IR 05000424/1987044
| ML20237K747 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 08/21/1987 |
| From: | Rogge J, Schepens R, Sinkule M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20237K705 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.G.1, TASK-2.E.1.1, TASK-2.F.2, TASK-2.K.3.05, TASK-TM 50-424-87-44, 50-425-87-30, GL-87-09, GL-87-9, NUDOCS 8708270283 | |
| Download: ML20237K747 (16) | |
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'inMog UNITED STATES -
o NUCLEAR REGULATORY COMMISSION
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' REGION il
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y g-j 101 MARIETTA STREET,N.W.
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ATLANTA, GEORGI A 30323 Y
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Report Nos.:
50-424/87-44 and'50-425/87-30 Licensee:
Georgia Power Company P. O. Box 4545 Atlanta,'GA '30302 Docket Nos.:
50-424 and 50-425 License Nos.:
NPF-68 and CPPR-109 i2 '
Facility Name:
Vogtle 1 and 2 Inspection Conducted:
June 20 - July 24, 198i
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Inspectors:
2/[87
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J. F. Rogge, Senior Resident Inspector, Operations Date Signed ksf h.
NhW)t%0 8lAl87
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'R. Qcfiepens,6 Resident Inspector, Operations Date Signed Accompanying Personnel:
C. W Burge
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s/[6 M 77 l
Approved by:
t M. V. Sinkuls,'Sectiorf Chief Ddte SWned'
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Division of Reactor Projects
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SUMMARY Scope:
This routine, unannounced inspection entailed Resident Inspection in the following areas:
plant operations, radiological controls,-maintenance,
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surveillance, fire protection, emergency preparedness, security, preoperational i
test program, quality programs and administrative controls affecting quality, and follow-up on previous inspection identified item.
Results:
One violation was identified in the quality programs area-failure to properly utilize a temporary modification procedure.
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1.
Persons Contacted Licensee Employees P. D. Rice, Vice-President, Vogtle Project Director
- G. Bockhold, Jr., General Manager Nuclear Operations E. M. Dannemiller, Technical Assistant to General Manager T. V. Greene, Plant Manager C. W. Hayes, Vogtle Quality Assurance Manager
- C. E. Belflower, Quality Assurance Site Manager - Operations E. D. Groover, Quality Assurance Site Manager - Construction
- W. E. Mundy, Quality Assurance Audit Supervisor G. A. McCarley, Project Compliance Coordinator S. F. Goff, Regulatory Specialist i
- W. C. Gabbard, Regulatory Specialist C. F. Meyer, Operations Superintendent
- R. M. Odom, Plant Engineering Supervisor C. L. Coursey, Maintenance Superintendent (Startup)
- M. A. Griffis, Maintenance Superintendent
- G. R. Frederick, Quality Assurance Engineer / Support Supervisor R. E. Spinnatu, ISEG Supervisor J. F. D'Amico, Nuclear Safety & Compliance Manager W. F. Kitchens, Manager Operations
- A. L. Mosbaugh, Assistant Plant Support Manager j
M. P. Craven, Nuclear Security Manager j
- J. E. Swartzwelder, Deputy Manager - Operations
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H. M. Handfinger, Assistant Plant Support Manager l
I Other licensee employees contacted included craftsmen, technicians, j
supervision, engineers, operations, maintenance, chemistry, inspectors, i
and office personnel, i
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Other NRC Personnel l
B. H. Little, Senior Resident Inspector-Callaway
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on July 24, 1987, with
those persons indicated in paragraph 1 above.
The inspector described the areas inspected and discussed in detail the inspection results.
No i
dissenting comments were received from the licensee.
The licensee did not i
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identify as proprietary any of the materials provided to or reviewed by the inspector during this inspection.
Region based NRC exit interviews were attended during the inspection period by a resident inspector.
This inspection closed four Inspector Followup Items (IFI), one Unresolved Item, and four Three Mile Island Task Followup Items.
The items identified during this inspection are:
a.
