ML20207H569
| ML20207H569 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 12/23/1986 |
| From: | Bailey J GEORGIA POWER CO., SOUTHERN COMPANY SERVICES, INC. |
| To: | Denton, Youngblood B Office of Nuclear Reactor Regulation |
| References | |
| GN-1233, NUDOCS 8701070563 | |
| Download: ML20207H569 (14) | |
Text
Georgia Pbwer Company Fcst Offica Box 282 Wayfusboro, Georgia 30830 Telephone 404 554-9961 404 724-8114 Southern Company Services,Inc.
Fest Office Box 2625 Birmingham, Alabama 35202 Telephone 205 870-6011 VOgtle Project December 23, 1986 Director of Nuclear Reactor Regulation File: X6BR01 Attention: Mr. B. J. Youngblood X7BC35 PWR Project Directorate #4 X7N14.2 Division of PWR Licensing A Log:
GN-1233 U. S. Nuclear Regulatory Commission Washington, D.C.
20555 NRC DOCKET NUMBER 50-424 CONSTRUCTION PERMIf NUMBER CPPR-108 V0GTLE ELECTRIC GENERATING PLANT - UNIT 1 DEFERRAL OF PRE 0PERATIONAL TESTING
Dear Mr. Denton:
Georgia Power Company (GPC) has provided the U. S. Nuclear Regulatory Commission, Region II, with a status for fuel loading (of Vogtle Electric Generating Plant (VEGP) Unit 1 on November 14, 1986 reference SL1578, 0529H). This included a list of preoperational tests that may not be completed at fuel loading. provides a revised list of preoperational testing that may not be completed prior to fuel load of VEGP-1 including a description of the deferred testing, justification for deferral, and a schedule for completion.
These preoperational tests or portions thereof are being included in power ascension testing.
s The deferral of the preoperational testing on the list is hereby referred to your office for concurrence since these are described in Chapter 14 of the VEGP Final Safety Analysis Report.
If your staff requires any additional information, please do not hesitate to contact me.
Sincerely,
.h.
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J. A. Bailey Project Licensing Manager Attachment JAB /dd xc:
R. E. Conway B. W. Churchill, Esquire P. D. Rice M. A. Miller (2)
R. A. Thomas B. Jones, Esquire G. Bockhold, Jr.
NRC Regional Administrator y
L. T. Gucwa NRC Resident Inspector AO C. W. Whitney D. Feig F
J. E. Joiner, Esquire Vogtle Project File l)
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ATTACHMENT 1 PREOPERATIONAL TESTING THAT MAY NOT BE COMPLETED PRIOR TO FUEL LOAD PREOPERATIONAL TEST 1-3QK Fire Detection System This system reduces the probability of fire damage to safety related equipment.
The portion of this system for which the testing is incomplete is in the Turbine Building and River Intake Structures which are non-safety related.
This system is not safety.related and is not addressed in Technical Specifications.
This testing will be completed prior to the completion of the 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> warranty l
run.
1-3SD Digital Radiation Monitoring System 1-3SD-02 The preoperational testing in 1-3SD-01 and 1-3SD-02 will be complete on those monitors that are required for Mode 6 per Technical Specifications Tables 3.3-9 and 3.3-10 and Technical Specification 3.9.9.
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.
1-3AB Main Steam System The MSIV's were successfully hot stroke tested during RFT.
Since then the MSIVs have been reworked and require retesting at temperature.
This testing is not required to be performed until the low power testing program per Regulatory Guide 1.68 paragraph 4.1.
There is no requirement for these valves to function to mitigate an accident until after initial criticality and when the plant is steaming.
The operability of the MSIVs is addressed in Technical Specification 3.7.1.5 which is a requirement prior to Mode 3 and will be performed prior to Mode 3.
The Stroke testing at temperature will be performed in Mode 3 when full temperature exists.
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1-3AL Auxiliary Feedwater System Testing During HFT The performance of the Auxillary Feedwater System was demonstrated as close as practical to rated conditions during Hot Functional Testing in preoperational test 1-3AL-03.
During the testing it was determined that the flow orifices allowed 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, to the extent this can be done at cold conditions.
Retesting needs to be performed to demonstrate that required 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 heat from the RCS.
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.
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.
The Control Room ESF ventilation filter package will be fully tested under surveillance procedures to meet preoperational and Technical Specification requirements prior to fuel load.
All systems which are addressed in Technical Specifications will be tested prior to fuel load.
These 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.
Testing on the filter packages serving the piping penetration areas may not be complete.
The other preoperational testing of this ventilation system will be 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.
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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 irradiated 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 having irradiated fuel would be present.
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 relea'se of radioactivity, no credit is taken for these filter packages to mitigate the consecuences of design bases accidents.
