ML19270G077
| ML19270G077 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 04/25/1979 |
| From: | Andognini G BOSTON EDISON CO. |
| To: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| 79-79, NUDOCS 7906020200 | |
| Download: ML19270G077 (15) | |
Text
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4 peg 9I BOSTON EOisON COMPANY GENERAL oFFICCs 800 DovLaTON SYNEET BOSTON. M ASS ACMuscTTs 0 219 9 BECo Ltr. #79-79
- a. cAa6 A=oaamiu, MANAGER wuct A= onc ATious or,A.Twe"7 April 25, 1979 Mr. Boyce Grier Office of Inspection and Enforcement Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pa 19406 License No. DPR-35 Docket Nc. 50-293 Response To IE Bulletin 79-08
Dear Sir:
The attached enclosure is submitted in response to the subject bulletin which was received at Pilgrim Nuclear Power Station on April 15, 1979, via telecopier.
This enclosure addresses Items 1 through 10 only, Item 11 will be responded to within the required 30 days.
We trust that this information is responsive to your requests in the subject bulletin.
If we can be of any further assistance, please contact us at your convenience.
Very truly yours,
- t 4
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2269 216 enc 1.
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Director Office of Inspection and Enforcement Division of Reactor Operations Inspection Washington, DC 20555 79060209 %
ENCLOSURE Responses To I.E.Bulletin 79-08 Item 1.
Review the description of circumstaneca described in Enclosure 1 of IE Bulletin 79-05 and the preliminary chronology of the TMI-2 3/28/79 accident included in Enclosure 1 to IE Bulletin 79-05A.
a.
This review should be directed toward understanding:
(1) the extreme seriousness and consequences of the simultaneous blocking of both trains of a safety system at the Three Mile Island Unit 2 plant and other actions taken during the early phases of the ac-cident; (2) the apparent operational errors which led to the eventual core damage; and (3) the necessity to systematically analyze plant conditions'and parameters and take appropriate corrective action.
b.
Operational personnel should be instructed to (1) not override automatic action of engineered safety features unless continued operation of engineered safety features will result in unsafe plant conditions (see Section 5a of this buttetin); and (2) not make operational decisions based solely on a single plant parame-ter indication when one or more confirmatory indications are available.
c.
All licensed operators and plant management and supervisors with operational responsibilities shall participate in this review and such participation shall be documented in plant records.
Response
The Pilgrim Station Training Department is in the process of establishing a trainine program to ensure that the respective Bulletins are reviewed in detail with all licensed operators and plant management.
This review will be conducted under formal classroom conditions and cover as a minimum the areas identified in a. and b. above.
This review will be documented in the training records.
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Item 2.
Review the containment isolation initiation design and procedures, and prepare and implement all changes necessary to initiate containment isolation, whether manual or automatic, of all lines whose isolation does not degrade needed safety features or cooling capability, upon automatic initiation of safety injection.
Response
A review of the containment isolation which occurs upon low-low water level or high drywell pressure prior to or simultaneous with initiation of safety injection was conducted to ensure isolation of all lines without degrading needed safety features or cooling capability.
It was concluded that the automatic containment isolation criteria was satisfied for all penetrations except the following:
1.
Reactor Building Component Cooling Water 2.
Instrument Air / Nitrogen Supply 3.
Instrumentation Lines 4.
Residual Heat Removal to Spent Fuel Pool Demineralizer 5.
Hydrogen Analyzer 6.
Reactor Building to Torus Vacuum Breakers 7.
Torus Make-up 8.
Main Steam Isolation Valves 9.
Main Steam Drain Isolation Valves
- 10. Reactor Water Sample Isolation Valves Items 3 and 5 provide intelligence necessary for the operator and initiating circuits to base actions upon and cannot be isolated.
These instrument lines are further provided with flow limiting orifices, excess flow check valves and manual isolation valves should isolation be necessary.
Item 6 is required by design to function upon establishment of a vacuum in the Torus to allow equalization with the Reactor Building.
The Torus is not designed for operation with a negative pressure.
Items i and 7, if in service at the time of receipt of the containment isolation signal will require manual isolation and procedures will be revised accordingly.
Items 1 and 2 will require manual isolation once it has been determined that:
(a) An off-normal condition exists that has resulted in a containment isolation and 2269 218
(b) the primary containment integrity has been jeopardized via these systems.
Items 8, 9 and 10 are the Group 1 isolation valves and do not automatically close on high drywell pressure.
It should be noted that primary containment isolation initiation provides necessary isolation of the containment in the event of accidents or similar critical conditions when the free release of containment atmosphere cannot be permitted.
