ML19289E723
| ML19289E723 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 04/25/1979 |
| From: | Widner W GEORGIA POWER CO. |
| To: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| NUDOCS 7905290029 | |
| Download: ML19289E723 (15) | |
Text
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Georgia Power Company pg&
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230 Peachtros Street M, b Post office Box 4545 O
Atlanta. Georgia 30302 Te!ephone 404 522-6060 m
Georgia Power Power Generation Department e,,
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April 25, 1979 United States Nuclear Regulatory Co:.si.ssion Office of Inspection and Enforcenent Region II - Suite 3100 101 Marietta Street ATTENTION:
Mr. James P. O'Reilly Gentlemen:
The following information is submitted in response to your letter of April 14, 1979 requesting information concerning I.
& E.
The ten attachments address the first ten items of I.
E.
Bulletin 79-08, " Events Relevant to Boiling Water Reactors Identified During Three Mile Island Incident".
As stated in the Buelltin response to item 11 is required in 30 days.
Should you have any questions, please contact my office.
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W. A. Widner JAB /mt xc:
U.
S. Nuclear Regulatory Commission Office of Inspection and Enforcement Division of Reactor Operations Inspection Washington, D. C.
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J. H. Miller, Jr.
D.
E.
Dutton W.
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Ehrensperger W. A. Widner F. G. Mitchell, Jr.
J.
R. Jordan C.
F. Whitmer i- /
W.
M. Johnston, Jr.
R. J.
Kelly i
M. Manry C.
R.
Thrasher
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C.
R. Miles, Jr.
K. M. Gillespie
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Belflower C. W.
Hayes 1
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Groover G.
E. Spell, Jr. fl 2 2.*$.tfH.j son, Jr.
Route to SRB Members Y $/
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& E.Bulletin 79-08 Question 1 The action for Item one is complete and documented on training data sheets except for 3 license personnel who were not available at the time the training was conducted. Training for these license holders will be conducted when they return The above training was also presented to to the plant site.
site.
all current license applicants in school at the plhnt 2045
- 80
I.
& E.Bulletin 79-08 Question 2 In review of the Unit 1 and 2 containment isolation systems, several penetrations have been identified as not automatically isolated during or prior to automatic initiation of the ECCS, but which can be isolate without being detrimental to the operation of the ECCS or Safety Ejection System.
There are eight T48 system valves on Unit 1 that are normally closed valves with the purpose of opening and inerting the primary containment atmosphere with nitrogen during normal or post accident conditions.
For Unit 2 containment valves, where the primary containment has been provided with hydrogen recombiners, eight T48 system valves are kept closed during all plant conditions since inerting is not required with two 100 percent hydrogen recombiners available.
Should it become necessary to open the Unit 1 inerting valves operator action will be required per the post accident venting procedures.
An annunciator is annunciated in the main control room at all times that these remote manual inerting valves are in the abnormally open position.
Four remote manual isolation valves on Unit 2 will be cycled post accident from the closed to open to closed position intermittently to provide a means of controlling combustive gas concentration within the primary containment.
For Unit 2 the four remote manual venting valves will be kept closed post accident.
The post accident venting procedure will be revised if required prior to start-up of U it 1 and 2 to assure that proper venting or treatment of cxplosive gases is made.
There are two drywell pneumatic system isolation normally open valves provided for each Unit for operation of pneumatically operated valves in the primary containment.
These valves will remain open during all modes of operation both normally and post accident since they are under a pressure greater than containment and are required for valve operation.
9 2045 181.
I. & E.Bulletin 79-08 Question 3 Automatic actions fulfill the requirements for proper functioning of the auxiliary heat removal systems that are Manual used when the main feedwater system is not operable.
actions are not necessary but are used to reduce the cycling of temperatures, pressures and water level within the Reactor vessel during the lost of all feedwater transient.
Loss of all reactor feedwater could be indicated by:
Rapidly decreasing reactor water level, loss of feedwater flow indication, loss of feedwater system pressure and several annunciators such as Reactor Low Water Level and Recirc Flow Limit.
The Automatic Actions would begin with Reactor Recirculation pumps run back (Feedwater 20%) and the Reactor would scram when the Reactor Vessel water level decreased to 12 1/2 inches.
