IR 05000424/1997009
| ML20212F378 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 10/20/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20212F296 | List: |
| References | |
| 50-424-97-09, 50-424-97-9, 50-425-97-09, 50-425-97-9, NUDOCS 9711040269 | |
| Download: ML20212F378 (30) | |
Text
.- .
. _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
. ,
.
U. S. NUCLEAR REGULATORY COMMISSION (NRC)
REGION 11 Docket Nos. 50-424 and 50-425 License Nos. NPF-68 and NPF-81 Report No: 50-424/97-09, 50-425/97-09 Licensee: Southern Nuclear Operating Company. In Facility: Vogtle Electric Generating Plant. Units 1 and 2 Location: 7621 River Road Waynesboro, GA 30830 Dates: August 3 through September 20, 1997 Inspectors: M. Widmann. Senior Resident Inspector (Acting)
K. O'Donohue. Resident Inspector T. Farnholtz Resident Inspector. V. C. Summer (Sections M1.3. E2.1)
G. Kuzo. Health Protection Inspector (Sections R1.1 R RI.3. R2.1 R3.1. R6.1. R8.1. and R8.2)
H. Whitener. Senior Reactor Inspector (Section M2.1)
Approved by: P. Skinner. Chief Reactor Projects Branch 2 Division of Reactor Projects Enclosure 2 9711040269 PDR 971020 ADOCK 05000424 G pm
_ _ _ _
-
. .
EXECUTIVE SUMMARY Vogtle Electric Generating Plant Units 1 and 2 NRC Inspection Report 50-424/97-09, 50-425/97-09 This integrated inspection included aspects of licensee operations, engineering maintenance, and plant support. The report covers a seven-week period of resident inspection. It also includes the results of an announced inspection by a regional health physics inspecto Doerations
. In general, the conduct of operations was professional and safety-conscious (Section 01.1).
- Plant control during the Unit 1 shutdown for the seventh refueling outage (1R7) and defueling activities was good. Activities were performed in accordance to written procedures and, in genera appropriate trending of data durir') the plant shutdown was performe In addition, pre-evolution briefings observed by the inspectors were informative and well-done (Section 01.2).
- The conduct of the core offloard was in accordance with written procedures and the evolutions observed by the inspectors in the spent fuel pool area were satisfactorily performed (Section 01.3).
. Licensee management's direct involvement in contingency )lan walkdowns and the heightened awareness of the o)erating staff to t w contingency plans during periods of increased risc was excellent (Section 01.4).
. The inspectors concluded that the revision process does not contain a positive mechanism to ensure that licensing document change requests is iniciated which will update the Updated Final Safety Analysis Report for procedure changes (Section 01.5).
. An example of poor work practices was identified in that the licensee did not properly establish adequate control of contractor work in the spent fuel pool. In addition, a decision by contract personnel to continue fuel pool activities after they had dropped a refueling tool was another example of a poor work practice (Section 01.6).
. The inspectors observed that plant review board (PRB) discussions during the inspection period were thorough and appropriately focused on safet The inspectors also concluded that PRB provided useful input and added value to the review process (Section 07.1).
,
Enclosure 2
_ - _ _ _ _ _ _ _ _ _ _
- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _
.
[o
-
. .
Maintenance
. Maintenance and surveillance activities were generally completed thoroughly and professionally (Sections M1.1 and M1.2).
. A violation was identified when maintenance personnel failed to generate the required work order prior to performing work on the 2B diesel generator to repair jacket water cooling system leaks (Section M1.3).
. Contractor (Cooper-Enterprise) personnel were highly skilled in the overhaul of the emergency diesel generator engine (Section M2.1).
. Good maintenance controls and management oversight were evident in performance of the Emergency Diesel Generator 1B engine overhaul (Section M2.1).
- A Non-Cited Violation (NCV) was identified due to an inadequate procedure revision which resulted in an inadvertent discharge of auxiliary feedwater (AFW) into the steam generators during the performance of a surveillance (Section M8.1).
Enaineerina
. The Design Change Packages (DCP) describing modifications to the Unit 1 auxiliary feedwater miniflow lines, local leak rate testing systems associated with the Unit 1 containment penetrations and the DCP to remove the last of the 3-hour Thermo-Lag fire barriers were reviewe No discrepancies or concerns were identified (Section E2.1).
Plant Sucoort
. In general, radiation and contamination controls for initial U1 RF7 outage activities were appropriate and met applicable TS and 10 CFR Part 20 requirements (Paragraph R1.1).
u
. Several instances of poor radiological work practices idantified by the inspectors were promptly addressed by management (Paragraph R1.1).
. Worker Shallow Dose Equivalent (SDE) and Committed Effective Dose Equivalent (CEDE) results from U1 RF7 outage activities were evaluated
"
properly and were within 10 CFR 20.1201 allowable limits (Paragraph R1.2).
. For the initial outage activities, approximately half of the Personnel Contamination Re) orts (PCRs) were attributed to poor radiological work practices by worcers (Paragraph R1.2).
. Skin dose assessments were to be enhanced by implementing ction points for inclusion of pure beta-emitting radionuclides (Paragraph R1.2).
Enclosure 2
-
-% -
_ _ _ _ . _ _ _ _ _ _
- - _ .
. .
. The reactor coolant cleanup and U1 Rr7 chemistry programs were managed a)propriately and contributed to observed reduced dose rates ()aragraph R1.3).
. The U2 containment minipurge ventilation equipment met design criteria and was installed in conformance with configuration control documents (Paragraph R2.1).
. Inspector followup item 50-424. 425/9'/-09-06 was identified to review the effects of containment minipurge leakage on site emergency preparedness activities during accident conditions (Paragraph R2.1).
. Program guidance and records for determining workers' prior yearly occupational exposure and recording particle SDE values met the intent of 10 CFR 20. Subpart L (Paragraph R3.1).
. Inconsistencies in documenting workers' previous annual occupational exposures were identified (Paragraph R3.1).
. Licensee Health Physics (HP) staffing and qualifications were adequate to support both routine and outage operations (Paragraph R6.1).
. A NCV for failure to monitor effluents released to unrestricted and controlled areas in accordance with 10 CFR 20.1302 requirements was identified (Paragraph 8.2).
. A violation was identified due to packages and materials permitted to enter a Unit 1 vital area without perform.ng a security search (Section S3.1).
. A violation was identified for fire extinguishers being used to su] port plant activities beyond the valid inspection date on the extinguisier An inspection of fire extinguishers was not performed within the 28-day
'
requirement for 54 extinguishers issued to plant personnel (Section F3.1).
Enclosure 2
_ _ _ _ _
- - _ _ _ _ _ _ _ _ _
.
.
. .
