ML20207K554
| ML20207K554 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 01/02/1987 |
| From: | Conway R GEORGIA POWER CO. |
| To: | Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| NUDOCS 8701090448 | |
| Download: ML20207K554 (14) | |
Text
'e, 7
Georg.a Power Company 333 Piedmont Avenue Atianta Gemma 30308 Tdephore 434 526 6724 Maeg A.ecss Past 0"ce 804 4545 At anta Gcera a 30302 m5 N0 : 0 4 R.E.Conway Sero V ce PreMert January 2, 1987 Dr. J. Nelson Grace, Director United States Nuclear Regulatory Commission Region II Suite 2900 101 Marietta Street, Northwest Atlanta, Georgia 30323 NRC Docket 50-424 CONSTRUCTION PERMIT NUMBER CPPR-108 VOGTLE ELECTRICAL GENERATING PLANT - UNIT 1 STATEMENT OF COMPLETION AND REQUEST FOR LOW 1
POWER OPERATING LICENSE
Dear Dr. Grace:
By letters dated November 14, 1986, and December 22, 1986, I advised you of the status regarding activities at the Vogtle Electric Generating Plant, Unit 1 (VEGP-1).
The purpose of this letter is to advise you that Georgia Power Company is essentially complete with the design, construction, and preoperational testing of Vogtle Unit 1 and will be ready for issuance of an operating license on Friday, January 9, 1987.
All design, construction, and preoperational testing has been essentially completed so as to ensure compliance and consistency with the Final Safety Analysis Report ("FSAR") and supporting documents, other licensing-related documents, the Safety Evaluation Report ("SER") and the Commission's regulations.
We have utilized, and will continue to utilize the normal update process in order to ensure that all current design, construction, and preoperational testing documents are converted into submittals to the Commission pursuant with the regulation and that design documents are incorporated into the FSAR's descriptions.
On Monday, December 29, Georgia Power Company Nuclear Operations personnel met with members of your staff in order g
to discuss the preoperational test program.
At that meeting S
it was agreed that preoperational tests which Georgia Power Company was proposing for deferral until after fuel load would be reviewed and that the listing of deferrals would be 8701090448 87ogog PDR ADOCK 05000424 k
,, E 01 Ju
l Dr. J. Nelson Grace January 2, 1987 Page Two revised.
The revised preoperational test deferral list is included herein as Attachment 1.
This revised Attachment incorporates a number of comments made at the December 29 meeting involving deletions of specific tests, revisions, and format changes based upon progress in the testing program, additional reviews of the FSAR, and observations from the December 29 meeting.
Preoperational testing activities, l
except for those deferred, will be complete on or before fuel load.
Punchlist activities necessary to support fuel load have been identified and will be complete prior to fuel movement.
Other work which will not be completed prior to fuel load has been carefully evaluated in order to ensure that work does not affect our ability to safely load fuel and begin power testing.
This review process is ongoing, and we will ensure that deferred work items are properly scheduled for completion prior to the operational mode for which they are needed, that the work is integrated with ongoing operational activities, and that none of the deferred preoperational tests and punchlist items will affect the orderly progression of the power ascension program.
Georgia Power Company Management, the Plant Review Board, and the Safety Review Board have reviewed the deferred l
preoperational tests described in Attachment 1.
Based upon these reviews, Management, the Plant Review Board and the l
Safety Review Board have concluded that fuel load and l
subsequent startup testing activities can proceed safely and that completion of the deferred items as described is not an impediment to safe fuel load and subsequent startup.
We anticipate commencement of activ;'les designed to support fuel movement on Sunday, January 11, 1987.
Completion of design, construction, initial testing, and the review of deferred testing by Project Management, the Plant Review i
Board, and the Safety Review Board, have indicated that there will be no adverse impact to the safe operation of the plant or to the health and safety of the public.
We, thernfore, request that the Commission issue a license permitting commencement of fuel load activity no later than Friday, January 9, 1987.
l l
t
Dr. J.
Nelson Grace January 2, 1987 Pago Three As always, we are prepared to provide any additional information that you may require.
Sincerely, o
Conway R.
E.