It is neither necessary nor desirable that every isolation valve close simultaneously with a common isolation signal.
For example, if a process pipe were to rupture in the drywell, it would be important to close all lines which are open to the drywell, and some effluent process lines such as the main steam lines.
How-ever, under these conditions, it is essential that containment and core cooling systems be operable.
For this reason, specific signals are utilized for isolation of the various process and safeguards systems.
The high drywell pressure signal is not utilized to isolate the Group 1 valves in order to maintain the normal heat sink available until a low-low level signal causes isolation.
This allows for an orderly and controlled shutdown in the event of a small break inside containment.
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s e
Item 3.
Describe the actions, both automatic and manual, necessary for proper functioning of the auxiliary heat removal systems (e.g., RCIC) that ara used when the main feedwater system is not operable.
For any manual action necessary, describe in suumary form the procedure by which this action is taken in a timely sense.
Response
The High Pressurt Coolant Injection System and the Reactor Core Isolation cooling System function as auxiliary heat removal systems when the main feedwater system is inoperable.
Both automatic and manual actions are summarized below:
I.
High Pressure Coolant Injection System A.
Automatic Actions HPCI system operation is initiated automatically by signals of either low-low water level in the vessel or high pressure in the drywell.
The system alignment is always maintained in standby status ready to receive an auto initiation signal.
When an initiation signal occurs the operator checks the following:
steam supply valve, and injection valves open and ficw being delivered to reactor vessel.
B.
Manual Actions 1.
HPCI Injection to the Vessel If it becomes obvious that a low-low water level condition is about to occur, the HPCI can be initiated manually as follows:
a.
Open pump injection valve b.
Start gland seal exhauster blower and auxiliary oil pump c.
Open steam inlet valve d.
Monitor system operation as with auto initiation 2.
Test Mode Operation Decay heat in the reactor after a shutdown results initially in about six percent of rated heat generation.
Af ter twenty minutes this heat generation has been reduced to about two percent rated.
The main condenser is normally the heat sink for the steam pro-duced by the decay heat.
However, if the reactor is isolated the heat must be removed by other means; the use of the relief valves is necessary for the first twenty minutes.
When the decay heat has been reduced to about 2% of rated, the HPCI can then be used to cool and depressurize the reactor to less than 100 psig. When this technique is required the HPCI is initiated manually with the pump flow routed through the test line return to the condensate storage tank.
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s Fast start instructions for both the injection mode and the test mode operation of the HPCI are inscribed on bakelite tags and permanently affixed to the HPCI Panel in the Control Room.
II.
Reactor Core Isolation Cooling The RCIC is designed to provide makeup to the nuclear system as part of the planned operation for periods when the normal heat sink is unavailable, A.
Automatic Actions RCIC system operation is initiated automatically by a signal of low-low water level in the vessel.
When an initiation signal occurs, the operator checks the following:
The pump injection valve is open and RCICS flow levels of f and stays as 400 gpm, indicating that the flow controller has control of turbine speed.
B.
Manual Action 1.
RCIC Injection to the Vessel If it becomes obvious that a low-low water level condition is about to occur, the RCIC can be initiated manually, a.
Open cooling water supply valve b.
Start gland seal vacuum pump c.
Open pump discharge valve d.
Open turbine steam supply valve e.
Observe that the flow to the vessel increase to 400 gpm and stabilizes.
Following a loss of feedwater and reactor scram, a low-low reactor water level signal will automatically initiate main steam line isolation valve closure. At the same time these signals will put the HPCI and RCIC Systems into the reactor coolant make-up inj ection mode.
These systems will continue to inject water into the vessel until a high water level signal automatically trips the systems.
Following a high reactor water level trip, the HPCI System will automatically re-initiate when reactor water level decreases to low-low water level.
The RCIC System must be manually reset by the operator before it will automatically re-initiate after a high water level trip.
Following the initiation of HPCI and RCIC, the system will function until they are caused to isolate on high reactor vessel level.
At this time, the operator has sufficient redundant and diverse instrumentation to allow a determination if the rate of coolant exiting the reactor will again cause HPCI to initiate on low-low level.
If so, the HPCI should be manually restarted when the high level isolation clears and the Flow Indicator Controller adjusted so that the input flow can maintain a stable water level.
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s.
The Emergency or Off Normal Station Operations Manual Procedures that could result in a loss of feedwater are being reviewed and revised, as necessary, to assure that instructions for manual initiation of both HPCI and RCIC are provided as follows:
A.
Verify that all automatic actions occur as expected.
B.
Verify the Reactor scram and monitor the parameters necessary to assure safe shutdown.
C.
Operate a relief valve for approximately 1 minute.