Reoort Details Summary of Plant Status Unit 1 operated at full power at the beginning of the inspection period until September 6, 1997, when the unit initiated power ' eduction activities in preparation for the Unit 1 seventh refueling out s3e (IR7). Unit I had completed 128 days of continuous operation prior to the planned shutdown. The unit entered Mode 6 on September 11. and fuel offload was completed on
. September 19. At the end of the inspection period, the unit was defueled with the reactor coolant system drained and outage activities in progres Unit 2 operated at full power throughout the entire inspection perio I. Ooerations 01 Conduct of Operations 01.1 General Comments'(71707)
Using Inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing plant operations. In general, the reviews indicated that the conduct of operations was professional, safety-conscious and performed in accordance with the technical specifications and written procedura .2 Unit 1 Shutdown (71707)
The inspectors observed the Unit 1 shutdown activities in preparation for 1R7. This included selected sections of the following evolutions:
power reduction and entry into mode 2: plant cooldown and entry into mode 3: placement of the residual heat removal system (RHR) in service:
entry into mode 5, and collapse of the pressurizer bubbl As part of Procedure 12006-C. Section D. Revision (Rev.) 41 " Reactor Coolant System (RCS) Cooldown to Cold Shutdown (less than 200*F)." a procedure ste) called for stabilizing the RCS temperature prior to the collapse of t1e pressurizer bubble. Operations considered that they had sto, ped the cooldown at 10:00 a.m. on September 8. 1997 and were sta)ilized. As a result, the control room oporators discontinued recording RCS and pressurizer temperature and pressure data. However, the inspectors noted that over the next 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> that the RCS temperature decreased an additional 36 F after the determination that the plant was " stable." Discussions with operations personnel indicated that the term " stable" in reference to the plant conditions was not well defined. At this time, the observations by the inspectors had minimal safety significanc Enclosure 2 l
_ _ _ _ _
_- .
..
-___- _ _ _ _ _ _ _ - - _ _ _
.
. .
,
?%
On September-11. 1997, the RCS was drained down to two feet below the vessel flange. .The inspectors observed portio.is of reactor disassembly activities. The reactor head lift activity was performed with no problems. Despite minor holdups associated with the unlatching of the control rods, the reactor internals were removed and 5tored without ite.iden Plant control was good, activities were performed in accordance with written procedures, and in general, appropriate trending.of data was observed by the inspectors. In addition, pre-evolution briefings observed by the inspectors were informative and weli-don .3 Cort Offload (60710)
-The inspectors observed portions of activities related to core offload and fuel handling in the spent fuel pool per procedures 93300-C. Re \
17. " Conduct of Refueling Operations." 93641-C. Rev. 7. "Develo) ment and Implementation of Fuel Shuffle Sequence Plan." and FP-GAE/GBE-FE "Vogtle Canister Sipping Procedure." The activities observed in the spent fuel pool included portions of the fuel shuifle to prepare for core reload, fuel maaping to verify that the assemblies were positioned in accordance with tie fuel shuffle data sheets, and fuel sipping of assemblies from 1R The conduct of the core offload was in acccrdance with written procedures and the evolutions observed by the inspectors in the spent fuel pool area were satisfactorily performe .4 1R7 Shutdown Risk Walkdowns and Continqency P_lans Insoection Scone (71707)
The inspectors reviewed the licensee's contingency plans implemented during 1R7 where periods of increased risk were schedule Observations and Findinas During 1R7. the licensee established cnntingency 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 review of these activitie Licensee management performed evolution briefings in support of draining the reactor coolant system to 192 feet for vessel disassembly. The inspectors observed portions of those briefings and concluded that they were appropriat Enclosure 2
'i
_ _ _ - _
$
-
. .
-3 Following core offload, the Train B 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 A available for heat removal. The inspectors verified that SFP cooling was maintained operational, and that a feed and bleed path, as described in Abnormal Operating Procedure 18030. Rev. 10. " Loss of SFP Level or Cooling." was availabl On September 23, 1997, following return of the Train B systems to service, the licensee removed diesel generator 1A from service. In additior to NSCW. CCW. and SFP cooling. Train A systems were removed from service. leaving the SFP Train B cooling to maintain temperature control of the )ool during this period. As before, the inspectors verified that t1e SFP cooling was the protected train and fully operationa Conclusions Licensee management's direct involvement in contingency )lan walkdowns and the heightened awareness of the o)erating staff to t1e contingency plans during periods of increased risc was excellen .5 Fuel Handlina Buildino (FHB) Ventilation Insoection Scone (71707)
Based on a Deficiency Card (DC) generated on a FHB ventilation issue, the inspectors reviewed portions of the FHB system design. The inspectors reviewed procedure 13320-C. " Fuel Handling Heatin Ventilating and Air Conditioning (HVAC) System." Revisions (Rev.) 17 and 18. DC 1-97-357, design change package 96-VAN 0039. " Fuel Handling Building Post Accident Ventilation." procedure 00051-C. " Procedures Review and Approval." Rev. 24, procedure 00402-C. " Licensing Document Change Request." Rev. 16. procedure 00056-C. " Safety and Environmental Evaluations." Rev._15. and system drawings. The inspectors also discussed with licensee operations and engineering management the review of this issue, Observations and Findinas During the inspectors * review of the FHB post-accident HVAC system, the inspectors determined that the licensee had implemented a revised operating procedure. The change was processed in accordance of 10 CFR 50.59 procedures: however, the procedure to revise the Updated Final Safety Analysis Report (UFSAR) was not initiated. No licensing document change request (LDCR) or design change package (DCP) was ap) roved at the time of the procedure revision. The licensee actnowledged the inspectors' findin Enclosure 2 i
I
- _ _ - _ _ -_ - - _ - _ . - _ _ - _ _ - - - - - - . _ _ - _ _ - - - - _ - .
.
.
. .
On July 17, 1997, procedure 13320-C. Rev. 18, was approve One of the changes in Rev.18 deleted the requirement to enable the low differential pressure (d/p) automatic actuation by taking the low differential pressure switch. A-HS-2533C out of block. In accordance with the 3rocedure revision. this action was permitted while moving loads wit 11n or over the SFP with normal FHB ventilation in servic As part of the procedure change process, a safety evaluation, procedure 00056-C. Rev.15. "10 CFR 50.59 Evaluation." was Jerformed and it determined that no unreviewed safety question existed. lowever, the safety review screening identified that the procedure change did constitute a change to procedures as described in the UFSAR because UFSAR section 9.4.2.2.2.3. System Operation, states that the actuation circuit for low d/p is normally blocked, except during fuel handiing activitie The safety evaluation screening further stated that the UFSAR would be updated as part of a design change The DCP process would include issuance of an LDCR which would ei ure that the UFSAR was updated to reflect the plant operations. However, the inspectors noted that if the DCP was not approved, the LDCR would not be initiated and the UFSAR would not be updated. The licensee stated that there was no formal mechanism that ensured that the UFSAR was updated when a procedure change is processed, Conclusions The inspectors concluded that the revision process does not contain a positive mechanism to ensure that an LDCR is initiated which will update
<
the UFSAR for procedure change .6 Fuel Handlina Tool Event Insoection Scone (71707)
The inspectors reviewed the dropping of a fuel handling tool into the SFP which occurred on August 7. 1997. The inspectors reviewed Procedures 00801-C. " Control of Onsite Contractors. ' Rev. 11. 11951- " Fuel Handling Operator and Fuel Handlirig Supervisor Qualification and Review Checklists." Rev. 15. 93100 C. " Refueling Tools and Equipment Preservice Inspection / Checkout." Rev. 17. and 93220-C. " Fuel Handling Machine Operating Instructions." Rev. 11. in addition to the D I personnel statements, and the recovery plan developed as a result of the '
event. The inspectors also interviewed the involved Westinghouse contractors and licensee health protection personne Enclosure 2 a
_ _ _ _ _ _ _ _ _ _ _ _ . __
- - -_ _ _ _ _ _ _ - _ - _ _ - _ _ _ _ _ -_ __ .
.