REC / erd cc:
Service List Harold Denton
Georgia Power Company
,e l
Project Management Post Offica Box 282 Waynesbors, Georgia 30830 Telephone 404 724 8114 404 554 9961 l
l Vogtle Project l
Dr. J. Nelson Grace, Director i
January 2, 1987 Page Four c:
U. S. Nuclear Regulatory Commission Document Control Desk Washington, D. C.
20555 H. G. Baker D. R. Altman L. T. Gucwa J. P. O'Reilly P. R. Bemis C. W. Hayes G. F. Head J. A. Bailey G. A. McCarley P. D. Rice
- 0. Batum R. W. McManus R. H. Pinson G. Bockhold Sr. Resident (NRC)
C. W. Whitney C. E. Belflower C. C. Garnett (OPC)
B. M. Guthrie J. F. D'Amico J. E. Joiner (TSLA)
D. E. Dutton W. D. Drinkard D. Feig (GANE)
R. A. Thomas E. D. Groover NORMS Melanie A. Miller e
s L.
i i
ATTACHMENT 1 PREOPERATIONAL TESTING THAT MAY NOT BE COMPLETED PRIOR TO FUEL LOAD PREOPERATIONAL
- TEST / STATUS 1.
1-300-07/1001 RCS Hot Functional Retesting of portions of various systems tested during HFT needs to be performed.
The MSIVs were successfully hot stroke tested during HFT in preoperational test 1-3AB-01, Main Steam System.
Since then the MSIVs have been reworked and require retesting at temperature.
These reworks have not involved seating surfaces and seat leakage was not a problem during HFT.
These valves are not required to function to mitigate an accident until after initial criticality
(
and when the plant is steaming.
The operability of 3
the MSIVs is addressed in Technical Specification 3.7.1.5 which permits entry into Mode 3 provided at least one main steam line isolation system in the affected steam line is maintained closed.
Required i
surveillance testing will be performed prior to f
Mode 3.
The Stroke testing at temperature will be l
performed in Mode 3 when full temperature exists.
l The performance of the Auxiliary Feedwater System was demonstrated as close as practical to rated conditions during Hot Functional Testing in preoperational test 1-3AL-03, Auxiliary Feedwater System Testing During UFT.
During the testing i
it was determined that the flow orifices allowed F
greater flow than assumed in the limiting accident
[
analysis for containment pressure.
Due to this excess flow, the pump discharge flow orifices were modified to reduce flow and a modification was required on the flow control valves.
Response
times and flow parameters have been demonstrated to meet acceptance criteria by testing under cold conditions since the modifications.
An additional modification to the discharge and miniflow valves i
timing is needed to improve flow control. Retesting needs to be performed to demonstrate that required I
flow can be injected into the Steam Generators at operating pressure in Mode 3.
During Mode 3 the above testing will be performed.
This is not required prior to initial criticality since the required flow is based on the need to remove decay c
heat from the RCS.
1
This system is addressed in Technical Specification 3.7.1.2 which requires the system to be operable prior to Mode 3.
All Surveillance requirements will be met as required by this specification.
The RCS Power Operated Relief Valves (PORVS) were not completely tested in the preoperational test 1-3BB-05, Pressurizer Pressure and Level Control, performed during Hot Functional Testing.
The PORV's developed mechanical problems and could not be repaired until after cooldown from HFT.
During HFT one PORV successfully passed its time response test but the data was subsequently invalidated by mechanical aroblems that occurred.
The blowdown test of eac3 PORV was conducted during HFT but because of excessive seat leakage the data was invalidated. The valves were subsequently reworked and since they use system pressure to operate, they cannot be tested until the RCS is pressurized after fuel load.
Technical Specification 3.4.9.3 permits entry into Modes 4, 5, and 6 with inoperable PORV's provided that 2 RHR suction relief valves are operable.
The RHR suction valves will be opened during Modes 4, 5, and 6 until the PORV's are demonstrated operable as required by 4.4.9.3.1.
Technical Specification 3.4.4 allows entry in Mode 3 without prior demonstration of PORV operability.
Each valve will be demonstrated operable as required in Technical Specification surveillance requirement 4.4.4.1 a and b.
The HFT testing, even though invalidated by valve mechanical problems, provides a high degree of confidence that the blowdown and time response tests can be successfully performed after the valves have been declared operable.
During MODE 3 and prior to MODE 2, time response and blowdown testing will be completed.
Preoperational test 1-3BB-04, RCS Leak Rate test was unsuccessful due mainly to leakage from seven small manual valves.
Further testing was performed to demonstrate that repair of these valves would result in acceptable unidentified leakage rates.
These valves have since been repaired and RCS leak rate testing needs to be reperformed.
RCS leak rate is addressed in Technical Specification 3.4.6.2.
