ML19289E725

Forwards Revised Responses to IE Bulletin 79-08, Events Relevant to BWRs Identified During TMI Incident. Replaces Previous 790425 Submittal (Mentioned as 790415 in Present Ltr).Includes Response to Item 11

ML19289E725
Person / Time
Site:	Hatch
Issue date:	05/09/1979
From:	Widner W GEORGIA POWER CO.
To:	James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
References
NUDOCS 7905290035
Download: ML19289E725 (16)

Text

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E Georgia Power Company fp 230 Peachtree Street Post Off,ce Box 4545 M OY gg Atlanta Georgia 30302

.' Telephone 404 522-0000 ,g..; .

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n Power Generation Department g, ,

79 tuY 10 A 9.16 it.o ern m ,wic saem May 9, 1979 United States Nuclear Regulatory Commission Office of Inspection and Enforcement Region II - Suite 3100 101 Marietta Street Atlanta, Ga 30303 ATTENTION: Mr. James P. O'Reilly Gentlemen:

On April 15, 1979 Georgia Power Company provided to the NRC information concerning I. & E.Bulletin 79-08, Events Relevant to Boiling Water Reactor Identified During Three Mile Island Incident. Subsequent to that submittal we have revised our response to clarify certain portions and to provide additional information where necessary.

The attached revised responses to the first ten items of I. & E.Bulletin 79-08 replace our previous submittal in its entirety. A response to item eleven is also included.

Should you have any questions, please contact my office.

7#'d. W2w W. A. Widner Manager Power Generation JAB /mt xc: U. S. Nuclear Regulatory Commission Office of Inspection and Enforcement Division of Reactor Operations Inspection Washington, D. C. 20555 D. E. Dutton

, :2043 253r bc: J. H. Miller, Jr.

W. E. Ehrensperger R. W. Staffa F. G. Mitchell, Jr. J. R. Jordan C. F. Whitmer W. M. Johnston, Jr.

R. J. Kelly M. Manry C. R. Thrasher C. R. Miles, Jr.

K. M. Gillespie C. E. Belflower C. W. Hayes E. D. Groover G. E. Spell, Jr. G. H. Burson, Jr.

2oute to SRB Members 79052900~$5 on'a u oomy.

I. & E.Bulletin 79-08 Question 1 4

The action for Item One is complete and documented on training data sheets for all license personnel. Training was also presented to all current license applicants in sch'ool at the plant site.

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I. & E.Bulletin 79-08 Question 2 In review of the Unit 1 and 2 containment isolation systems, several penetrations have been identified that do not automatically isolate during or prior to automatic initiation of the ECCS, but which can be isolated without being detrimental to the operation of the ECCS or Safety Ejection System. The penetrations identified are used for the Containment Nitrogen Inerting System make up lines and the Containment Atmospheric Dilution (CAD) System lines.

The Nitrogen Inerting make up lines each contain three normally closed remotely operable two inch valves in series which are used for inerting the primary containment atmosphere with nitrogen during normal or post accident conditions.

Should it become necessary to open the inerting valves post accident, operator action would be required per the post accident venting procedures. An annunicator is annunciated in the main control room at all times that these remote manual inerting valves are in the open position.

The CAD system is used as an alternate method of venting and purging the Unit 1 containment. The valve arrangement for the CAD system is described in the response to question Unit 2 primary containment has been provided with two 100% hydrogen recombiners. Therefore, the Nitrogen Inerting and the CAD valves are kept closed during all plant conditions since inerting is not required.

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e I. & E.Bulletin 79-08 Question 3 Automatic actions fulfill all the requirements for proper functioning of the auxiliary heat removal systems that are required when the main feedwater system is-not operable. Manua) actions are not necessary but are used to reduce the cycling of temperatures, pressures, and water level within the Reactor Vessel during the loss of feedvater transient.

Loss of reactor feedwater could be indicated by: Rapidly decreasing reactor water level, loss of feedwater flow indication, loss of feodwater system pressure and several annunciators such as Reactor Low Water Level and Recirc Flow Limit.

The Automatic Actions would begin with Reactor I<ecirculation pumps run back at Feedwatern 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 trip and the Main Steam Isolation Valves will close. At this same low Reactor vessel level (-38 inches) , the High Pressure Ccolant Injection System and the Reactor Core Isolation Cooling System will initiate and inject to the Reactor Vessel.