Violation 50-424/87-44-01, " Failure to properly implement a temporary modification to the train "A" electric tunnel ventilation system" -
paragraph 5.b.(1).
b.
Unresolved Item 50-424/87-44-02,
" Review licensee's findings regarding missed Technical Specification surveillance not reperti#
J paragraph 4.
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Inspector Followup Item 50-424/87-44-03, " Review licensee procedure and FSAR changes regarding the fire protection program"
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paragraph 5.b(3).
The following previous inspection item remains open due to incomplete licensee action:
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50-424/86-111-02 " Review PRB Procedure for Proper Incorporation of Technical Specifications" - Paragraph 7.c.
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Licensee Action on Previous Enforcement Matters Not inspected.
4.
Unresolved Items - Units 1 & 2 (92701)
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Unresolved items are matters about which more information is required to determine whether it is acceptable or may involve a violation or deviations.
The new unresolved item identified during this inspection is discussed in paragraph 5.b.(6).
5.
Operational Safety Verification - Unit 1 (71707) (93702)
The plant began this inspection period in Hot Standby (Mode 3) conducting an outage to repair feedwater system problems.
On June 21 the unit
. returned to Power Operation (Mode 1) until the reactor tripped on a turbine trip from 100% power due to high generator field amperage sensed by an out of calibration sensor.
The unit again returned to Mode 1 on June 25.
The unit remained at 100% power level until July 9 when the reactor tripped on a turbine trip due to a problem with the stator cooling system.
The unit was not able to return to Startup (Mode 2) until July 12 due' to design modification work on the Control Room Emergency Ventilation
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System.
On July 13 Mode I was again achieved, but the unit remained at l
low power level until July 15 while work on Steam Generator Water level I
control problems associated with the #4 Steam Generator was accomplished.
On July 16 the unit achieved 100% power and remained at this level until
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July 22 when the unit again tripped due to a problem with the stator
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Due to the failure to identify the root cause on the first trip the licensee had installed additional monitoring and this effort allowed the licensee to locate a faulty temperature switch as the root cause.
a.
Control Room Activities Control Room tours and observations were performed to verify that facility operations were being safely conducted within regulatory requirements.
These inspections consisted of one or more of he following attributes as appropriate at the time of the inspection.
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Proper Control Room staffing
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Control Room access and operator behavior
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Adherence to Technical specification (TS) Limiting Conditions i
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for Operations (LCO)
Observance of instruments and recorder traces of safety related
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and important to safety systems for abnormalities
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Review of annunciators alarmed and action in progress to correct Control Board walkdowns
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Safety parameter display and the plant safety monitoring system
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operability status Discussions and interviews with the On-Shift Operations l
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Supervisor, Shift Supervisor, Reactor Operators, and the Shift
Technical Advisor to determine the plant status, plans and assess operator knowledge Review of the operator logs, unit log and shift turnover sheets
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No violations or deviations were identified.
b.
Facility Activities I
Facility tours and observations were performed to assess the j
effectiveness of the administrative controls established by direct j
observation of plant activities, interviews and discussions with licensee personnel, independent verification of safety systems status and LCO's licensee meetings and facility records.
During these j
inspections the following objectives are achieved:
(1) Safety System Status (71710)
Confirmation of system
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operability was obtained by verification that flowpath valve alignment, control and power supply alignments, component conditions, and support systems for the accessible portions of the ESF trains were proper.
The inaccessible portions are confirmed as availability permits.
Additional indepth inspection of the train A diesel generator system was performed to review the system lineup procedure with the plant drawings l
and as-built configurations, compare valve remote and local indications, walkdowns were expanded to include hangers and supports, and electrical equipment interiors.
The inspector
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verified that the lineup was in accordance with licensee
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requirements for system operability.
On July 20, 1987, the Resident Inspector followed up on the j
reason for the illuminated a' arm " Train A Electrical Tunnel
Ventilation System" on the safety system monitoring panel I
(SSMP).