Also, at this same low water level, the Standby Gas Treatment System initiates and Primary Containment Isolation valves close.
Should the Reactor Vessel water level continue to decrease to -38 inches, the Reactor Recirculation pumps will this trip and the Main Steam Isolation Valves will close. At same low (-38) Reactor vessel level, the High Pressure Coolant Injection System and the Reactor Core Isolation Cooling System will initiate and inject to the Reactor Vessel.
Further Automatic action would be initiation of RHR (low pressure coolant injection) and the Core Spray System should the Reactor Vessel level reach - 146.5 inches. Also at this level, the diesel generators would start and the drywell cooling fans would trip.
No operator action Required.
The manual actions are performed concurrently with the Automatic Actions and serve as a verification and a back-up to the Automatic Actions.
The operator attempts to restore feedwater flow to the reactor vessel and if this fails, he will scram the reactor by placing the mode switch in SHUTDOWN.
The operator will manually close all main steam isolation valves and manually initiated the Reactor Core Isolation Cooling System to provide makeup water to the Reactor Vessel.
the If the Reactor Water level decreases to -38 inches, operator will monitor automatic operation of the Primary Containment Isolation System, Emergency Core Cooling Systems and Nuclear Steam Supply Shutoff Valves.
The operator can manually secure the High Pressure Injection System when reactor water level is confirmed Coolant to be above the low level scram point and return the system to normal standby condition.
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I.
& E.Bulletin 79-08 Question 3 continued The operator will monitor automatic initiation of Core RHR (low pressure coolant injection) diesel generators,
- Spray, and drywell cooling fans trip should the reactor vessel water level decrease to -146.5 inches.
When reactor water level is restored and stablizied at 37 inches (normal level), the operator can secure Reactor Core Isolation Cooling and return the sytem to normal standby condition.
Reactor shutdown will continue per HNP-1025 (Fast Reactor Shutdown) and HNP-2000 (Annunciator Response) procedures.
Emergency Core Cooling Systems are no longer needed for the Reactor shutdown therefore the operator can manually initiate modes of these systems to bring the reactor to a controlled cold shutdown condition.
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I.
& E.
Bulletin 79-08 Question 4 Both units have the same vessel level instrumentation in the control room which indicate directly vessel level.
Indication varies from 200 to 900 inches vessel level measured from the bottom head drain.
The various control room indications are listed below.
TABLE 1 TYPE HO.
RANGE USE Gemac Ind.
3 0-60" Instrument zero is at 517" vessel level, used for level control in normal range, loop has alarms at 32" and 42".
Pressure and temperature compensated.
Gemac Rec 1
0-60" Recorder for normal water level.
Yarway Ind.
2
-150 to +60" 367" to 577" vessel level used for monitoring abnormal water level above active fuel.
Primary C/R indication for ECCS initiation levels, and LOCA Temperature compensated.
Gemac Ind.
1 200" to 500" Used for monitoring water level in core during LOCA active fuel is from 203.5" to 352.5" 2/3 core height is 304.5" Yarway 1
200 to 500" Used for monitoring water level during LOCA.
Gemac 1
500 to 900" Used for vessel flooding during shutdown, cold, calibration.
. 3 I O.}
Bulletin 79-08 Question 4 continued I.
& C.
In addition to C/R indication there are several local One distinction is indications in the Reactor Building.
that most of these do not require electrical power tofrom 200" to 577 in operate.
They range' The various local indications are the same for both units and are listed below:
TABLE 2 TYPE NO.
RANGE USE Yarway Ind.
2
-150 to +60 Local and remote (control room indication.
Yarway Ind.
10
-150 to +60 Abnormal water level monitoring and ECCS initiation i.e. see Table 3.
Yarway Ind.
2 200 to 500" Used for monitoring level during LOCA.
Containment Switch Spray permssive at 313.5" Yarway 1
200 to 500" Used for Unit 1 Remote Shutdown Panel indication.
Gemac 1
-150 to -60" Used for Unit 2 Remote Shutdown Panel Indication.
Barton Ind.
4 0 to +60" Local Normal Level monitor and Rx Scram and Turbine Switch Trip on High Level.