-
. .
i 5 Observations and Findi.ngs I On August 7. Westinghouse f,ontract personnel were in the process of perform 1nt SFP tool handling checkouts in the Unit 1 bFP when a burnable poison rod assecbly (BPRA)- tool was knocked off its storage rack in the SFP and fell. The BPRA tool came to rest with-one end against the east fuel pool cask loading pit door and the other end on the fuel pool cavity floor. A camera was used to inspect the SFP hner for damage rid revealed no significant indications of damage as a result of the dropped tool. However. A small " thumb size" indentation was identified. but no cracks were observed in the general area of the indention. On August recovery plan T-ENG-97-11. " Recovery of_A Drop)ed BPRA Tool in the Unit 1 Spent Fuel Pool." Rev. . was implemented. Recovery of the tool was achieved without incident
, Based on interviews, the Westinghouse personnel involved stated that 4 during a thimble plug tool checkout a fuel pool bridge operator inadvertently bumped the BPRA tool. However. instead of step)ing and addressing the incident, the bridge operator continued on wit 1 the thimble plug tool checkout. After completion of that activity the cperator then stopped work. The inspectors were informed by the lead health physics (HP) technician that the bridge o)erator did not respond to a "stop" command after it was recognized by tie HP tec'inician that-the BPRA tool had been dropped. The licensee attributed the dropped tool incident to cognizant personnel erro A review of contract Jersonnel qualifications as fuel handling machine operators indicated tlat involved personnel were qualified in accordance ( with Procedure 11951- As-a result of this event, the licensee met with contract personnel to address issues of conservative decision making. In addition, the licensee rebuilt the dropped BPRA tool to support spent fuel refueling operations for the ongoing 1R7 outage, Conclusions An example of poor work practices was identified in that the licensee did not properly establish adequate control of contractor work in the spent fuel pool. In addition, a decision by contract personnel to continue fuel pool activities after they had dropped a refueling tool was another example of a poor work practic .
Enclosure 2
_ ___
j
_
.
.
. .
+
071 . Quality AssuranceLinl Operations
,
LO7;1- Plant Review Board Meetinas (4050Q).
- The inspectors attended a Plant Review Board (PRB) meeting on August 1 !1997.-- - The meeting was a normally-scheduled--PRB meeting and the majority-of the items discussed were routine in nature. Included in the PR . review at the meeting was =the licensee's response to NRC violation-50---
- 424. 425/97-07-01. concerning the-improper entry of a technical specification-(TS) action statement on the containment personnel'
airlock. In addition, draft Licensee Event Report (LER) 1-97-00 " Hydrogen Monitoring System Train Rendered Inoperable." was reviewe The PRB discussions were thorough and appropriately focused on safet The inspectors also concluded that PRB provided useful input and overall added value to the review proces '
-II. Maintenance M1 Conduct of Maintenance M1.1 Maintenance Work Order Observations
! Insoection-Scone (627071-The inspectors observed portions--of maintenance activities-involvirg the following work orders:
=196025051 -Disconnect control rod drive motor cables
'
19602783- -Polar crane preventive maintenance 19604001 Install mid-loop sightglass
-- 19604108 - Remove transfer tube blind flange 19604109 Disassemble'conoseals:
19604112 --Remove reactor head insulation-
=
19604114- Preparations for reactor head. detention 19604120 Lift reactor head and place in stand 19700229 Inspection of-1PSV3001. Main Steam (MS) Safety Relief Valve Loop 1
,
19700235 Inspection of 1PSV3002. MS Safety Relief Valve Loop 1
' 19700240 Inspection of IPSV3003.:MS Safety Relief Valve Loop 1 19700241- Inspection of 1PSV3004. MS Safety Relief Valve Loop 1
- 19700242 Inspection of 1PSV3005..MS Safety Relief Valve Loop 1 19700857 Reactor vessel seal plate placement and Unit 1 guide tube support pins-(split pin) replacement (
19701634 Centrifugal charging pump train A gear inspection
- 19702362 Diesel generator IB air dryer replacement ;
29702214' - Auxiliary component cooling water pump 2 discharge drain valve weld repair Enclosure 2
.
--
. . . . _ - . . . . . . . .
.. _
- - - _ - _ _ _ _ _ _ _ _ ,
.
.
.- .
7 Observations and Findinos
'The_ observed maintenance activities were generally completed thoroughly-and professionall M1.2 Surveillance Observation Insoection Scoce (61726)-
,
'
The inspectors observed the performance or reviewed the following surveillances and plant procedures:
14445-1 Remote shutdown monitoring instrumentation channel check, Rev. 4 14445-2 Remote shutdown monitoring instrumentation channel check, Rev. 3 14450-1 Reactor coolant system pressure isolation valve inservice leak test Rev. 23 14617-2 Solid state protection system (SSPS) slave relay K 609 train B test safety in'ection. Rev. 6 14619-2 SSPS slave relay rain B test safety injecti o Rev. 5 14623-2 SSPS slave relay K 615 train B test safety injection. Rev. 4 14802-2 Nuclear service cooling water (NSCW) pumps and check valve inservice test (IST) and response time test Rev. 13 14808-2 Centrifugal charging pump train A and check valve IS Rev. 15 1481011 Turb1ne driven auxiliary feedwater pump and check valve
- response time test, Rev. 24 14903-1 Containment emergency sump inspection. Rev. 7'
14915-1 Special conditions surveillance logs: data sheet 12, containment temperature monitoring with function A fire detection instruments inoperable. Rev. 31 24697-1 Nuclear instrumentation system intermediate range channel IN35 (compensating-voltages). Rev. 23 24739-1 Peak acceleration recorder 1AXR-19910. Rev 5 24741-1 Peak acceleratien recorder 1AXR-19913. Rev. 5 28210-C Main steam line code safety valve setpoint verificatio Rev. 13 Observations and Findinos The' observed surveillance activities were generally completed thoroughly and professionall ,
Enclosure 2 l
l
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _>
"
y l . :.
3 9
[IL
=8
-
M1.3 r Diesel Generator-(DGf Ma'intedance Without AcoroDriate Work Orders (WO)
'a; InsDection Scooe (62707)
.
.
l .
m The inspectors = observed the monthly surveillance of the 2B DG during which several leaks in the jacket water cooling system were identifie =The. licensee conducted maintenance on this system to repair.these' leak Observations and Findinas- 3 On Se)tember 11:.~ 1997 bduring a routine monthly surveillance of the 28 -
DG, tie inspectors identified several significant leaks in the DG jacket water cooling system. The inspectors notified the DG operator who in turn' contacted the res)onsible system engineer to evaluate these leak All the identified leats were from threaded union connection #
Maintenance personnel were contacted to repair the leaks while the DG continued to operate and the jacket water cooling system was at normal operating pressure,. With the system engineer observing, maintenance-
-technicians tightened the' unions which stopped the leak The_ inspectors reviewed the requirements of procedure 00350-C. " Work Request Program." Rev, 29. Paragraph 4.1.1 requires that a-WO be
. generated to provide ad accurate description of the cndition prior to maintenance being_ performed. A W0 was not' processed to perform this
. activity, Conclusions Maintenance personnel failed to ' generate the required WO prior to T performing work on-the 2B-DG to repair several jacket water cooling system-leaks as required by Procedure 00350-C. This is identified as
. Violation.(VIO) 50-425/97-09-01. Failure to Obtain a Work Order Prior to Conducting Mdintenance on a Diesel Generato i M2" -Maintenance and Material Condition of Facilities and Equipment-
'M2.1 Diesel Generator Ten Year Insoection and Overhaul
.. a . Insoection Scoce (62700)
The ins)ectors observed various activities associated with -the tear-down-of the Emergency Diesel Generator (EDG) IB Engine for the 10-year overhaul and inspection including contractor.(Cooper-Enterprise)
maintenance. practices:-disassembly of the machine: NDE testing of components: quality documentation of compcnents: and coordinated management oversigh >
Enclosure 2
. _ - - _ - _ _ - _
_ _ _
..