Discussions with the operators revealed the cause for J
the alarm to be the turbine buile;ing (TB) and auxiliary building
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(AB) Train A tunnel supply ventilation fan (Equipment Tag i
No.1-1540-B7-005) which was taken out of service on April 8, i
1987, via clearance No. 1-87-1070 for an indefinite period of time.
This fan is part of the electrical tunnel ventilation i
system which is described in section 9.4.9.2 of the FSAR.
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safety function is to prevent excessive temperature rise by
providing minimum ventilation for the TB & AB Train A cable f
tunnel during normal, shutdown, refueling and accident q
conditions.
The fan motor is designed to start automatically by the Thermostats when the tunnel temperature reaches 90 F.
The fan may also be activated manually by a remote hand switch located on the Q-HVAC panel in the control room.
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The inspector reviewed clearance No. 1-87-1070 which was hung for personnel safety since that with the fan running extremely high pressures were reported behind the doors on Level II of the AB going into the electrical tunnel.
Also a review was conducted of request for er.gineering review (RER) No. 1-87-0557 which identified that the design air flow cannot be achieved due to a sealed wall containing a closed fire door between the Auxiliary and Fuel Handling Building not allowing air flow past this door so as to be exhausted through the turbine building.
The engineering approved disposition was to create an air flow path and that ventilation was not required as long as temperatures are maintained below 100 F.
Also, temporary ventilation should be provided to prevent temperatures from exceeding 100 F until a permanent resolution is implemented.
The inspector raised the following concerns regarding the subject item:
(1) no deficiency card was found to document this deficient condition as required per Administrative Plant Procedure 00150-C " Deficiency Control," (2) operations personnel were aware that the subject fan was out of service but had not questioned how the subject fan might have impacted the operability of the ESF systems indirectly, (3) a clearance was hung which effectively implemented a temporary plant modification to a safety related system described in the FSAR
preventing it from performing its intended safety function i
without initiating the appropriate plant administrative
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procedure 00307-C " Temporary Modifications" which would have
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required that a safety evaluation be performed per Plant
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Administrative Procedure 00056-C " Safety and Environmental
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Evaluations" to justi fy continued plant operation in this I
condition.
Upon informing the licensee of the above concerns immediate action was taken by the licensee to (1) initiate a deficiency card, (2) review it to determine that it does not indirectly affect the operability of an ESF system, and (3) evaluate the tunnel temperatures to ensure that they will not be exceeded during accident conditions and to adequately document the engineering basis via a safety evaluation to justify continued operation in this condition.
Due to the above noted deficiencies and the fact that the RER did not adequately document the basis of an engineering review for accident conditions nor were the appropriate plant procedures implemented prior to effectively implementing a temporary plant modification to the Train "A" electrical tunnel ventilation system.
The foregoing is considered to be in violation of Technical Specification 6.7.la and will be identified as " failure to properly implement a temporary modification to the Train
"A" electrical tunnel ventilation system."
As part of the inspectors review of the above matter, the inspector reviewed operations procedure No. 10005-C " Operability Status Indication for Plant Safety Systems." The purpose of this procedure is to provide instructions for indicating and recording the status of operable and inoperable plant safety systems. The inspector identified the following concerns to the licensee:
(1) procedure 10005-C was very general in nature and did not require any documenting of operator actions regarding its use.
Since this system is very complex in that it has numerous inputs and can lead the operator to identifying systems which may be inoperable as a result of being either directly or indirectly impacted it would be beneficial to require some form of documentation of the action taken by the operator in response, (2) Table I was not found to be up to date in that it did not have the electrical tunnel ventilation system listed nor what actions to take should it be found in the alarmed condition.
Other system listed in the Table, e.g.,
diesel generator fuel oil and air start system no longer existed on the panel, but rather had another system designated by and information tug. which was not found in Table 1 (3) the operators were basically familiar with the purpose, were the procedure was located and how to use the SSMP procedure.
However, their knowledge on how to implement the procedure was somewhat vague due to the procedure being very general in nature.