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& E.Bulletin 79-08 Question 4 continued TABLE 3 ECCS ACTUATION SETPOINTS TYPE MO.
RANGE SETPOINTS Gemac 3
0-60" Rx Low Level Alarm +32" Rx High Level Alarm +42" Barton 4
0-60" Rx Scram +12.5" Turbine RCIC HPCI Trip +58" Yarway 4
-150 to +60" Close MSIVs
-38" Trip Recirc
-38" Yarway 4
-150 to +60" Start RCIC
-38" HPCI
-38" LPCI
-146.5" CS
-146.5" D/G
-146.5" ADS
-146.5" Yarway 2
-150 to +60" ADS perm +12.5 Yaruay 2
200 to 500" Cont. Spray Perm 313.5 Table 3 listed the level instrumentation which automatically initiates safety systems.
These are located in the reactor building on the 158' elevation.
The 200 to 500" instrumentation and remote shutdown panel is on the 130' elevation or ground level.
The control room is on the 164' elevation.
Manual initation of safety system can be based on information of 19 separate indication, 14 (15 on Unit 1) which do not require an external power source to indicate water level of the reactor vessel.
All instruments measure directly the level in the vessel at all times.
The range of the instrumentation covers from just below the active fuel to well above normal water level.
Each range has several indications available for use, either in the control room or local.
Licensed personnel have been instructed to utilize all available instrumentation as indicated in the response to Question 2.
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I.
& E. 79-08 Question 5 A review has been made of the HNP 1 and 2 procedure concerning operator actions to be taken in overriding automatic actions of safety features.
The HNP procedures caution the operators to ensure that the water level as well as other that the condition is normal.
plant parameters are indicatin s The condition is stabilized before changing the system from its automatic function.
The HNP procedures are written so that under these conditions continued system operation when systems are no longer required will be cdrtailed to avoid progressing into an unsafe condition which could result in a unsafe plant condition.
5.b.
Refer to Response to Question 1.
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I.
& E.
Bulletin 79-08 Question 6 HNP-1 and HNP-2 are currently in cold shutdown.
Prior to returning the units to service, a complete valve line-up is performed for each unit for safety related valves.
A review of operation procedures for safety related systems and piping drawings (indicating normal valve position) was conducted for both Hatch Units to identify discrepancies between procedures and drawings.
Each unit is equipped with a safety system status panel that provides visual indication of system (s) not operable.
The shift foreman on duty maintains the status panel for his unit.
When a safety related valve is removed from service for maintenance, a clearance sheet is initiated by the shift foreman to isolate the valve for the work to be performed.
This clearance sheet requires double verification signatures, in addition to the shift foreman's signature, when reviewing the valve from service and when returning the salve to service.
Maintenance procedures require repair work to be performed using and the shift foreman to be notified prior to performing any work.
Should initial investigation into the problem reveal additional malfunctions, the procedure instructs personnel to notify the shift foreman immediately.
Surveillance requirements for technical specifications compliance are listed in the maintenance procedures.
When maintenance is complete on a safety related valve, the shift foreman reviews the worksheet and performs a functional test on the valve.
Double signature verification plus the shift foreman's signature on the clearance sheet verifies the valve is returned to the proper position.
Surveillance is performed frequently on safety related valves to prove operability.
The procedures contain restoration valve line-up data sheets that are performed by plant personnel and reviewed and signed by a licensed operator.
The data sheets are also reviewed by a shift foreman.
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I.
& E.Bulletin 79-08 Question 7 There are two systems at our facility that are designed to transfer potentially radioactive gases and liquids out of These systems are the Drywell Vent the primary containment.
and Purge system and the Radwaste Sump subsystem.
The Unit 1 Vent and Purging system consists'of arrangements of three remotely operable two inch valves in series which are normally and administratively kept closed valves during all modes of plant operation except when venting is required the control panel and an operator is constantly required at at all times that the 3 valves are open per the " Primary These 3 Containment Atmospheric Control system" procedure.
valves on the Unit 1 Purge and Vent System are part of the Containment Atmospheric Dilution (CAD) system and are used in lieu of the normal vent valves.