Further Automatic action would be' the 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.

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 SHUTDONN.

The operator will manually close all main steam isolation valves and manually initiate the Reactor Core Isolation Cooling System to provide nakeup water to the Reactor Vessel.

If the Reactor Water level decreases to -38 inches, the 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 Coolant Injection System when reactor water level is confirmed 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 Spray, RIIR (low pressure coolant injection) diesel generators, and drywell cooling fans trip should the reactor ves,sel water level decrease to -146.5 inches.

When reactor water level is restored and stabilized at

+37 inches (normal level), the operator can secure Reactor Core Isolation Cooling and return the system to normal standby condition.

Reactor shutdown will continue per HNP-1, 2-1025 (Fast Reactor Shutdown) and.HNP-1, 2-2001 (Annunciator Response) procedures. Emergency Core Cooling Systems are no longer needed for the Reactor shutdown therefore the operator can manually control these systems to bring the reactor to a controlled cold shutdown condition.

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I. & E.Bulletin 79-08 Question 4 Hatch Units 1 and 2 have similar vessel level instrumentation in 'he 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 NO. 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.

Gemac Ind. 1 200 to 500" Used for monitoring water level in core during LOCA, active fuel is from 208.5" to 352.5" 2/3, co e height is 304.5" Yarway Ind. 1 200 to 500" Used for monitoring water level during LOCA.

Gemac Ind. 1 500 to 900" Used for vessel flooding during shutdown.

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I. & E.Bulletin 79-08 Question 4 continued .

In addition to C/R indication there are several local indications in the Reactor Building. One distinction is that most of these do not require electrical power to operate.

The majority of these instruments are located in the reactor building on the 158' elevation. The 200 to 500" instrumentation and remote shutdown panel are on the 130' elevation or ground level. The control room is on the 164' elevation in the control building. They range from 200 to 577 inches measured form the bottom head drain. 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. See Table 3.

Yarway Ind. 2 200 to 500" Used for monitoring level Switch during LOCA. Containment Spray permissive at 313.5" Yaruay Ind. 1 200 to 500" Used for Unit 1 Remote Shutdown Panel indication.

Gemac Ind. 1 -150 to -60" Used for Unit 2 Remote Shutdown Panel indication.

Barton Ind. 4 0 to +60" Local Normal Level monitor Switch Rx Scram and Turbine Trip on High Level.

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I. & E.Bulletin 79-08 Question 4 Continued Table 3 lists the level instrumentation which automatically initiates safety systems.

TABLE 3 ECCS ACTUATION SETPOINTS TYPE NO. RANGE SETPOINTS Gemac 3 0-60" Rx Low Level Alarm +32" Rx High Level Alarm +42" Barton 4 0-60" Rx Scram +12.5" Main Turbine, RCIC, HPCI Trip +58" Yarway 4 -150 to +60" Close MSIVs -38" Trip Recirc -38" Yarway 4 -150 to +60" Start RCIC -30" HPCI -33" LPCI -146.5" CS -146.5" D/G -146.5" ADS -146.5" Yarway 2 -150 to +60" ADS permissive +12.5 Yarway 2 200 to 500" Containment spray permissive 313.5 Manual initiation of safety system can be based on information of 19 separate indication, 14 ('S 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 .antrol room or local.

Licensed personnel have been instructed to utilize all available instrumentation as indicated in the response to Question 1.

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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 are written so that continued system operation will be curtailed to avoid progressing into an unsafe condition when systems are no longer required. The HNP procedures caution the operators to ensure that the water level as well as other plant parameters are normal before changing the system from its automatic function.

5.b. Refer to Response to Question 1.

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I. & E.Bulletin 79-08 Question 6 Prior to returning the units to service *from aa extended

- outage, a complete valve line-up will be performed of safety related valves. A review of there procedures agairst Piping and Instrument Drawings, which indicate normal valve; position, was conducted for both Hatch Units to identify discrepancies between the procedures and drawings. All discrepancies identified will be corrected prior to start-up of each unit.