The inspector suggested that the licensee consider implementing a specialized training program, in addition to the
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existing training being given to the operators in the simulator, as how to properly use the SSMP through scenarios along with a more detailed procedure. The licensee indicated that they would review these concerns of the inspector's and would implement procedure changes deemed appropriate.
Storage of material and j
(2)
Plant Housekeeping Conditions
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components and cleanliness conditions of various areas throughout the facility were observed to determine whether safety and/or fire hazards existed.
(3)
Fire Protection - Fire protection activities, staffing and
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I equipment were observed to verify that for brigade staffing was appropriate and that fire alarms, extinguishing equipment, actuating controls, fire fighting equipment, emergency equipment, and fire barriers were operable.
During an visit by the Senior Resident Inspector from the Callaway Station on July 1,1987, he noted that all three fire pumps were in an inopcrable status with a seven day allowance to be restored.
He also noted that the fire protection engineer appeared to be acting independently of the Shift Supervisor in regard to fire protection alarm response.
These issues were presented to the General Manager.
The inspection followup determined that the third fire pump was restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
In addition the licensee suspended operations order No.1-87-29 which routinely placed two fire pumps in an inoperable status to prevent spurious starts during Unit 2 fire loop flushing evolutions.
Discussions were held with the Fire Protection Engineer regarding the use of fire protection LCO's and the performance of surveillance. The exect nature of procedure changes and FSAR connections is under review.
To follow the licensee's actions the following is identified, Inspector Followup Item 50-424/87-44-03, " Review Licensee Procedure and FSAR changes regarding the fire protection program."
(4)
Radiation Protection (71709) - Radiation protection activities, staffing and equipment were observed to verify proper program implementation.
The inspection included review of the plant program effectiveness.
Radiation work permits and personnel compliance were reviewed during the daily plant tours.
Radiation Control Areas (RCAs) were observed to verify proper identification and implementation.
(5) Security (71881) - Security controls were observed to verify that security barriers were intact, guard forces were on duty, and access to the Protected Area (PA) was controlled in accordance with the facility security plan.
Personnel within the PA were observed to verify proper display of badges and that
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Personnel within vital areas were observed to ensure proper authorization for the area.
Equipment operability of proper compensatory
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activities'were verified on a periodic basis.
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(6) Surveillance (61726) (61700) - Surveillance tests were observed I
to rerify that approved procedures were being used; qualified f
personnel were conducting the tests; tests were adequate to
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verify equipment operability;' calibrated equipment was utilized,
and TS requirements were followed.
The inspectors observed
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portions of the following surveillance and reviewed completed i
data against acceptance criteria:
i Surveillance No.
Department Title 14701-101 Operations Train "A" Reactor Trip Breakers and Shunt Trip Test 24565-102 I&C RCP-2 Train A Reactor Trip Under Frequency and Undervoltaga Relays Trip Actuating Device Operational Test At the request of Region II and the Office of the Analysis and Evaluation of Operational Data, the resident inspectors conducted a review of the licensee's reporting criteria for missed surveillance.
The inspector held discussions with licensee personnel within the Nuclear Safety and Compliance (NSAC) Organization at the site.
It was determined that up to July 24, 1987, it it was the licensee's position to only report missed Technical Specification Surveillance which resulted in a violation of a Technical Specification action statement for a limiting condition of operation or one which was safety significant.
Furthermore, NSAC had identified that their position was contrary to a recently issued NRC staff official position contained in Generic Letter 87-09, dated June 4,1987.
The NRC staff stated that missed Technical Specification surveillance are reportable pursuant to 10 CFR 50.73(a)
(2)(1)(b).
On July 24, 1987, NSAC issued a letter to all site organizations informing them that effective immediately all missed Technical Specif ication surveillance are reportable pursuant to 10 CFR 50.73(a)(2)(i)(b) per the NRC staff position stated in Generic Letter 87-09.
The licensee is conducting a review to identify all previous missed surveillance for deportability in accordance with their newly established reporting criteria.