Unidentified Leakage will be verified within acceptable limits by performing the surveillance procedure for RCS inventory balance within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after entering MODE 4 per Technical I
2
Specification requirement 4.4.6.2.ld.
The deferred leak rate test will be performed at normal operating temperature and pressure during Mode 3.
During HFT in preoperational test 1-3BB-09, RTD Cross Calibration, the cross calibration of RTD-433C was not performed.
In addition, following HFT the wires have been replaced or respliced for all RTDs, except RTD 413B.
This requires reperformance of insulation and wire resistance measurements at ambient and hot conditions.
The ambient temperature resistance measurements have been completed.
A comparison of the original data taken in the preoperational test at ambient conditions to the retest data after completion of the rework has been performed.
This comparison demonstrates with a high confidence level that the rework had no affect on the operability of the RTDs and the RTDs are considered operational.
Insulation resistance and wire resistance measurements at hot plant conditions during Mode 3 will provide further assurance prior to entering Mode 2.
The cross calibration of RTD-433C and the hot resistance measurements will be completed after fuel load.
Seven of these RTDs are addressed in Technical Specification 3.3.3.6 which requires one RCS T and one RCS T instrumentchannelperloop$8tbe operableprio9gkNMode3.
RTD-433C is associated with the reactor vessel level instrumentation.
Its cross calibration will be completed prior to entering Mode 2, when full operating temperature is present.
The other RTDs are addressed in Technical Specifi-cation 3.3.1 which requires them to be operable prior to Mode 2.
The hot resistance measurements will be completed prior to Mode 2 when operating temperature is present.
During Hot Functional Testing in preoperational test 1-3BB-10, RTD Bypass, bypass flowrate were high because flow orifices had not been installed.
Based on data taken during Hot Functional Testing, orifices have been sized for installation in the RCS RTD bypass lines.
Demonstration that the RTD bypass lines flowrate is sufficient to meet the transport time criteria and function of the low flow alarm cannot be accomplished until after fuel load and the RCS is at normal operating temperature, pressure and flow.
Technical Specification Table 3.3-1 requires that the RTD channels for overtemperature delta T and Overpower delta T reactor trips be operable in MODES 1 and 2.
Operability per the Technical 3
Specification surveillance table 4.3-1 includes the bypass flowrate.
Bypass flowrate and low flow alarm will be tested during MODE 3 prior to Mode 2.
Prior to Hot Functional Testing in 1-3SQ-01, Digital Metal Impact Monitor test the DMIM system was functionally tested and demonstrated to detect loose parts.
That portion of the testing which demonstrated the system sensitivity to impacts by objects of various masses did not adequately meet the acceptance criteria of Regulatory Guide 1.133.
Additional testing is required with the reactor head installed to determine the actual sensitivity of the system so that the required report can be submitted to the NRC.
Loose parts detection is not addressed in Technical Specifications and is not required until the Reactor Coolant Pumps are run.
The deferred impact testing of sensors will be completed prior to starting the pumps.
The preoperational testing of the 1-3BB-06, Reactor Coolant Pumps Initial Operation, was successfully completed.
The number one seal leakoff for each RCP was found out of calibration during post calibration after Hot Functional Testing.
Since HFT the seal leakoff instruments have been recalibrated and seal leakoff data will be taken during RCP operation on plant heatup after fuel load.
Seal leakoff for the RCPs is not addressed in the Technical Specifications.
Due to replacement of the seal injection flow transmitters seal injection flow needs to be rechecked.
The RCP seal injection flowrate was checked during Initial RCP opndion.
However, prior to a confirmation post calibration of the flow transmitters, the transmitters were replaced and the validity of the seal injection flowrate during RCP operation cannot be verified.
After replacement of the transmitters the seal injection flowrate was reset via preoperational test 1-3BG-01 with simulated conditions.
Seal injection is considered controlled leakage as addressed in the Technical Specifications 3.4.6.2 which is required prior to Mode 4.
- However, according to this Technical Specifications the requirements of specification 4.0.4 are not applicable for entry into Mode 3 or 4.
Plant Operating Procedures direct that the injection flowrate is monitored and adjusted as required during RCP operation for RCS fill and venting, and plant heatup.
Seal injection will be checked and 4
recorded during startup testing at normal operating temperature and pressure prior to expiration of the 31 day surveillance period which begins upon entry into Mode 4.
Minor portions of 1-300-08 Thermal Expansion Testing, 1-300-09 Dynamic Response Testing, 1-300-11 Steady State Vibration were either incomplete or require retesting after fuel load.
