delineated in
,
Procedure 00652-C. " Personnel Escort-Duties and Responsibilities."
'
L Rev. 5.
Upon identification of the issue, the licensee took appropriate immediate compensatory action and counse'cd the involved individuals.
l-In accordance with Procedure 00652-C. the escort had the responsibility
!.
to control the movement of and maintain visual contact with escorted l
individuals. During observations inside the 2B EDG building on March l
30. the inspectors observed that the escort walked away from the visitor i
to the other side of the EDG without maintaining visual contact with the visitor. This was brought to the immediate attention of the Shift Superintendent.
However, on March 31 the inspectors again observed the same escort involved in a conversation on the upper EDG platform with
,
the visitor standing on the floor below. The inspectors observed that
,
l during the several minute conversation with maintenance technicians, the I
escort stood with his back to the visitor the entire time. As a result, the escort did not maintain proper control or visual contact of the Enclosure 2 J
- _ __ _ _ _ _ ___ _______- ________- _ _
visitor.
The inspectors concluded that the failure of the escort to maintain control and visual contact of an escorted individual on two occasions was contrary to the requirements of Procedure 00652-C.
This is identified as VIO 50-425/98-03-06. " Failure to Properly Escort j
Individual Inside 2B EDG Building.~
c.
Conclusions The inspectors concluded that a person responsible for escorting
'
visitors onsite did not follow required procedures.
The failure to properly maintain control and visual contact of an escorted individual was contrary to procedural requirements and was identified as a violation.
l l
F8 Fire Protection Miscellaneous Issues (92904)
F8.1 (Closed) VIO 50-424. 425/97-09-04: " Invalid Fire Extinguisher Inspections."
The licensee's corrective actions were summarized in a response dated November 14. 1997.
Based on a review of the licensee's corrective actions to date in response to fire extinguisher inspection and fire watch issues, this issue is closed.
l V.
Manaaement Meetinas and Other Areas X
Review of Updated Final Safety Analysis Report A recent discovery of a licensee o)erating its facility in a manner contrary to the UFSAR description highlighted the need for a special focused review that compares plant practices procedures and/or parameters to the UFSAR descriptions.
While performing the inspections discussed in this resort the inspectors reviewed the applicable portions of the UFSAR that related to the areas inspected. The inspectors verified that the UFSAR wording was consistent with the
'
!
observed plant practices procedures and/or parameters X1 Exit Meeting Summary I
l The inspectors ) resented the inspection results to members of licensee management at t1e conclusion of the inspection on March 13. 1998.
Interim exists were conducted on March 16, 27, April 3 and 10,1998.
l The licensee acknowledged the findings presented.
(
Enclosure 2
..
'
_ _ _ _ - _ _ _ _ _. _ - _ - _ _ _ _
_ _ _ _ _ _ _ _ - - _ - _ _ _ _ _ _
PARTIAL LIST OF PERSONS CONTACTED
!
Licensee J. Beasley. Nuclear Plant General Manager S. Chestnut. Manager. Operations G. Fredrick Plant Support Assistant General Manager J. Gasser. Plant Operations Assistant General Manager l
i l
K. Holmes. Manager. Maintenance
-
M. Sheibani. Nuclear Safety and Compliance Supervisor C. Tippins. Jr.
Nuclear Specialist I l
INSPECTION PROCEDURES USED IP 37550:
Engineering IP 37551:
Onsite Engineering IP 40500:
Effectiveness of Licensee Controls in Identifying. Resolving, and Preventing Problems IP 61726:
Surveillance Observation IP 62707:
Maintenance Observation IP 71707:
Plant Operations IP 71750:
Plant Support Activities IP 73753:
Inservice Inspection IP 81502:
Fitness for Duty Program IP 81700:
Physical Security Program for Power Reactors 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 IP 92902:
Followup - Maintenance IP 92903:
Followup - Engineering IP 92904:
Followup - Plant Support ITEMS OPENED AND CLOSED Ooened Typ_g Item Number Status Description and Reference IFI 50-424, 425/98-03-01 Open Reactor Trip Switchgear Seismic Determination (Section 02.3)
VIO 50-425/98-03-04 Open Failure of Contractor Examiner to Implement a PT Procedure Requirement (Section M1.4)
l Enclosure 2
_ _ _ _ _ - _ _ - _ _ _
-_
.
l l
!
VIO 50-425/98-03-06 Open Failure to Properly Escort Individual Inside 2B EDG Building (Section S3.1)
Closed Tvoe Item Number Status DescriDtica and Reference NCV 50-425/98-03-02 Closed Failure to Properly Perform Procedure During Performance of ESFAS Test (Section 04.1)
NCV 50-425/98-03-03 Closed Failure to Follow Procedure l
Resulting in Incorrectly Landed Electrical Jumper (Section M1.3)
NCV 50-425/98-03-05 Closed Inadequate Engineering Temporary ESFAS Test Procedure (Section E3.2)
URI 50-424. 425/97-08-03 Closed Determine Design Basis for Ambient Temperatures Used in Qualified Life Evaluations of CCP Motors (Section E8.1)
URI 50-424. 425/97-05-04 Closed Missile Protection for the Turbine Driven Auxiliary Feedwater Pump Exhaust Line
(Section E8.2)
VIO 50-424, 425/97-01-04 Closed Inadequate Testing of 701 Governor (Section E8.3)
l i
VIO 50-424. 425/97-01-05 Closed Failure to Maintain Records of 701 DSC Setpoints (Section
E8.3)
VIO 50-424. 425/97-01-06 Closed US0 Determination for Woodward 701 Governor (Section E8.3)
.
URI 50-424. 425/96-10-06 Closed Spent Fuel Pool Boraflex
!
Concerns (Section E8.5)
EEI 50-424. 425/97-02-02 Closed Failure to Meet 49 CFR 173.475 Package Dose Rate Limits (Section R8.1)
Enclosure 2
,
j
- _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _
l l
VIO 50-424, 425/97-129-01014 Closed Failure to Meet 49 CFR 173.475 Package Dose Rate Limits (Section R8.1)
!
IFI 50-424, 425/97-06-04 Closed Review Licensee Evaluation of REMP Offsite Crosscheck Results (Section R8.2)
URI 50-424, 425/97-09-05 Closed Review Licensee Clarification of U1 and U2 Containment Minipurge UFSAR Design and Test Criteria and Evaluate the Adequacy of the Containment Minipurge Pressure Boundary Initial Test Criteria and Associated Results (Section R8.3)
IFI 50-424. 425/97-09-06 Closed Review the Effects of Containment Minipurge Leakage on Site Emergency Preparedness Activities during Accident Conditions (Section R8.4)
VIO 50-424, 425/97-208-01014 Closed Failure to Maintain Confidentiality of a Preliminary Drug Test Result (Section S1.1)
VIO 50-424. 425/97-09-04 Closed Invalid Fire Extinguisher Inspections (Section F8.1)
Discussed Trae Item Number Status Description and Reference i
IFI 50-425/97-10-05 Open Sampling Program for the Unit 2 Eaton Cable Splices (Section E8.4)
(
l Enclosure 2
)