Pending the results of the licensee's review of all missed Technical Specification surveillance for deportability this item will remain unresolved and be identified as unresolved item 50-424/87-44-02, " Review licensee's findings
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regarding missed Technical Specification surveillance not reported."
(7) Maintenance Activities (62703)
The inspector observed
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maintenance activities to verify that correct equipment clearances were in effect; work requests and fire prevention work permits, as required,- were issued and being followed; j
quality control personnel were available for inspection activities as required; retesting and return of systems to service was prompt and correct; TS requirements were being
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followed.
Maintenance backlog was reviewed.
Maintenance was observed and work packages were reviewed for the following maintenance activities:
MWO'No.
Department Description 1-87-07565 Mechanical Implement OCP No. 87-V1E0245 to Install Backdraft Dampers in the Supply and Outside Air Ductwork - Train B CB CR HVAC System 1-87-06944 I&C NSCW Train "A" Tower -
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Investigate, Rework, and Recalibrates Make UP Level Control Valve ILV-1601 (8) Emergency Preparedness - The inspectors participated in the licensee's drill conducted on July 15, 1987.
Positions manned were in the Control Room (Simulator) and Technical Support Center.
Inspector observations were provided to the evaluators.
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6.
Review of Licensee Event Reports - Unit 1 (90712) (92700)
Licensee Event Reports (LER's) and Deficiency Cards (DC's) were reviewed for potential generic impact, to detect trends, and to determine whether corrective actions appeared appropriate.
Events which were reported immediately, were reviewed as they occurred to determine if the technical specifications were satisfied.
a.
The below listed LER's were reviewed to determine if the information
provided met NRC requirements.
The determination included:
Adequacy l
of Description, Verification of Compliance with Technical i
l Specifications and Regulatory Requirements, Corrective Action Taken,
Existence of Potential Generic Problems, Reporting Requirements l
l Satisfied, and the Relative Safety Significance of each event.
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Additional implant reviews and discussions with plant personnel, as
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appropriate were conducted at the time of the event for those reports indicated by an (*).
The following LER's were reviewed and are ready for closure pending verifications of satisfactory completion of the licensee's corrective action:
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424/87-001, Rev. O Incorrect Transmitter Circuit Board Leads to Missing a Required Flow Rate Estimation
- 424/87-002, Rev. O Potential failure of MSIV's to Close
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Following Small Steam Line Break
- 424/87-003, Rev. O Restriction of Pipe Movement With Incorrect Penetration Sealant Material
- 4E4/87-004, Rev. O Containment Isolation Activations Cause By Faulty Circuit Board
- 424/87-006, Rev. 0 ESF Actuation of Auxiliary Feedwater Due to Inadvertent Trip of Main Feed Pumps
$424/87-007, Rev. 0 ESF Actuations caused by Steam Generator Water Level 424/87-009, Rev. 0 ESF Actuations caused by Adjustments to Steam Generator Level Control Systems.
424/87-87-010, Rev. O RPS Actuations Caused by Adjustments to Steam Generator Level Control Systems 424/87-011, Rev. O Reactor Trip Due to Lo-Lo Steam Generator Level 424/87-012, Rev. O Reactor Trip Due to Feedwater Control Problems Following Generator / Turbine Trip 424/87-013, Rev. O Feedwater System Valve Malfunctions Result
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in Reactor Trip i
424/87-014, Rev. O Steam Generator High Level Results in Reactor Trip 424/87-015, Rev. O Inadvertent Steam Dump Oper<. tion Results in ESF Actuation b.
The following LER's were reviewed and are considered to be closed based on a satisfactory review by the inspector which included verification that the licensee's identified corrective action was complete.
- 424/87-008, Rev. O Reactor Trip on High Positive Flux Rate 424/87-016, Rev. 0 Fuel Handling Building Isolation Due to High radiation Signal i
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424/87-017, Rev. O Control Room Ventilation Isolation
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424/87-026, Rev. O AFW Actuation Occurs Due to Procedure I
Inadequacy After Switching Feedwater Flow.
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424/87-029, Rev. O Reactor Trip on SG Lo-Lo Water Level Due to
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Misaligned Valve.