.
. .
.The 10-year tear-down and inspection was performed according to procedures and guidance as follows:
-
55052-C. " Standby Diesel Generator Maintenance Program." Rev. January 30. 199 GEN-95. " Diesel-Generator Surveillance Manual." Rev. January 23. 199 Cooper-Enterprise Service Information Memo. S.I.M. #402A. " Cooper-Enterarise R4/RV4 Preventive Maintenance Program (PMP) For Nuclear Stand)y Applications." Rev. Letter. "10 CFR Part 21 Report Power Cylinder Liner." from Cooper to the NRC. d6ted July 9. 199 Oberservations and Findinas The inspectors reviewed appropriate sections of the above document witnessed portions of disassembly activities and observed contractor maintenance practice The principle document utilized in this portion of the EDG overhaul was GEN-95 which implements the preventive maintenance (PM) program, including the inspection and measurement surveillance requirements recommended by the Cooper-Enterprise Owners Group (formally the-Transamerica Delaval Inc. (TDI) owners group). Cooper now provides the services as the EDG manufacture The ins)ectors observed that the contractor was well organized. The job was brocen down into specific individual tasks. such as " remove heads" or " remove liners." Work packages containing step-by-step instructions, detailed data sheets and acceptance criteria were developed and used for each task. Controls were established for foreign material control and for identifying, tagging. covering and protecting disassembled component Contractor personnel were skilled at their assigned tasks. activities were well documented, and coordinated management oversight was maintained by Cooper 0A/0C representatives and Southern Nuclear Company maintenance and engineerin c .- Conclusions The EDG contractor (Cooper-Enterprise) was highly skilled in the overhaul of the Transamerica Delaval engin Good maintenance controls and management oversight were evident in the performance of the EDG 18 Engine overhau Enclosure 2 l
.
..
. .. ..
. .. .-. .
.
=-- - - - - - - - -_ - _ ---- ___ _ - - - - - -
+# .
'
. ,
-10 -
-M8 -Miscellaneous Maintenance issues (92902)
--M8.1; (Closed) Licensee Event Reoort (LER) 50-424/97-005:' Test Procedure-Inadequacy Results In Auxiliary feedwater-Actuation (AFW)
- .Insoection Scoce (61726)
-The inspectors reviewed the August 4, 1997 inadvertent discharge of AFW-into the steam generators during the performance of Surveillance 14810-1, " Turbine Driven Auxillary Feedwater (TDAFW) Pum-Inservice Test-(IST) Response Time-Test " Rev. -The 23.p inspectors and Checkalso Valve-reviewed the following documents:
Surveillance 14810-1. "TDAFW Pump and Check Valve IST Response Time Test." Rev.:2 Surveillance 14810-1. "TDAFW Pump and Check Valve IST Response Time Test." Rev.' 2 LER 50-424/97-005. '" Test Procedure Inadequacy Results In Auxiliary Feedwater Actuation." --
~ Unit-1-Reactor 0perator and Unit Shift Supervisor control- room lo The inspectors also interviewed the involved personnel and discussed the licensee's root cause determination with licensee managemen '
' Observations and Findinas:
,
On August 4. '1997. operations personnel performed procedure 14810- During the test, the control room operators noticed a slight rise in reactor power. realized that the TDAFW pump discharge. valves were open, and immediately closed the valves. The addition of AFW to the steam-generators resulted in a power increase to 100.5 percent. The reactor power was returned to 100 percent within approximately ten minute Operations personnel later determined that when step 5.2.5 of procedure 14810-1 was performed, which energized TDAFW actuation relay AX-1. an ,
open signal was received at all four of the TDAFW motor operated
- discharge flow control valves. Because the TDAFW pump was operating at that time. AFW was fed to all four steam generator .-
The licenst:e had revised Procedure 14810-1 on-May 30, 1997, to delete a step that manually closed the TDAFW discharge isolation valv U4-015. Procedure 14810-1 had been revised to comply with an NRC commitment to maintain that valve open at full power operation. The procedure revision review performed by o)erations management did not recognize that-the: deletion of closing t1e manual valve would result in AFW injection into the steam generator .
,
Enclosure 2
-
o l' 't,
- .
.,
. .
11- o
'
The licensee reviewed the event and determined that it was caused by an
. incorrectly revised procedure. Corrective: actions identified -included W revision of procedure 14810-1--to allow response time testing without tallcwing AFW to flow into the steam generators. dn addition. the findividual who prepared .the procedure revision and management involved in the procedure review
,
of attention to detail. process Also.were' counseledofregarding a discussion the the the event-in importance-next-segment of requalification training is scheduled for operations perso.ne '
- Conclusion-
'
i The inspectors concluded that this event was the-result of an inadequate ~
procedure and cognitive personnel error during the review proces Conristent!with Section VII of the Enforcement Policy, this is'
ider'.ified:as Non-Cited Violation (NCV) 50-424/97-09-02. Procedure
,
-Revision Results in Auxiliary Feedwater Actuatio III'. Enaineerina E2 Engineering Support of Facilities and Equipment E2.1 -Review of Desian Chance Packaaes (DCPs)
a. -Insoection Scooe (3755 H The inspectors reviewed three DCPs for modifications to the-Unit 1 Auxiliary Feedwater (AFW) system, the LocalzLeak Rate Testing-(LLRT)
systems associated with the Unit 1 containment penetrations. and the
. removal of Thermo-lag from within the Unit .1 containment. The, ,
ins)ectors also discussed the-DCPs-with the responsible engineers and
.tec1nician b. ' Observation's and Findinas The inspectors reviewed-DCP 95-V1N0018. "AFW-Pamp Miniflow Line
- Modification." Revision (Rev.) 1. -The modification described in the DCP would provide' the ca) ability to return the miniflow of an o)erating 1 Unit 1 AFW Jump to tie same Condensate Storage Tank (CST) tlat was supplying tie pump suction. The exi:tino uesign directed miniflow from n the two motor driven AFW pumps to CST 2 and the turbine driven AFW pump to CST 1.:with suction capability for all the AFW pumps from-either CST 1:cr 2. In addition.-the DCP would add a second orifice in parallel with the existing orifices for the motor driven AFW pumps to increase the miniflow capability to that recommended by the manufacturer of the pum Enclosure 2
. , . .
..
.
.. .
_
- __ _ - _- _ _____ - ___ ___
.
. .
.