The alternate charging line was not tested for thermal expansion or steady state vibration during HFT.
In order to avoid subjecting the alternate charging line to unnecessary thermal cycles, only the normal charging line was used.
The alternate line will be used in the start-up phase.
l The preoperational test of the system has shown it i
to be operational.
The vibration and thermal expansion testing will be done in Mode 3 when operating temperature and pressure conditions are available.
The steam generator blowdown sample lines were rerouted due to a design change.
The flow transmitter lines on the RCL crossover legs required retesting because of vibration observed during HFT and a subsequent modification made to correct the problem.
This testing will be reperformed during Mode 3 when operating temperature and pressure is available.
The auxiliary spray and pressurizer spray require retesting due to modifications made as the result of thermal expansion problem observed during HFT.
Preoperet intial testing has shown pressurizer spray to be fully functional.
The pressurizer is addressed in Technical Specification 3.4.3 which requires it to be operable prior to Mode 3.
All Technical Specification surveillance requirements will be met prior to Mode 3.
The thermal expansion testing will be completed during Mode 3 when operating temperature and pressure is available.
An RTD bypass line flow transmitter instrument line was rerouted due to thermal expansion problems noted during Hot Functional Testing.
This flow transmitter is required to consider the RTDs in that loop operational since the transmitter causes an alarm on low flow.
Technical Specifications requires the RTDs operational prior to Mod (
2.
The observation of the line will be performed in Mode 3 allowing the RTDs in this loop to be declared fully operable prior to Mode 2.
5
1 i
During heat-up of the main steam lines during HFT, two rigid supports at one support were bent due to bowing of the steam lines.
The struts were determined not to be rec uired for deadweight on thermal loads and were c.isconnected for the remainder of HFT.
After HFT the struts were replaced with a spring.
The cold set of the spring will be checked after fuel load prior to heat-up.
The hot set of the spring will be checked during Mode 3.
During HFT a cross-tie line from RHR to safety injection did not expand as predicted.
After evaluation of the data, it was determined that a snubber was probably bound.
Analysis has shown the stresses to be acceptable with the measured motion.
Thermal expansion of this line will be measured and i
again. compared to the model during Mode 3.
The spent fuel pool cooling system piping has not been observed for dynamic effects.
However, no indication of problems were found during the preoperational testing of the system.
This system is not required to operate until irradiated < fuel is in the storage pool.
Spent fuel pool cooling is not addressed in Technical Specifications.
This testing will be complete prior to Mode 2 which will be prior to t:1e potential of-irradiated fuel being unloaded in the storage pool.
2.
1-3SD-01/971 Digital Radiation Monitoring System 1-3SD-02/83%
The preoperational testing in 1-3SD-01 and 1-3SD-02 will be complete on tiiose monitors that are required for Mode 6 per Technical Specifications Tables 3.3-9 and 3.3-10 and Technical Specification 3.9.9.
Preoperational testing of the other monitors in these tests will be complete except for the performance of the Technical Specification Surveillance which is part of the test.
Remaining testing on Monitors which are addressed in Technical Specifications Tables 3.3-4, 3.3-8 and 3.3-10 will be tested prior to entering Mode 4.
The remainder of the system will be tested prior to entering Mode 2.
4 6
~,
3.
HEPA and Carbon Filter Package Testing The purpose of the HEPA and Carbon filter packages is to minimize the exposure of the public and plant personnel to radioactivity.
Three systems are addressed in Technical Specifications.
They include the ESF HEPA and Carbon filter systems which service the Control Room, Piping Penetration Area, and Fuel Handling Bldg.
No credit has been taken for the other HEPA/ Carbon filter systems in accident analysis and they are not required for the protection of the health and safety of the public.
The Control Room ESF ventilation filter package will be i
fully tested under surveillance procedures to meet preoperational and Technical Specification requirements prior to fuel load.
Testing on the filter packages serving the piping penetration areas may not be complete.
The other preoperational testing of this ventilation system will
)e complete with the exception of items addressed elsewhere in this attachment.
The Piping Penetration System is required per Technical Specification 3.7.7 for entry into form Mode 4.
The system functions in a recirculation mode during a LOCA to minimize the escape to the atmosphere of radioactivity released by a primary coolant leak through a penetration or any of the associated piping and pumps outside containment.
As such, this system is not required prior to initial criticality since the RCS will not be radioactive.