LER 424/87-017 Described an incident where an instrument and control (I&C) technician, while trouble shooting radiation monitor 1RE-12117,
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performed a source check on radiation monitor 1RE-12116 without i
following appropriate plant procedures.
The source check on i
1RE-12116 was performed without the test signal being blocked.
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resulted in radiation monitor 1RE-12116 sensing a high radiation signal from the source check which initiated a control room ventilation isolation.
The failure to follow the appropriate plant procedures indicates an apparent violation of Technical Specification 6.7.la.
This requires that written procedures be established, implemented and maintained covering activities delineated in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978.
This a3 parent violation was reviewed and since all the requirements specified in 10 CFR Part 2, Appendix C, Section V, were satisfied, this violation is not cited.
Therefore, this item will be identified and tracked as licensee identified violation 50-424/ LIV 87-01 "LER 87-17 - failure to follow appropriate plant procedures".
LER 424/87-24 described an event where an operator assigned to isolate SG #3 main feedwater regulating valve for a maintenance check did not isolate it with the equivalent valves used to isolate other SG's for the same maintenance.
Upon completion, the plant equipment operator assigned to restore the manipulated valves to their normal position verified that valves equivalent to those used for the same maintenance on the other loops were restored.
This resoluted in valve 1-1305-U4-004 being left closed and subsequently prevented i
normal feedwater to SG #3 when the bypass feedwater isolation valve
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was opened during the feedwater flow transfer process.
The incorrect j
closure and failure to restore manual valve 1-1305-U4-004 in the
SG #3 bypass feedwater regulating valve piping eventually resulted in I
a SG #3 lo-lo level reactor trip.
The failure to restore valve 1-1305-U4-004 to its normal position following maintenance due to personnel error and inadequate work instruction indicates an apparent violation of Technical Specification 6.7.la.
This required that written procedures be established, implemented and maintained covering activities delineated in appendix A of Regulatory Guide 1.33, Revision 2, February 1978.
This apparent violation was reviewed and since all the requirements specified in 10 CFR Part 2, Appendix C, Section V, were satisfied, the violation is not cited.
Therefore, this item will be identified and tracked as licensee
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identified violation 50-424/ LIV 87-02 "LER 87-29 - Failure to Properly Restore Valve 1-1305-04-004 Following Maintenance".
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7.
Followup on Previous Inspection Items - Units 1 & 2 (92701)
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(0 pen) Inspector Followup Item 50-424/86-111-02.
" Review PRB Procedure for Proper Incorporation of Technical Specifications".
Procedure 00002-C, Revision 5 dated March 2, 1987 " Plant Review Board
- Duties and Responsibilities" was compared to the Technical Specification and FSAR.
While the licensee has reconciled the procedure with the Technical Specification one discrepancy remained regarding TS 6.4.1.6.h.
This TS requires " Review of plant operations to detect potential hazards to nuclear safety" while the procedure requires " review of plant operations to detect nuclear safety hazards".
When verifying consistency with the FSAR the inspector noted that the commitment becomes " review evaluations of plant operations to detect potential nuclear safety hazards".
Further review of the FSAR indicates that responsibility F needs revision to be consistent with TS and two new items in TS (items M. and N.)
should be added to the FSAR.
Procedure 0002-C was also noted during this review did not specify that the Chairman was to be either the Nuclear safety and compliance representative or the operations representative as stated in the FSAR and the procedure needs clarification of the term " member".
This item remains open to track I
the licensee resolution.
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(Closed) Inspector Followup Item 50-424/86-111-03 " Review Revision 2 00005-C Regarding Overtime Minimizing".
Procedure 00005-C, Revision 2 dated January 19, 1987, was reviewed and the licensee corrective action was verified.
c.
(Closed)
Inspector Followup Item 50-425/85-13-01
" Review Implementation of FEC0 N-17BF to Install 2 piece Balance Drum Locknuts into the Charging Pumps".
The inspector observed work in progress for one pump and the documentation for both pumps.
d.