NOTE:
One minute operation of a relief valve is equal to approximately one foot of water in the vessel capacity change.
D.
After the relief valve is closed, put the HPCI in service in the test mode.
This will control the Reactor pressure.
E.
Use the RCIC for make-up.
F.
Run two RHR pumps in the torus cooling mode.
G.
Begin a controlled cooldown using the HPCI in the test mode.
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4 Item 4.
Describe all uses and types of vessel level indication for both auto-matic and manual initiation of safety systems.
Describe other redundant instrumentation which the operator might have to give the same informa-tion regarding plant status.
Instruct operators to utilize other available information to initiate safety systems.
Response
The attached table is a complete listing of all level instrumentation and includes use types and location in the plant.
These instruments marked with an asterisk provides redundant instrumentation to the Control Room Operator.
The Training Department is preparing an outline which will be given to the Watch Engineers who will conduct the lectures on shift.
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i VE*SEL LEVEL IC ICAft0M INSTRLHENT USE TTPE LOCATION LIS 71A Initiate / Trip HPCI. RCIC Level 2205 RK 725 Pereirsives to ADS. C.S.. LPCI indicating 2:06 RK 72C tattiate Standby Diesel Generators Switches 2205 RK 72D Indicates level between 432.5 6 532.5 inches 2206 RK El. 51' Reactor 81ds.
LIS 57A Reactor Protection System Level 2205 RK 575 Primary containment Isolation Indicating 2205 KK 58A Indicates level between 432.5 & $32.5 Switches 2206 KK 58s 2:06 RK E1. 51' Reactor 514a.
LITS 59A Turbine Trip Level indic-2205 KK Indicates level,etween 432.5 & 532.5 inches ating trans*
E1. 51' mitter/ switch Reactne Bldg.
LZ 100A Indicates level between 432.5 & 532.5 inches meter 905 pn1 *
(Control Room)
LITS 598 Turbine Trip Level indic.
2206 KK Indicates level between 432.5 & 532.5 inches ating stans-E1. 51' eitter/ switch Reactor side.
Lt 1005 Indicates level between 432.5 6 532.5 inches meter 905 pal *
(Control Room)
LT 646A Feedwater Control system input Level 2205 KK Transmitter E1. 51' Resetor Bldt.
LI 29A Indicates level between 482.5 & 532.5 inches meter 905 pal
- Control Room LT 6465 Feedwater Control System input Level 2206 RK Transmitter E1. 51' Reactor 81de.
L1 298 Indicates level between 482.5 6 532.5 inches meter 905 pal
- Control noos LR 26A Records level of indication.either (LI 29A or L1298)
Recorder
! 905 pn1
- l Control Room LITS 73A Permissive for containment spray Level indic.
2251 RK Indicates level between 205 & 505 inches ating trans-El. 23' Reactor B1dg.
sit ter/ switch l l
1 LZ 10&A Indicates level between 205 & 505 inches l meter j 901 pal
- Control Room LITS 738 Permissive.ar containment spray Level indic-2252 RK Indicates level between 205 & 505 inches ating trans- ! Et, 2 3' aitter/ switch Reactor Bldg.
L1 1065 Indicates level betwaen 205 & 505 inches l meter 903 pal
- Centrol Room LT 79 Transmit signals for level between 355 & 455 inches Level trans- ' 2232 RK I
E1. 23' mitter i Reactor Bldg.
LR 28 Records level between 355 & 455 inches Recorder l 905 pal *
' contrni Rocr LT 61 Transmits signale for level between 397 & 797 inches Level trana ' 2205 RK l
sitter
, tl. 51'. Reactor Bids.
l 904 pal
- Lt 101 Indicates level between 397 & 797 inches meter I Control Room e available to control room operators
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2269.,224
Item 5.
Review the action directed by the operating procedures a'id training instructions to er.sure that:
a.
Operators do not override automatic actions of engineered safety features, unless continued operation of engineered safety features will result in unsafe plant conditions (e.g. vessel integrity).
b.
Operators are provided additional information and instructions to not rely upon vessel level indication alone for manual actions, but to also examine other plant parameter indications in evaluating plant conditions.
Response
A review of PNPS Emergency and System Operating procedures was conducted and, although existing procedures do address concerns about overriding automatic actions, Procedure Changes are bcing initiated to provide more explicit direction and to emphasize the necessity for verifying conditions with redundant and diverse instrumentation prior to taking actions relative to vessel level.
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Item 6.
Review all safety-related valve positions, positioning requirements and positive controls to assure that valves remain positioned (open or closed) in a manner to ensure the proper operation of engineered safety features.
Also review related procedures.