The modification described in this DCP had already been completed on the Unit 2 AFW system. The-insp'ctors walked down the completed Unit 2 AFW miniflow line modification and the Unit I work in progress. No concerns were identified with either the design modification or the progress of the work on Unit The inspectors also reviewed DCP 94-V1N0002. " Replacement of Blind Flanges to Support LLRT Testing." Rev. O. The purpose of this modification was to replace the existing isolation valve and blind flange arrangement at the LLRT test points for each containment penetration on Unit I with either a single isolation vaive and a threaded cap or a double isolation valve and a threaded cap. This would provide an easier and more accessible connection point for LLRT tests, in addition, this modification would reduce the possibility of damage to LLRT test connections due to stresses introduced during the flange torquing process. A total of 46 valve / flange connections were included i in the scope of this modification. Forty-four of these were designated to be replaced with a single valve and threaded cap while two. located at penetration 50. were to receive double isolation valves and a threaded ca DCP 94-VIN 0061 " Containment Building Elimination of Thermo-Lag Material" was performed to delete the last of the 3-hour fire barrier materials located in containmen This DCP was successfully implemented and complete c. Conclusions The DCPs describing modifications to the Unit 1 AFW miniflow lines. LLRT test points associated with the Unit 1 containment penetrations, and the removal of Thermo-Lag were reviewed in detail. No discrepancies or concerns were identified during these review IV. Plant Sucoort R1 Conduct of Radioloaical Protection and Chemistry (RP&C) Activities (83750. 84750)
R1.1 Unit 1 (U1) RP&C Controls Insnection Scone Radiological controls and chemistry operations associated with Unit 1 Refueling Outage Seven (U1 RF7) activities were reviewed and discusse The reviewed controls included area )ostings, radioactive waste (radwaste) and material container la3els, high and locked-high radiation area controls, and procedural guidanc Enclosure 2 I
_
___ __
.
. .
.
- The ins)ectors made frequent tours of the radiologically controlled areas (RCAs) associated with radioactive material and equipment storage areas. outage equipment staging areas and radwaste processing and storage facilitie Planning, special radiological controls, and procedural guidance associated with projected high person-rem outage tasks were reviewed and discussed in detail-. Procedural and radiation work permit (RWP) guidance for on-going activities and associated survey records also were reviewed and discussed with res)onsible Health Physics (HP)-staff. The inspectors directly observed teclnician performance and discussed radiation-and contamination controls for s)ecific U1 RF7 tasks conducted within selected containment and auxiliary auilding location Established guidance and implemented controls wers compared against Updated Final Safety Analysis Report (UFSAR) SNtion (S)12 detail Technical Specification (TS) S 5.7 and documcoted requirements in applicable sections of 10 CFR Part 2 Observations and Findinas Excluding some staged egoioment which extended across RCA boundaries ,
located in outside areas physical controls associated with radioactive material, equipment, and waste storage and staging areas were identified and posted properly. Area postings and container labels were accurate and in accordance with TS or 10 CFR 20 Subpart J requirement Radiological controls associated with containment high radiation and locked-high radiation areas were in accordance with TS requirement The inspectors directly verified implementation of specific radiological controls associated with refueling operations specified in licensee procedures 12007-C. Refueling 0)erations (Mode 5 to Mode 6) Revision (Rev.) 40 and 43014-C. Special Radiological Controls. Rev.1 During tours of containment, the inspectors noted several examples of poor radiological practices associated with protective clothing (PC)
use. Identified poor practices included unfastened or opened hoods and coveralls, reaching within PCs with potentially-contaminated gloves to
'
retrieve personal items, and carelessness in removal of PCs when exiting the U1 containment area. On September 17. 1997, licensee management addressed improper work practices during morning meetings and implemented a 15-minute stand-down of work activities at noon to enhance worker awareness of proper radiological work practice Conclusions
,
In general, radiation and contamination controls for initial U1 RF7 outage activities were appropriate and met applicable TS and 10 CFR Part 20 requirement Several instances of poor radiological work practices identified by the inspectors were promptly addressed by managemen Ei.elosure 2 l
s
.. . _ _ _ _ _ _ _ _ _ _ .
.
. .
R1.2 RP&C Dose Assessments Insnection Scone (83750)
The inspectors reviewed and discussed worker doses attributed to U1 RF7 activitie Specifically, personnel contamination reports (PCRs), i.e. .
skin contaminations exceeding IC00 counts per minute (cpm) per probe area, and positive internal whole body counter (WBC) analysis results were reviewed and discussed with responsible licensee representative Dose assessment-methods and assumptions were reviewed for technical adequacy. Dose assessment results were compared to 10 CFR Part 2 Subpart C. occupational dose limit Results of * hot particle" shallow dose assessments were reviewed against guidance documented in Information Notice 90 48 Enforcement Policy for Hot Particle Exposure and NUREG/CR 5569. Health Physics Position (HPPOS) Data Base. Rev. Observations and Findinas The licensee established a 1R7 outage goal of 30 PCRs. As of September 18, 1997, 10 PCRs were documented for assessment of shallow dose equivalent (SDE) to the skin or extremities. Five of the identified PCRs were attributable to poor radiological practices and failure to follow established procedural guidanc Two of the PCR assessments associated with " hot particle" contamination events documented SDEs exceeding 100 millirem (mreh0. which were to be included in the individuals' permanent dose records in acc;rtiance with licensee administrative policy. The SDE assessments were conducted in accordance with procedure 44019 C. Dose Assessment from Contamin? tion and immersion in Noble Gas. Rev. 10. dated May 14, 1997. Preliuinary calculation of the SDEs for the two events were 1.334 and 21.520 millirem (mrem) and were within 10 CFR Part 20 limits. Licensee assumptions regarding location of rad.ioactive contamination or particles shielding and exposure time. were appropriate. However, the inspectors noted that the radioisotope m!xtures included in the SDE evaluations did not include pure beta-emitting radionuclides such as Carbon (C)-14. Strontium (Sr)-89. Sr 90, and Yttrium (Y)-90. Using ratios of pure beta emitting to selected gamma emitting radionuclides from recent 10 CFR Part 61 dry active waste (DAW) radiochemical analyses, licensee representatives calculated that the major beta-emitting radionuclides added an additional 81 mrem to the maximum calculated SDE hot particle exposure. Following discussions of concerns associated with the potential to overlook pure beta emitting radionuclides in skin dose assessments which approach regulatory Wposure limits, licensee representatives stated that action limits would be established to include pure beta emitting radionuclides in skin dose assessment Enclosure 2 l
_ - - -
_ - _ - _ _ _ _ _ _ _ ___
-
. .
Licensee records of WBC analyses conducted between September 7 and September 19, 1997, documented ten positive whole-body counts, eight identified during routine termination analyses and two attributed directly to investigative analyses associated with potential worker contamination events. All WBC results were less than 10 mrem committed effective dose equivslent (CEDE). Based on industry guidelines adopted by the licensee, none of the results were to be included in the individuals dose record Conclusions Worker SDE and CEDE results from contamination events and work activities during the U1 RF7 outage activities were evaluated properly and were within 10 CFR 20.1201 limit For the initial outage activities, approximately half of the PCRs were attributed to poor radiological work practices by worker Skin dose assessments were to be enhanced by establishing action points for including pure beta-emitting radionuclide RI.3 Primary System Chemistry (84750) Insoection Scope (84750)
Unit 1 primary coolant cleanup and shut-down chemistry initiatives were reviewed and discussed with responsible licensee representative Licensee activities and results were reviewed against applicable T procedural requirements, and industry standards, Observations and Findinal Licensee coolant cleanup initiatives included continued upgrades in sub-micron filtration. For the last six months the reactor coolant sub-micron filtration was reduced from 0.2 to 0.1 micron Licensee
, representatives presented data indicatin contact dose rates on cold leg hot leg.g a continued reduction ofand intermediate lo In addition, significant reductions between 1R6 and 1R7 outages in dose rates measurements associated with steam generator drain line ressurizer, regenerative heat exchanger, and reactor head conoseals
)etween the IR6 and 1R7 outages were note Routine shut-down chemistry was employed during the U1 RF7 outage and no concerns were identified during cool down and initial clea'up. Licensee representatives presented trend data indicating a continueu reduction of
.
radionuclide inventory available for removal from the syste Enclosure 2
, _ , . - - -
-
-
-
_ _ - _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _
'
g .