When a safety related valve is removed from service for maintenance, a clearance sheet is initiated by the shf.ft 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 removing the valve from service and when returning the valve to service. Maintenance procedures require repai. T work to be performed using a maintenance request 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.

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 requirements for technical specification compliance are listed in the maintenance procedures. The procedures contain restora. ion 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. In addition to the above precautions 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.

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I. & E.Bulletin 79-08 Question 7 ,

There are two systems at our facility that are designed te transfer potentially radioactive gases and liquids out of the primary contaiament. These systems are the Drywefl Vent 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 remain closed during all modes of plant operation except when venting is required. An operator is constantly required at the control panel at all times that the 3 valves are open per the " Primary Containment Atmospheric Control System" procedure. These 3 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 either a containment isolation signal, reactor building high radiation signal, or refueling area high radiation signal while the three CAD system valves require an operator to manually close them by procedure when an isolation signal is received. Any time the CAD system valves are opened there is an audible and visible alarm of the abnormal condition in the main control room. There should be no chance of inadvertently opening these valves since the operator must manipulate three specific .ontrols 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 close on either a containment isolation signal, reactor building high radiation signal, or refueling area high radiation signal. To allow for venting post accident the containment isolation signal may be intentionally by passed by operation of an " Override Interlock" switch.

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I. & E.Bulletin 79-08 Question 7 continued The Radwaste Sump subsystems are provided to collect and transport normal liquid leakage from the primary containment to the Padwaste Treatment systems. The pump discharge valves are automatically closed by the containment isolation signals. In addition, the annunciator response procedure requires the operator to manually close these pump discharge valves onca either the containment isolation signal, reactor buidling high radiation signal, or refueling area high radiation signal is receive prior to resetting the isolation signal. The existing annunciator response procedure states that following investigation of the isolation and correction of conditions causing the isolation, the isolation signal can be reset and the pump discharge valves opened. A procedure revision has been 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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E I. & E.Bulletin 79-08 Question 8:

Item 8.a: Operability in redundant system is verified through the inspection of clearance sheets and tags, v,,erbal communications, and visual inspection of System Inoperability Status board. Procedures for maintenance and testing are being reviewed to insure each procedure has a precautionary step to verify through inspection redundant system operability.

This review is to be completed prior to return to startup of the applicable Unit.

Item 8.b: The review and modification of maintenance and test procedures to verify the operability of all safety-related systems when they are_ returned to service following maintenance or testing has been completed and some revisions to procedures found to be necessary. The revision will be completed prior to return to startup of the applice'sie Unit.

Item 8.c: The review of maintenance and test procedures that require safety-related systems to be removed from and returned to service has been completed and found to be adequate.

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. I. & E.Bulletin 79-08 Question 9 Review of the plant prompt reporting procedure, emergency procedure, and the emergency plan indicates the following:

1. Technical Specifications and procedure HNP-450 I' 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 off-site parties by the emergency call list on which the NRC is listed.

The plant emergency procedures will be revised by July 1, 1979 to require plant personnel to establish and maintain an open continuous communication channel with the NRC within one hour after it has been determined that an emergency condition exists. A standing order will be used beginning immediately to place this policy in effect for the interim.

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I. & E.Bulletin 79-08 Ouestion 10:

A review has been conducted by the operations department of HNP 1 and 2 procedures which deal with significant amounts of hydrogen gas that may be generated during a transie,nt or other accident that would either remain inside the primaryDuring the review system or be released to the containment.

of the procedures dealing with the hydrogen concentration within the primary containment several discrepancies have been identified. These discrepancies are Thisnoted in the Post procedure fails Accident Venting procedure for HNP-2.

to refer the operator to the procedure on utilizing the two 100% hydrogen recombiner systems designed to control and recombine the buildup of hydrogen within the containment.

It should be noted that the operator is referred to the procedure for utilizing the hydrogen recombiner syste During the review of the procedures concerning hydrogen no discrepancies were concentration in the primary system, Procedures with discrepancies which were i identified.

during this review will be revised prior to HNP-2 reactor returning to power.

The Unit 1 containment is inerted with nitrogen to preclude the possibility of a hydrogen explosion.

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I. & E.Bulletin 79-08 Question 11 .

As a result of implementing items one thru ten above, it has been determined that no modifications to existing technical specifications are required.

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