Such as those for mainten-ance, testing, plant and system startup, and supervisory periodic (e.g.,
daily /shif t checks) surveillance to ensure that such valves are returned to their correct positions following necessary manipulations and are maintained in their proper positions during all operational modes.
Response
Prior to a plant start-up following an extended outage all safety system valve line-ups are verified.
The station procedure to verify locked open/
closed valves in safety systems is performed semi-monthly.
In addition to this surveillance, the Operating Supervisor will verify daily on the 4-12 watch that the HPCI, RCIC, CS and LPCI systems are lined up in accordance with operating procedures for stand-by conditions and the Watch Engineer will perform the same verification for the RBCCW, SSW DG &
SBLC systems.
The PNPS administrative procedure for Maintenance Request requires all that isolations for maintenance activities be identified on the Maintenance Re-quest and the Operating Supervisor must review the MR and approve system status prior to initiation of the activity.
The daily supervisory checks will identify any off-normal conditions and provide documentation to all operations personnel.
Upon completion of the respective work activity the SUt cannot be signed off as complete until such time that the required surveillance tests are perforced and the system declared operable.
All PNPS surveillance procedures require verification that systems have been returned to normal operation with proper valve line-up.
The Watch Engineer must review and concur with all surveillances before signing off as completed.
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Item 7.
Review your operating modes and procedures for all systems designed to transfer potentially radioactive gases and liquids out of the primary containment to assure that undesired pumping, venting or other release of radioactive liquids and gases will not occur inadvertently.
In particular, ensure that such an occurrence would not be caused by the resetting of engineered safety features instrumentation. List all such systems and indicate:
Whether interlocks exist to prevent transfer when high radiation indication a.
exists, and b.
Whether such systems are isolated by the containment isolation signal.
The basis on which continued operability of the above features is assured.
c.
Response
There are two separate systems for transfering potentially radioactive liquids from primary containment and two separate systems for transfering potentially radioactive gases.
The two liquid systems are the containment sumps and RHR to radwaste and the two gas systems are Containment Atmospheric Control and Stand-by Gas Treatment.
The Containment Sumps, RHR to radwaste and Containment Atmospheric Control Systems isolate on reactor vessel low water level and/or containment high pressure.
The containment sump isolation and RHR to radwaste isolation cannot be overridden as long as low-low water or high drywell pressure exist.
When the isolation is reset the RHR to radwaste will remain closed; however, if the sump isolation valves are in the auto position they will respond to a high level condition and enable water to be pumped to radwaste from the containment.
Station procedures require that these valves are maintained closed and that the sumps are pumped on high level alarm. To preclude the switch from being inadvertently placed in the auto position this function will be removed from the valve permanently.
The atmospheric Control System has e erride capability via a key lock switch to allow venting the drywell via the SBTS when a primary containment isola-tion signal exists.
This function is procedurely controlled and requires obtaining the key from the Watch Engineer There are no interlocks that prevent transfer when high radiation exists either in the system or in the containment.
Presently plant conditions, local samples and local radiation readings would be used to determine if a transfer should be either initiated, con-tinued or terminated.
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Item 8.
Rev,iew and modify as necessary your maintenance and test procedures to ensure that they require:
Verification, by test or inspection, of the operability of a.
redundant safety-related systems prior to the removal of any safety-related system from service.
b.
Verification of the operability of all safety-related systems when they are returned to service following maintenance or testing.
Explicit notification of involved reactor operational personnel c.
whenever a safety-related system is removed from and returned to service.
Response
The station procedure for maintenance requests was reviewed and verified to contain adequate direction to ensure Items 8a, b & c are performed.
In addition to these actions, all out of service conditions for ECCS are identified in the Operations Log and all shift operations personnel are re-quired to review the log.
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Item 9.
Review your prompt reporting procedures for NRC notification to assure that NRC is notified within one hour of the time the reactor is not in a controlled or expected condition of operation.
Further, at that time an open continuous communication channel shall be established and main-tained with NRC.
Response
A review of the Emergency and Reporting procedure was made and it was deter-mined that changes are necessary to meet the above requirements.
The pro-cedures which address conditions of limited control or unexpected conditions of operation are being revised to include NRC reporting requirements among the subsequent operator actions.
In addition, the Failure and Malfunction Reports Procedure, is being revised to include requirements to notify the NRC anytime conditions addressed above exist.
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Item 10.
Review operating modes and procedures to deal with significant amounts of hydrogen gas that may be generated during a transient or other ac-cident that would either remain inside the primary system or be released to the containment.
Response
C The Emergency Procedure for Post Accident Venting was reviewed and determined to be adequate in regards to dealing with hydrogen releases to containment.
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