The normal vent valves are used on Unit 2 for normal venting and they isolate on a containment isolation signal and/or high reactor building and/or refueling area radiation signal while the three CAD system valves require an operator to manually close them by Any time procedure when an isolation signal is received.
the CAD system valves are opened there is an audible and visible alarm of the abnormal condition in the main control There should be no room for inadvertently opening room.
these valves since the operator must manipulate three specific controls for each valve arrangement to vent gas from the containment atmosphere.
The normal Vent and Purge system valve arrangement consist of two 2 inch automatic isolation valves in series which isolate closed on a containment isolation signal or high reactor and/or refueling area radiation signal. To allow for venting post accident the containment isolation signal may be intentionally by passed by operation of an "Overrride Interlock" switch.
The Radwaste Sump subsystems are provided to collect and transport normal liquid leakage from the primary containment to the Radwaste Treatment systems.
The pump discharge valves are automatically closed by the containment isolation The annunciator response procedure then states for signals.
the operator to manually close those pump discharge valves once containment isolation or reactor and/or refueling floor high radiation signal is receive prior to resetting the isolation signal. Although the existing annunciator response procedure states that following investigation of the isolation the and correction of conditions causing the isolation, isolation signal can be reset and the pump d sch}qgg valves 20 J
opened.
I.
& E.Bulletin 79-08 Question 7 continued A procedure revision was submitted to assure that the discharge valves will not be opened without shift foreman approval until it is confirmed that the discharge valves are not required to be closed.
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IEB 79-08 Ouestion 8:
Item 8.a. requested the review and modification as necessary of maintencnce and test procedures to verify the operability of redundant saf ety-related systems Operability in prior to the removal of any safety-related system from service.
systems is serified through the inspection of clearance sheets and redundant tags, verbal communientions, and visual inspection of System Inoperability Status insure each
. Procedures for maintenance and testing are being reviewed to board.
procedure has a precautionary step to verify through inspection redundant system This review is to be completed prior to return to startup of the operability.
applicable Unit.
Item 8.b. requested the review and modification as necessary of maintenance and test procedures to verify the operability of all safety-related systems when they This review has been are returned to service following maintenance or testing.
The revisions completed and some revisions to procedures found to be necessary.
will be completed prior to return to startup of the applicable Unit.
Item 8.c. requested the same review as 8.a. and 8.b. in the area of notification of reactor operational personnel whenever a safety-related system is removed from This review has been completed with plant procedures and returned to service.
found to be adequate.
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IEB 79-08 Ouestion 9:
Review of the plant prompt reporting procedure, emergency procedure, and the emergency plan indicates the following:
1.
Technical Specification and procedure HNP-450 " Reportable Occurrences" require notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as a reportable occurrence for an incident of this type.
2.
The emergency procedure for site and general emergency require notification of off-site parties by the emergency call list on which the NBC is listed.
No time limit is given however.
Complying with the one-hour rule would require revision of emergency procedures or generating new procedures, if required.
Complying with the open continuous communication channel is a problem since it would not be clear where or how to do this.
The emergency control center may or may not be at the plant site during an incident.
It will be necessary to study this requirement in depth to determine the best available method for communication with the NRC during an incident.
This study will be completed by July 1, 1979.
Input will be solicited from the on-site resident inspector and NRC NRR Project Manager in an on-site meeting to be held on April 30, 1979.
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IEB 79-08 Ouestion 10:
A review of UNP 1 and 2 procedures which dmis with significant amounts of that hydrogen gas that may be generated during a transient or other accident or be released to the containment would either remain inside the primary r-stet, has been conducted by the Operations Department.
During the review of the procedures dealing with the H2 e neentration within the primary containment several discrepancies were identified. These discrepancies were noted in the Post Iccident Venting procedure for HNP-2.
This procedure failed to refer the rec mbiner systems de-operator to the procedure on utilizing the two 100% H2 Itsigned to control and recombine the buildup of H2 within the containment.
should be noted that the operator was referred to the procedure for utilizing 2 recombiner systems in procedure titled Pipe Break Inside Primary Con-the H tainment. During the review of the procedures c0ncerning H2 concentration in the primary system, no discrepancies were identified. Procedures with dis-crepancies which were identified during this review will be submitted to the PRB f or approval and re-issue prior to HNP-2 reactor return to power.
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