(Closed) Inspector Followup Item 50-424/87-12-04 " Review Completed Walkdown Verification Regarding Communication".
The inspection reviewed the documentation of the completed inspection and noted that it was performed as satisfactory.
e.
(Closed) Inspector Followup Item 50-424/87-32-07 " Complete Training for All Construction Personnel".
This was closed NRC Report 50-425/87-31 and this action document Unit 1 item closure since the item concerns Unit 2 personnel.
f.
(Closed) Unresolved Item 50-424/86-18-06 " Live Load Reduction Use of 25% Dive Load in Seismic Calculation for Structural Steel".
This item was resolved by NRR as documented in NRC letter dated September 30, 1986 and is closed here for administrative purposes.
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(Closed) IE Bulletin 50-424 and 50-425/86-80-02, " Static "0" Ring Differential Pressure Switches".
This item was closed in NRC Report 50-424/87-17 for Unit 1.
Baseu on regional review this closure is applicable for Unit 2 and is closed in this report for administrative purposes.
8.
Preoperational Test Program Implementation / Verification - Unit 2 (70302)
(71302)
i The inspector reviewed the present implementation of the preoperational test program.
Test program attributes inspected included review of administrative requirements, document control. documentation of major test events and deviations to procedures, operating practices, instrumentation calibrations, and correction of problems revealed by testing.
Periodic facility tours were made to assess equipment and plant conditions, maintenance and preoperational activities in progress.
Schedules for program completion and progress reports were routinely monitored.
Discussions were held with responsible personnel, as they were available, to determine their knowledge of the preoperational program.
'The Inspector reviewed numerous operation deviation reports to determine if requirements were met in the areas of documentation, action to resolve, justification, corrective action and approval.
Specific inspections conducted are listed below:
a.
Freoperational Tests Test Witnessing (70312)
The inspector witnessed selected portions of the following preoperational test procedures as they were conducted.
The inspection included attendance at briefings held by test supervisors to observe the coordination and general knowledge of the procedure with the test participants.
Overall crew per/ormance was evaluated during testing.
A preliminary review of the test results 'was compared to the inspector's own observations.
Problems encountered during performance of the test were verified to be adequately documented, evaluated and dispositioned on a selected basis.
frocedure NRC Insp.
Test Title Activity Observed No.
No.
2-3PE-01 70441 Class IE - Standby SafetyInjection Power System Train A This is the first preoperational test to be conducted for Unit 2 and the test duration is expected to last 25 days due to set up time.
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b.
Flushing Activities The inspector witnessed portions of the following flushing activities to verify cleanliness of fluid systems and associated components.
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Procedure No.
Description 2-188-05 Pressurizer Relief Tank Flush Flow Path No. 5.21.
2-18G-09 Charging Pump Discharge to Regen.
HX Flush.
Flow Path No. 9.5.
'2-1BG-07 Chemical and Volume Control System Volume Control Tank and Piping System Flush.
Flow Path No. 7.10.
No violations or deviations were identified.
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9.
Three Mile Island Task Action Plan Followup - Units 1 & 2 (4252018)
l This inspection consists of verification that the licensee has implemented i
the requirements of NUREG 0737, " Clarification of TMI Action Plan Requirements" as committed u in the facility FSAR or other appropriate documents.
Verification consisted of one or more of the following attributes, as appropriate, to determine acceptability for each listed action item:
Program or procedure established
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Personnel training or qualification
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Completion of item
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Installation of equipment
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Drawings reflect the as-built configuration
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Component tested and in service or integrated into the preoperational
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test program The following documents were utilized in performing the review, as appropriate:
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NUREG 0578 THI-2 Lessons Learned Task Force Status report NUREG 0660 NRC Action Plan Developed as a Result of the THI-2 Accident i
NUREG 0694 TMI-Related Requirements for New Operating Licenses
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NUREG 0737 Clarification of TMI Action Plan Requirements t
Supplement 1 I
FSAR and Final Safety Analysis Report l
Amendments t
NUREG 1137 and Safety Evaluation Report Supplements
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a.