This system is addressed in Technical Specification 3.7.7 and is required prior to Mode 4.
The system will be fully tested under surveillance procedures to meet the Technical Specification and preoperational requirements prior to Mode 4.
Testing of the filter package serving the Fuel Handling Building may not be complete.
The other preoperational testing of this ventilation system will be complete.
The purpose of this system is to minimize radioactivity release from a dropped irradicted fuel assembly accident.
At fuel load the new fuel will be removed from the fuel handling building and loaded into the reactor.
The Fuel Handling Building filter package is addressed in Technical Specification 3.9.12 and is required whenever irradiated fuel is in the storage pool.
The system will be fully tested under surveillance 'procedures prior to Mode 2 when the first potential of 11aving irradiated fuel would be present.
7
- e Testing of additional filter packages may not be complete.
They are non-safety and are not addressed in Technical Specifications.
The other preoperational testing on these systems will be complete.
While the purpose of these packages is to minimize the release of radioactivity, no credit is taken for these filter packages to mitigate the consequences of design bases accidents.
These systems will be fully tested under non-technical specification surveillance test procedures to meet preoperational requirements prior to MODE 2.
4.
1-3GL-04/90%
Piping Penetration Filter Exhaust Preoperational testing will be complete except for the performance of the Auxiliary Building negative pressure test due to the air balance not being complete.
This system functions in a recirculation mode to minimize escape to the atmosphere of radio-activity resulting from a design basis accident and leakage of reactor coolant through penetrations of piping and pumps outside containment.
This system is required operable by Technical Specification 3.7.7 and is required prior to Mode 4.
Signifi-cant radioactivity is not present in the core or reactor coolant until after initial criticality (Mode 2).
The Auxiliary Building negative pressure test will be completed prior to Mode 4.
5.
Radwaste System Testing 1-3HB-02/51% - Waste Processing System-Liquid 1-3HE-01/85% - Boron Recycle System Testing of systems associated with the deferred radwaste building except as needed to support the Alternate Radwaste Building will not be completed.
Contractor facilities will be utilized.
Performance testing c2 the waste evaporator and Boron Recycle Evaporator may not be complete.
Other portions of the preoperational test will be complete.
These systems are not required to maintain the boric acid concentration in the RCS and are not safety related with the exception of safety related isolation valves.
These valves have been successfully tested in the preoperational test.
The normal function of these systems may be fulfilled by the demineralizers contained in the alternate radwaste facility.
These systems are not addressed in Technical Specifications.
The evaporator systems testing will be complete by the completion of the 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> warranty run or the alternate radwaste facility will be used on a continued basis.
8
6.
1-3RJ-01/86% - Proteus Computer Preoperational Test Portions of the testing of the Proteus computer system may not be complete at fuel load.
This includes the checkout of the following software:
Moveable Detectors, Watch program, and the Operational Communications link. The Proteus computer is a non-safety related aid to the licensee for plant operation.
It is not addressed in the Technical Specifications.
System testing will be completed prior to MODE 2.
7.
No Number Breathing Air System Assigned The system which will supply breathing air inside containment is not yet complete.
This system is not safety related or addressed in Technical Specifications.
It is, however, a system described in the Vogtle FSAR.
It is used to permit entry into the containment under conditions of airborne contamination.
Contamination should not occur until initial criticality.
The breathing air system will be complete and tested prior to initial criticality or other portable bottled air will be used when required.
8.
1-3RP-03/15% - Post Accident Monitoring System The plasma display provides a seismic class 1 display device for certain safety related instrumentation.
The system has been functionally tested; however, the preoperational test may not be completed.
Modifications have been made on the PROM's (Programmable Read Only Memories), however testing is not complete.
Instrumentation displayed on the plasma display (and elsewhere) is addressed in Technical Specification 3.3.3.6 but is not required to be opercble until Mode 3.
The required testing will be completed prior to MODE 3.
9.
IFI 424/86-121-02 Inspection Followup Item on NSCW Testing This item addresses the heat removal capacity testing of the NSCW system.
It is the licensee's position that adequate testing has been performed to the extent practical under cold conditions in the preoperational test.
However, additional testing for heat removal capacity will be performed as part of the power ascension testing when significant heat loads are present.
9
ra
- NOTE:
If the deferred testing involves associated testing of.
computer points, response timing or annunciators,.then portions of those associated preparational tests may not t
be completed for Fuel Load.
They will be completed consistent with the completion mode for the deferred.
testing.
4 f
4 10
--