These systems will'be fully tested under vendor test procedures to meet preoperational requirements prior to commercial operations.
1-3GL 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 3/4 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 3
i (Mode 2).
The Auxiliary Building negative pressure test will be completed prior to Mode 4.
1-3GK Control Room Pressure Test The Main Control Room pressure test may not be complete at fuel load.
That is, the acceptance criteria of 1/8 inch water positive pressure may not be yet achieved.
Preliminary testing has achieved about 1/16" and sealing activities are reducing leakage.
The basis for the 1/8 inch water positive pressure is thyroid and whole body radiation doses resulting from infiltration of radioactivity after a design basis accident.
The system is required operable by Technical Specifications prior to Mode 4.
Significant radioactivity will not be present in the core or reactor coolant until after initial criticality (Mode 2).
The Main Control Room pressure test will be completed prior to Mode 4.
1-3BB Pressurizer Pressure and Level Control The RCS Power Operated Relief Valves (PORVS) were not completely tested in the preoperational testing performed during Hot Functional Testing.
The PORV s developed mechanical problems and could not be repaired until after cooldown from HFT.
They were subsequently reworked and since they use system 7ressure to operate, they cannot be tested until tie RCS is pressurized after fuel load.
Per the Technical Specification 3.4.4 both valves are required to be operational prior to Mode 3.
Prior to Mode 3 each valve will be demonstrated operational as required in the Technical Specification surveillance requirement 4.4.4.1 a & b.
That is each valve will be operated through one complete cycle of full travel and a channel calibration will be performed.
During MODE 3 and prior to MODE 2 time response and blowdown testing will be completed.
During HFT one PORV successfully passed its time response test but the data was subsequently invalidated by mechanical problems that occurred.
The blowdown test of each PORV was conducted during HFT but because of excessive seat leakage the data was invalidated.
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This testing, even though invalidated b valve mechanical problems provides us a high degree of confidence that the blowdown and time response tests can be successfully performed after the valves have been declared operable.
For Cold Overpressure Pressure protection the l
Technical Specification requirement is that either two PORVs or two RHR suction relief valves be operable (RHR suction relief set at 450 + 1% psig) for MODES 4, 5, and 6 (Reactor head installed).
The RHR suction valves will be open during MODES 4,
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5, and 6 until the PORVs are demonstrated operational as identified above such that the requirements of Technical Specification 3.4.9.3a and b are met.
The pressurizer level control was successfully tested during HFT.
After this a problem with the height of the condensate pots was identified resulting in two level transmitters requiring respanning and a third requiring replacement because of more stringent spanning requirements..
Since modification or adjustment of the associated electronics for level control and indication is not required and the controls successfully past preoperational testing during Hot Functional Testing there is no need for retest of the control circuits.
The only testing required is to respan and reperform a calibration on the transmitters.
The operability of the level control is addressed in Technical Specification 3.4.3 which requires the pressurizer operable prior to Mode 3.
The recali-bration of these instruments will be completed prior to entering Mode 3.
1-3BB RCS Leak Rate The Hot Functional 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 will be reperformed.
RCS leak rate is addressed in Technical Specifica-tion 3.4.6.2 which requires that the Unidentified Leakage be within acceptable limits prior to Mode 4.
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To comply, the surveillance procedure for RCS inventory balance will be run within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of entering MODE 4 per Technical Specification requirement 4.4.6.2.ld.
The deferred leak rate test will be performed at normal operating temperature and pressure during Mode 3.
1-3BB RTD Cross Calibration During HFT 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 measuremu.:s 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.
l Seven of these RTDs are addressed in Technical Specification 3.3.3.6 which requires one RCS T l
l and one RCS T instrumentchannelperloop$8Ube operableprichb$ Mode 3.
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.
1-3BB RTD Bypass During Hot Functional Testing the RTD bypass flow lines flowrate was 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.
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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 Specification surveillance table 4.3-1 includes the bypass flowrate.
Bypass flowrate and low flow alarm will be tested during MCCE 3 prior to Mode 2.
1-3SQ Digital Metal Impact Monitor l
Prior to Hot Functional Testing the DMIM system i
was functionally tested and demonstrated to detect loose parts.
n at portion of the testing which demonstrated the system sensitivity to impacts by objects of various masses did not adequately meet the acceptance criteria of Regutatory Guide 1.133.
Additional testing is required to determine the actual sensitivity of the system so that the i-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.
1-3RP Post Accident Monitoring System l
The plasma display is the seismic class 1 display device for the instrumentation listed in Technical Specifications 3.3.3.6.
The system has been functionally tested; however, the preoperational test will not be completed until a design modification is implemented that changes out some of the PROM's (Programmable Read Only Memories) and modifies the internal wiring.
This display is not requi' red to be operable until Mode 3.
The change out of the PROMS and completion of testing will occur prior to MODE 3.