(Closed) I.G.I.." Training During Low Power Testing".
This requirement is to provide training during the testing phase.
FSAR 14. 2. 5.
describes that certain operators would be designated to conduct the test and others would observe the conduct of the test as schedules permitted.
In SER - Supplement 5 the staff concerned with the performance of the test after achieving 100% power.
The staff determined that plant operators had received adequate training in natural circulation procedures on the simulator.
In addition during the performance of the Natural Circulation Test the inspector l
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witnessed the test and observed training.
Once natural circulation had been achieved thirty minutes was allowed for the operators to be familiar with the real plant response.
The total number of operators
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which experienced the training was 21.
This item is closed for l
Unit 1 only.
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b.
(Closed) II E.1.1 " Auxiliary Feedwater System Evaluation".
This item
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requires the applicant to p' Generic Evaluation of Feedwater Transientserform modifications.
NUREG-611 and Small Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plants" provides the NRC generic recommendations.
The SER paragraph 10.4.9 documents the NRC staff review.
NRC. Region to inspection of the onsite implementation of this item was conducted in
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' conjunction with TMI item II.E.1.2, however the review was-not documented in an inspection report.
This inspection reviewed' the FSAR and SER to determine hardware or procedure modifications which are applicable regarding NUREG-611.
From this review the inspector determined that since few of the recommendations are applicable due to the Vogtle design the inspector verified that for recommendation GS-4 that the site maintains the two series in lat valves to the pumps are in a locked open status and switchover to the second CST is in a controlled procedure and for additional generic recommendation
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No. 2.that 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> endurance tests were conducted by satisfactory completion of reoperation test 1-3AL-03.
This item is closed for Unit 1 only.
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c.
(Closed) II.F.2 " Instrumentation for the Detection of Inadequate Core i
Cooling".
This item requires the design and implementation instrumentation to provide an unambiguous, easy-to-interpret f
indication of inadequate core cooling.
The installed system consists
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of three subsystems.
These subsystems are subcooling margin monitors, core exit thermocouple, and a reactor vessel level instrumentation systems (RVLIS).
This inspection included a review of the FSAR, SER and recent letters dated May 29 and July 20, 1987, i
submitted pursuant to License Condition 2.C(7)a.
Previous NRC inspection which reviewed procedures and hardware were conducted as part of NRC Report 50-424/86-117.
Licensee Action at the time of licensing was to complete testing of RVLIS prior to 100% power.
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To verify that testing and surveillance were complete the following documentation was examined.
Procedure Title
1-5RP-01 RVLIS Final Calibration and Operational Checkout (Including NSID-EIS-87-10)
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14228-1 Operations Monthly Surveillance Logs J
24620-1 RCS Subcooling Margin Monitor Calibration 24677-1 RVLIS Transmitter Calibration 24690-1 Core Exit Thermocouple Calibration The inspector performed a walkthrough of the PSMS displays and discussed the operator use of the displays.
This discussion included the use of the ERF computer and E0Ps.
This item is closed for Unit 1 only.
d.
(Closed) II.K.3.5, " Automatic Trip of Reactor Coolant Pumps During Loss-of-Coolant Accident".
This item concerns a modification to provide automatic tripping of the Reactor Coolant Pumps.
The NRC evaluation of the licensee's proposal was documented in Supplement 3 of NUREG 1137.
The NRC evaluation concludes that no modification are required, and that appropriate reactor coolant pump trip criteria has been established.
Emergency Procedure 19000-1 was verified to ensure the trip criteria was included.
This item was considered closed for both units based on no required modifications are necessary.
10.
Management Meetings - Unit 1 (30702)
The resident inspectors attended a management meeting at Vogtle Electric Generating Plant with the NRC staff and the Licensee on June 30, 1987, to discuss Vogt1c's initial operating history.
The resident inspector attended an NRC Enforcement Conference on operational matters held with the licensee at Vogtle Electric Generting Plant on July 1, 1987.
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