16 Conclusions The reactor coolant cleanup and U.' RF7 chemistry programs were managed appropriately and contributed to caserved reduced dose rate R2 hP&C Status of Facilities and Eouioment (84ZM).
'
R2.1 Containment Minipurge Exhaust Ventilation Design, Instellation and Testing Insoection Stone -
As a result of the identification of an unmonitored, unplanned release pathway documented in Paragraph R8.2 of this report, the inspectors reviewed the design, installation, and testing of the Unit 1 and Unit 2 containment minipurge exhaust system System design, installation, and testing were compared against details documented in UFSAR SS 1.9.140. 9.4.6.1.3. and Table 3.2.2-1; Vogtle Design Manual. Design Control Number-1506. Rev. 8: P&l Diagram Drawing Number (No.) 1X4DB213 Purification and Cleanup System 1506. Rev. 2 Regulatory Guide (RG) 1.140. Design. Testing, and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water Cooled Nuclear Power Plants. Rev. 1. RG Information Relevant to Maintaining Occupational Radiation Exposure As low As is Reasonably Achievable. Rev. 1. and American Nuclear Standards Institute (ANSI) N509-1980. Nuclear Power Plant Air Cleaning Units and Components._and ANSI N 510 1980. Testing of Nuclear Air-Cleaning Systems, Observations and Findinas From direct walk-downs of the Unit 2 containment minipurge ventilation system and discussions with system engineers, the inspectors verified that the equipment met the design base document specifications and was installed in accordante with the applicable configuration control drawing The UFSAR. S. I.9.140 specifies that, excluding specific sections, the containment minipurge system conforms to RG 1.14 Regulatory Guide 1.140. SS 2.d and 3.f specify that the cleanup systems should be designed to control leakage and facilitate maintenance in accordance with guidelines of RG 8.8 and that ductwork associated with the system should be designed to exhibit on test, a maximum leak rate as defined in ANSI N509 1976. S4.1 From review of UFSAR S 1.9.140_and discussions of system purchase orders, the inspectors noted that the licensee was committed to ANSI N509 1980 guidance in Ifeu of ANSI N509-1976 for the containment minipurge ventilation system. Further. RG 8.0. 5 specifies that ventilation systems will be designed to ensure control of Enclosure 2
- _ _ _ _ _
. _ _ ____ -__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
. .
airborne contaminants. ANSI N509 1980. S4.12 specifies that ventilation system pressure boundary leakage test criteria and methcds are based on the lowest value of: air cleaning effectiveness, health physics requirements, or duct and housing quality requirements, with the minimal value to be used for allowable leakage for design and testing Although initial test data were not available for review prior to the end of the onsite inspection, cognizant licensee representatives stated that the system was designed and tested for air cleaning effectiveness and not health physics consideration Air velocity data and calculated leak rates through the minipurge exhaust duct fan delt openings were reviewed and discussed. Leak rate estimatas ranged from 123 to 900 cubic feet per minute (cfm), depenrient upon the Unit 1 HJ-2632 1 solation damper being in the open or closed position and 60 cfm for the Unit 2 system independent of the associated isolation damper position. With the U2 2HV-2632 isolation damper-closed and the U2 contair, ment minipurge not operating, the inspectors verified leakage into the U2 equipment room as a result of backflow from the U2 main stack. The inspectors noted that, based on observed flow rates into the equiament room and dependent upon radionuclide releases through the main stac( vents, the current design may not meet ALARA considerations of a dressout area. The inspectors also questioned the original test methods and criteria. Further, the startup licensee evaluations and associated test data were not available prior to the end of the inspection to verify that the system pressure boundary was evaluated to determine the appropriate test criteria and methods documented in ANSI N509-1980. S4.12. During the review, the ins)ectors also noted inconsistencies among UFSAR s)ecifications regarding RG 1.140 compliance and Table 3.2.2-1 details. T1e inspectors noted that, pending clarification of UFSAR containment minipurge design and testing commitments, and determination of pressure boundary test criteria and associated results, this-issue would be identified as unresolved item (URI) 50-424. 425/97-09-05: 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 Result The inspectors questioned potential effects of the containment minipurge leakage regarding onsite emergency pre)aredness activities. Althoug licensee representatives stated that tie equipment rooms were inaccessible during selected accident scenarios, the inspectors questioned the impact of the releases during accident condition on adjacent auxiliary building location The ins)ectors identified this issue as IFI 50-424. 425/97-09-06: Review the Effects of Containment Minipurge Leakage on Site Emergency Preparedness Activities during Accident Condition Enclosure 2
_ _ _ _ _ _ _ _ _ _
_ - _ __ _ - _ ___ __ _ _--___ _ _-__-_ ___-__ ___-_
. .
I f
!
18 Conclusions The Unit 2 containment minipurge ventilation equipment met design :
criteria and was installed in conformance with configuration control l document )
Unresolved item 50-424. 425/97-09 05 was identified to review Unit 1 and Unit 2 containment minipurge pressure boundary design criteria and evaluate minipurge pressure boundary test criteria and result Inspector followup item 50 424, 425/97-09-06 was identified to review the effects of containment minipurge leakage on site emergency preparedness activities during acciden '
R RP&C Procedures and Documentation (83750)
R3.1 Dose Records St.one The inspectors reviewed and discussed licensee program guidance and results for determining current-year prior occupational dose and for documenting individual SDE doses from hot particle exposures. The inspectors reviewed and discussed both NRC Form 4. or equivalent, and NRC Form 5 records for selected contractor personnel involved in health physics, scaffolding or insulation containment work during the U1 RF7 outage, Licensee program guidance and corresponding records were compared against 10 CFR 20 Sub) art L requirements. Licensee procedure 45012- Individual Radiation Exposure Records and Reports. Rev. 15. was reviewed'
for adequac Results and Observations Licensee guidance met the intent of 10 CFR 20. Subpart L requirement However. inconsistencies in completing the NRC Form 4 equivalent documentation signed by selected workers to provide initial estimates of previous current-year occupational dose were identified. Previous-annual occupation doses for contractors involved in scaffolding and insulation activities were completed appropriately: however. records of selected contract HP technicians only included the total effective dose equivalcot (TEDE) and not the separate occupational doses specifically listed on NRC Form 4. Licensee representatives stated that for individuals unable to provide accurate estimates of occupational doses other than their TEDE doses equal to the TEDE were assumed for the remaining occupational dose values. Based on discussion with selected HP staff, the inspectors noted that the blank data entries were sometimes interpreted as zero or not recorded. rather than the TEDE value. Licensee representatives stated that, in general, the TEDE dose Enclosure 2 a
.
!' . .
i l
.