1-3HB Waste Processing System-Liquid The testing of the waste evaporator with boric l
acid solutions may not be complete.
Other portions of the preoperational test will be complete.
This 7
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system is not required to mafatain the boric acid concentration in the RCS and is not safety related with the exception of safety related isolation valves.
These valves have been successfully tested in the preoperational test.
The system is not addressed in Technical Specifications.
The system -
testing will be complete by completion of the 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> warranty run.
1-3HE Boron Recycle System The testing of the recycle eva) orator with boric acid may not be completed.
Other portions of the preoperational test will be complete.
This system is not required to maintain the boric acid i
concentration in the RCS, is non-safety related, and is not addressed in Technical Specifications.
The system testing will be complete prior to Mode 3.
1-3RJ 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, Graphics program, Watch program, Operational Communications link, and testing of computer points associated with portions of deferred testing described in this attachment.
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 as follows:
Prior to Moveable detectors Mode 5
(
Graphics Mode 4 Watch Mode 2 l
OCL Mode 3 Computer Points Mode of completion i
of applicable deferred item.
l l
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1-3RP Emergency Response Facility (Computer)
Testing of a small number of computer points in ERF computer system may not be complete for fuel
- load, nese are the computer points associated with portions of the deferred testing described in this attachment.
The testing will be completed consistent with the completion mode referenced in this attachment for the deferred testing.
1-300 Reactor Trip & ESFAS Process Channel Logic Response Time 1-300 Response Time Data Response time testing of a few items may not be complete by fuel load.
This consists of the response time testing associated with deferred testing described in this attachment.
This testing will be completed consistent with the completion mode referenced in this attachment for the deferred ecsting.
Radwaste Systems Testing Associated with the Deferred Radwaste Building Testing of systems associated with the deferred i
radwaste building except as needed to support the Alternate Radwaste Building will not be complete.
I Contractor facilities will be utilized.
1-3BB RC Pump Initial Operation The preoperational testing of the Reactor Coolant Pumps was successfully completed.
The number one seal leakoff for each RCP was found out of i
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 operation.
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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 3.4.6.2.1 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 inj action flowrate is monitored and adjusted as required during RCP operation for RCS fill and venting, and plant heatup.
Seal injection will be checked and 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.
1-300 Thermal Expansion Testing 1-300 Dynamic Response Testing 1-300 Steady State Vibration Minor portions of these tests were either incomplete or require retesting after fuel load.
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.
This system is indirectly addressed in Technical Specification 3.9.11 which requires a minimum level when irradiated fuel is in the storage pool.
This testing will be complete prior to Mode 2 which will be prior to the potential of irradiated fuel being unloaded in the storage pool.
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.
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The preoperational test of the system has shown it 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 i
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.
Preoperational 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 cperating 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 Mode 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.
During heat-up of the main steam lines during HFT, i
two rigid supports at one support were bent due to bowing of the steam lines.
The struts were determined not to be required for deadweight on thermal loads and were disconnected 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.
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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 again compared to the model during Mode 3.
1-300 Common Annunciators 1-3RK Plant Annunciators The testing of 5 breakers in 480 V switchgear is incomplete because these breakers lack a 52b contact to cause the "Switchgear Trouble" common alarm upon tripping.
New breakers with these contacts have been ordered.
These breakers are in MCCs which are safety related.
However, these breakers feed 1-E battery chargers which renders the common MCC trouble alarm redundant since each charger has a trouble alarm which would alarm on loss of power to the charger.
Because of this, these common MCC trouble alarms are not required for safe operation even though MCC trouble alarms are described generically in the FSAR.
These l(
alarms are not addressed in Technical Specifications.
New breakers will be obtained, installed and tested prior to Mode 1.
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 to which Vogtle has committed in the FSAR.
It is required to permit entry into the containment under conditions of airborne contamination.
This condition cannot 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.
1-3EF NSCW Heat Capacity Test A proposed Inspector Followup Item (Inspection Report 424/86-121) has stated that the licensee has not performed preoperational testing to demonstrate the cooling capacity of the ultimate heat sink.
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The licensee's position is that adequate testing has been performed in the preoperational test through the testing of design parameters such as pump flowrates, and individual heat exchanger flowrates.
The ability of the system to operate under initiating signals and controls has been demonstrated to maintain acceptable water temperature.
All of these requirements have been met in the system preoperational tests.
In addition, the NSCW system was demonstrated to have adequate capacity and control to provide cooling water to maintain all loads in place during Hot Functional Testing.
However, Hot Functional Testing did not simulate the major NSCW normal operating load of decay heat removal.
Such a test cannot be performed until after operation at power and is not required by the committed regulatory guides.
Technical Specifications limit NSCW tower water temperatures to 90 F to assure operation within the design envelope.
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