'
J :
,- was most limiting and was used to establish an administrative dose i i
margin significantly below regulatory limits. Following discussions of I L the potential for an inconsistent interpretation of the dose record i
! licensee representatives stated that the estimates of the individual i l
occupational dose categories would be documented.
i During review and discussion of hot particle exposures. the inspectors i t
noted that the calculated SDE values from the particle were added to l i skin or extremity SDE doses, as applicable. However, the licensee i
'
records did not indicate the magnitude of dose contribution from the hot .
,.
"
particle exposure. Following discussion of possible future compliance '
concerns when hot particle SDE values are added directly to the SDE - ,
values from normal skin or extremity _ exposure. licensee representatives !'
stated that subsecuent NRC Form 5 documents would provide additional
- information regarcing the SDE values.
i Conclusions- ,
I Program guidance and records for determinin i occupational exposure and recording particl! worker's prior yearlSDEvaluesm: Subpart L requirements.
, ' Inconsistencies in documenting workers' estimated previous annual !
, occupational exposures were identifie ,
,
'
R6 Rad <ation Protection-and Chemistry Oraanization and Administration +
183750. 34750) i R6.1 Health Physics Staffing and Qualifications L .
, a, Scope
.-
l
.
"
Current HP staffing levels to support the U1 RF7 outage activities and routine Unit 2 operations were discussed and evaluated. In additio '
n qualifications-and training provided to HP staff involved in the outage j; activities were; reviewe ,
.-
The-inspectors directly observed HP coverage of outage tasks and
'
- evaluated technician proficiency in conducting selected tasks.
! . Observations and Findinas- :
L For the outage. the licensee's 30 3ermanent HP senior technicians were
-
! supplemented by approximately 35 A9SI 3.1 qualified contractor -
1- 3ersonne The Vogtle Nuclear Plant Chemistry. Training and Decon
'
)epartments provided seven senior and ten junior HP technicians-to sup) ort the outage. In addition. eight senior and three junior HP
'
teclnicians were provided from other Southern Company nuclear plant *
,
i Enclosure 2 b
, . - - - - , , - . , _ , , ~ - - . . . - . , . . - . . _ . - - . . . - ~ _ , . - _ , . - . _ - - , . . - - - . . _ , . - - - - , , - - - - - - ,, -
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ __-_ ____ _ _ _ _______ _
. . .
. 20 The inspectors verified that the licensee had established and provided specf 't outage HP and chemistry training to personnel supporting the
.
U1 RF, outage activities. HP coverage of observed outage activities was adequat Excluding an isolated example of documentation and calculation errors regarding breathing zone air sample analyses, the inspectors noted that HP. chemistry, and counting room technicians'
proficiency was adequate to complete their assigned tasks Follow u the air sampling errors identified that a chemist had completed pro)p er of qualifications for the specific task but had not been included in t1e proper outage refresher training. Additional discussions determined that ambiguous air flow data were provided by field HP technicians collecting the air samples. Licensee representatives stated that actions to improve the coordination of outage training and material presented would be reviewed, Conclusions Licensee HP staffing and aualifications were adequate to support both routine and outage operation R Miscellaneous Radiation Protection and Chemistry Issues (84750, 86750)
R8.1 (Onen) eel 50 424.425/97-02-02: Failure to Meet 49 CFR 173.475 Package Dose Rate Limit The inspector reviewed and discussed status of licensee corrective actions presented during an April 18, 1997 Enforcement Conference, and also detailed in a May 20. 1997 letter from C. K. McCoy, Vice Presiden Vogtle Project. Southern Nuclear Operating Company, which documented
'
corrective actions to a Notice of Violation issued on April, 25, 199 The violation was issued for radioactive material package dose rates
which, subsequent to receipt at a vendor facility exceeded Department of Transportation (DOT) limits specified in 49 CFR 173.475. The subjcct Jackages contained vendor equipment used during refueling operation r urther discussions regarding the corrective actions were detailed in a June 3. 1997. letter from J. P. Jaudon Director. Division of Reactor Safety NRC Region 11. to C. K. McCoy, Vice President. Vogtle Projec Southern Nuclear Operating Company.
'
As of September 19. 1997. revised vendor procedures were received and a) proved for use. However, responsible licensee representatives stated tlat additional review of the procedures was to be conducted by HP staff and specialized training to be used in are-job briefings had not been developed. The inspectors noted that t1is item will remain open pending completion of HP review of applicable vendor procedures and development of specific pre-job briefings for removing and packaging vendor equipment following removal from the sperit fuel poo Enclosure 2
. . .
'
l
'
l
. . ,
l
l 21
'- r R8.2 (Closed) Inspector Followun item (IFI) 50-424. 425/97-07 03- l Unplanned /Unmonitored Release from Ul Equipment Bu1101n The inspectors reviewed Licensee Event Report 1-97-003, associated with identification of unmonitored and unplanned airborne releases from the U1 equipment building on July 28. 1997. The noted release pathway was l identified during release of a Volume Control Tank (VCT) sample for the
'
chemistry laboratory which resulted in a high radiation alarm in the U1 equipment building. At the time of the alarm. tne U1 containment mini- <
purge system was not in service. Licensee evaluations determined that the releases resulted from backflow from the U1 main vent stack into the U1 containment mini-purge system and subsequent release through the opening for the system exhaust fan belt. The evaluation determined that .
I a backdraft and an exhaust damper. located between the main vent stack !
and containment minipurge exhaust fan belt opening did not provide a !
complete seal against air flow into the ventilation system. The room is vented directly to atmosphere, thus resulting in an unplanned and ,
unmonitored release pathway. -The inspectors noted that the failure to "
monitor effluents released to unrestricted and controlled areas to demonstrate compliance with dose limits for individual members of the
'
public was a violation of 10 CFR 20.1302 requirement Licensee corrective actions included dose evaluations for the releases during current and previous years of commercial operations. Based on maximum releases which occurrti during 1995. the licer.see estimated that the unmonitored release pathway could have increased dose estimates to offsite individuals by 0.001 percent of the limits specified in the Off Site Dose Calculation Manual. Currently, the licensee initiated continuous effluent release permits for both Unit 1 and Unit.2 equipment room release points. In addition, operations procedures were changed to maintain the normally-opened 1/2HV-2632B exhaust dampers closed to minimize backflow into the containment mini) urge system from the plant *
vent and resultant effluent releases throug1 the equipment room. The licensee is evaluating long-term corrective actions to seal the
'
containment ) urge system duct exhaust fan belt openings to eliminate all releases. T1e inspectors noted that. consistent with Section Vll.B.1 of the NRC Enforcement Policy this non repetitive licensea-identified and corrected violation is being treated as Non-Cited Violation (NCV)
50-424, 425/97-09-07: Failure to Monitor Effluents Released to Unrestricted and Controlled Areas in Accordance with 10 CFR 20.1302(a)
Requirements.
l
!
t i
-
Enclosure 2
. _ _ _ . _
. _ _ _ _ _ _ _ _.._. _ __
- _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ - _ _ _ _ _ _ _ _ _ _ _ _ _
.
. .
S3 Security and Safeguards Procedures and Documentation S3.1 Material Searches ! Insoettion Scone (71750)
Due to an increased amount of materials and equipment received at the
)lant to support outages activities, the inspectors reviewed the process
)y which these items enter the protected areas (PA) and vital areas (VAs). The inspectors reviewed security procedure 90019 C. "Narehouse Materials Access Controls." Rev. 7. Vogtle specific Physical Security Plan. Amendment 34. dated April 28. 1997, and 10 CFR 73.55(d)(3). Access Rec uirements. in addition, the inspectors discussed the m6terial access anc search process with security managemen b. Observations and Findinas From August 8 through 10. 1997, during routine observations of Unit 1 outage activities, the inspectors observed that materials can enter the PA without being searched if the materials met the " positively controlled" criteria of procedure 90019 C. However, the inspectors also determined during additional review of the process that the licensee permitted materials to enter a VA without a search when moved from a controlled area within the PA. This process of material entry into a VA did not meet the search requirements of the licensee approved Physical Security Plan or procedure 90019-C. Materials and equipment permitted to enter a VA from August 8 through 10. 1997, were ultimately released to the responsible departments or vendor personnel for use in support of outage activitie Upon identification of this issue, the licensee performed a proper search of the subject materials and equipment. No items were identified that would have threatened personnel or plant equipment safet As a result of this event, the licensee has implemented specific training to heighten security personnel adherence to search procedure requirement Conclusions
-
The inspectors concluded that the entry of materials and equipment into a Unit 1 VA without being searched by security personnel is contrary to the requirements of the Physical Security Plan. Amendment 34. and procedure 90019-C. This is identified as Violation (VIO) 50-424.425/97-09 03. Improperly Performed Material Searc Enclosure 2
_
'
I
. . .
I
F3 Fire Protection Procedures and Documentation F3.1 Fire Watch Fire Extincuishers insoections a. Inspection Scone (71750)
Based on the identification of a fire extinguisher with an invalid inspection sticker, the inspectors reviewed the licensee's 3rocess for control, inspection, and issuance of fire extinguishers. T1e inspectors
- reviewed procedure 92020 C. " Control of Ignition Sources," Rev. 11, and UFSAR Section 9.5.1, Fire Protection Program, in addition, the 1 inspectors interviewed involved personnel and licensee fire protection supervisio b. Observations and Findinas
'
On August 15 during routine observations of maintenance, the inspectors identified that a fire extinguisher, being used in support of a work i activity had exceeded the valid ins)ection date. The fire extinguisher inspection date was July 1, 1997. Jpon identification, the licensee immediately removed the fire extinguisher from the work area and obtained a valid replacement. Followup inspection identified that all fire extinguishers located in the maintenance shop tool room on August 15 were dated July 1, 1997. A review of the burn permit log, used to track fire extinguishers issued for hot work activities, indicated that r 54 fire extinguishers issued between July 30 and August 15, 1997 had invalid inspection sticker Procedure 92020-C provides instructions that fire extinguishers must be inspected within 28 days of the date printed on the inspection sticker on the extinguisher. The fire protection team is responsible for those inspections. However, due to a 3rocedure revision in March 1996 the process by which those extinguislers were inspected changed. The revision had qualified fire protection personnel perform the required inspections at the maintenance shop after the extinguisher had been issued to responsible fire watch personnel. The new arocess had the responsible individual ensure that the fire extinguisler issued had a valid inspection sticker. This task would be accom the valid inspection date on the burn permit form.The plished by filling in
-
inspectors *
review of numerous burn permit forms indicated that the verification of valid inspections was not being performed by fire watch personne Previously, the precess had fire protection personnel retrieve the extinguishers from the field, inspect them, and then placed them in the tool room to be issue To correct this 3roblem, the licensee: increased training to fire watch personnel to enplasize the extinguisher inspection date requirement:
qualified mainteunce shop tool room personnel responsible for issuance of fire extinguist.ers to inspect extinguishers: and trained fire technicians to verify that fire watch personnel document burn permit Enclosure 2
- - .
- - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
. . .
information properly, in addition, the licensee plans to revise procedure 92020 C-to have fire watch personnel qualified to perform fire extinguisher inspections, Conclusions The inspectors concluded that the failure to inspect fire extinguishers being used to su) port plant activities beyond the valid inspection date on the extinguisler is contrary to procedure 92020 C requirements. This is identified as Violation (VIO) 50-424, 425/97 09-04, Invalid Fire Extinguisher Inspection Manaaement Meetinas and Other Areas X Review of Updated Final Safety Analysis Report (UFSAR)
A recent discovery of a licensee o)erating its facility in a manner contrary to the UFSAR description lighlighted 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 re) ort, 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 except as noted in this repor X1 Exit Meeting Summary The inspectors ) resented the inspection results to members of licensee management at tie conclusion of the inspection on September 23, 199 The licensee acknowledged the findings presente The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identifie PARTIAL LIST OF PERSONS CONTACTED Licensee J. Beasley, Nuclear Plant General Manager J. Gasser, Plant 0)erations Assistant General Manager M, Griffis, Plant iaintenance and Modifications S. Chestnut. Manager Operations K. Holmes, Manager Maintenance B. Burmeister, Manager Engineering Technical Support D. Huyck, Manager Security M. Sheibani, Nuclear Safety and Compliance Supervisor C. Tippins, Jr., Nuclear Specialist 1 Enclosure 2
. _ - - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _
> . .
INSPECTION PROCEDURES USED IP 37551: Onsite Engineering IP 40500: Effectiveness of Licensee Controls ii. Identifying, Resolving, and Preventing Problems IP 60710: Refueling Activities IP 61726: Surveillance Observation IP 62700: Maintenance Implementation IP 62707: Maintenance Observation IP 71707: Plant Operations IP 71750: Plant Support Activities IP 83/50: Occupational Radiation Exposure IP 84750: Radioactive Waste Treatment, and Effluent and Environmental Monitoring IP 86750: Solid Radioactive Waste Management and Transportation of Radioactive Materials ITEMS OPENED.-CLOSED. and DISCUSSED Ooened 50-425/97-09 01 VIO Failure to Obtain a Work Order Prior to Conducting Maintenance on a Diesel Generato (Section M1.3)
50-424/97-09-02 NCV Procedure Revision Results in Auxiliary Feedwater Actuation (Section M8,1)
50 424. 425/97-09-03 VIO Improperly Performed Material Search (Section S3.1)
50-424, 425/97-09 04 VIO Invalid Fire Extinguisher Inspections (Section F3.1)
50-424, 425/97-09-06 URI 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 (Paragraph R2.1).
50-424, 425/97-09-06 IFI Review the Effects of Containment Minipurge E
Leakage on Site Emergency Preparedness Activities during Accident Conditions '
(Paragraph R2.1).
Enclosure 2
'
. _ _ - _ - _ _ _ _ _ _ _ _ - - _ - _ _ _ _
_
,
. . .
50 424, 425/97-09 07 NCV Failure to Monitor Effluents Released to Unrestricted and Controlled Areas in Accordance with 10 CFR 20.1302(a) Requirements -
(Paragraph R8.2).
Closed 50 424/97 05 00 LER Test Procedure Inadequacy Results in Auxiliary Feedwater Actuation (Section M8.1)
50-424/97 09 02 NCV Procedure Revision Results in Auxiliary Feedwater Actuation (Section M8.1)
50-424, 425/97-07-03 IFI Unplanned /Unmonitored Release from Ul Equipment Building (Paragraph R8.2).
50-424, 425/97-09 07 NCV Failure to Monitor Effluents Released to Unrestricted and Controlled Areas in Accordance with 10 CFR 20.1302(a) Requirements .
(Paragraph R8.2).
Discussed 50-424, 425/97 02-02 eel Failure to Meet 49 CFR 173.475 Package Dose Rate Limits (Paragraph R8.1).
.
Enclosure 2
. ..
. -