ML19259C256
| ML19259C256 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/24/1979 |
| From: | Counsil W NORTHEAST UTILITIES |
| To: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| NUDOCS 7906140481 | |
| Download: ML19259C256 (6) | |
Text
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Mr. Boyce H. Grier Director, Region I Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pennsylvania 19406
Dear Sir:
This responds to IE Bulletin 79-08 regarding review of operational errors and system misalignments identified during the Three Mile Island incident.
The paragraph numbers correspond to the bulletin's action items.
1.
The actions requested by this item have been completed. A lesson plan covering the points specified in 1.a and 1.b was prepared and presented to all licensed and unlicensed operators as well as plant management and supervisors with operational responsibilities. Documentation is available.
2.
The primary containment isolation design has been reviewed and it has been confirmed by this review that the required containment isolation does occur in parallel with the automatic initiation of any of the safety injection systems.
The BWR design containment and reactor coolant pressure boundary (RCPB)provides isolation (excluding emergency core cooling and make-up systems).
The isolation occurs upon reactor vessel low water level or high drywell pressure prior to or simultaneous with initiation of emergency core cooling and safety injection systems.
The isolation valves will remain closed until operator action is taken, even if the initiating signal clears.
(A detailed description of the containment and system isolations can be found in the fiillstone 1 FSAR and they are summarized in the Millstone 1 Technical Specifications.)
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Mr. Boyce H. Grier,,
3.
Millstone 1 has an isolation condenser system as an auxiliary heat removal device when the main feed water system is not operable.
The system will automatically initiate when reactor pressure reaches 1085 psig for fifteen seconds.
This system relies upon the natural circulation of steam from the reactor vessel through the isolation condenser and returns condensate to the vessel. Make-up for the shell side is automatic and is supplied from the station fire water or condensate transfer systems. The isolation condenser system may also be manually initiated, by opening one valve either from the control room or locally.
Procedures exist for these evolutions.
For long term operation in this cooling mode, the control rod drive pumps may be used to replenish the coolant lost by nominal primary system boundary leakage.
If the isolation condenser system were not available for cooling, the plant has the ability to maintain cooling using the low pressure coolant injection system (LPCI) after manually depressurizing with the automatic pressure relief (APR) system.
Procedures currently exist for these evolutions.
4.
Reactor vessel water level in the BWR is continuously monitored by seven indicators or recorders for normal, transient and accident conditions. Those monitors used to provide automatic safety equipment initiation are arranged in a redundant array with two instruments in each of two or more independent electronic divisions. Thus, adequate information is prc,vided to automatically initiate safety actions and provide the operator with assurance of the vessel water level at all times.
These water level measurement devices have operated in BWR plants for twenty years. Tests of BWR water level instrumenta-tion under simulated steam and water line breaks F've been conducted showing satisfactory performance.
The range of reactor vessel water level from below the bottom of the active fuel area up to the top of the vessel is covered by a combination of narrow and wide-range instruments.
Level is indicated and recorded in the control room.
A separate set of narrow-range level instrumentation on separate condensing chambers provides reactor level control via the reactor feed water system. This set also indicates and records in the control room.
The safety related systems or functions served by safety related reactor water level instrumentation are:
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Mr..Boyce H. Grier,,
Reactor scram Feed water coolant injection system (FWCI)
Core spray system (CS)
Lowpressurecoolantinjectionsystem(LPCI)
Automatic pressure relief system (APR)
Main steam isolation valve closure Primary containment isolation All systems automatically initiate on low reactor water icvel.
The FWCI syssem will control in level control mode if and when level is restored to the normal operating range.
The core spray and LPCI systems will continue to operate until manually shutdown.
In the unlikely event that vessel level indication were in doubt, tha operators would continue to allow the FWCI, core spray and LPCI systems to operate, overflowing the vessel to the torus via the APR valves.
Existing procedures have been modi.ied to clarify this operation.
5.
a.
The Millstone 1 plant's procedures and training currently are in agreement with the NRC's position on not overriding automatic safety functicns.
b.
Over a dozen other types of instrumentation in the BWR provide the operator with indirect indication of reactor vessel coolant inventory changes and could inform the operator of the need to take corrective actions.
Some of the instrumentation which the operator can use to determine changes in reactor coolant inventory or other abnormal conditions are:
Drywell high pressure Drywell high radioactivity levels Suppression pool high temperature Safety relief valve (SRV) discharge high temperature High feedwater flow rates High main steam flow High containment and equipment area temperatures High differential flow-reactor water clean up system Abnormal reactor pressure High suppression pool water level High drywell and containment sump fill and pumpout rate An example of the use of this additional information by the operator is as follows:
Drywell high pressure is an indirect indication of coolant loss.
Coincident high suppression pool temperature further verifies a loss of reactor coolant.
High SRV discharge temperature would pinpoint loss of coolant via an open valve.
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Mr. Boyce H. Grier,,
Other instrumentation that can signal abnormal plant status, but not necessarily indicative of loss of coolant are:
High neutron flux High process monitor radiation levels Main turbine status instrumentation Abnonnal reactor recirculation flow High electrical current (Amperes) to pump motors 6.
We have reviewed our procedures which relate to periodic surveillance of safety related valve positions.
These procedures have been revised to make them more inclusive. We have also reviewed the administrative procedures governing surveillance testing, maintenance and system / plant startup relative to safety related valve position verification. The existing procedures for surveillance testing are adequate.
The procedures for control of maintenance on safety related equipment have been revised to specifically assure correct positioning of valves which were worked on or were used for isolation purposes.
7.
Systems designed to transfer potentially contaminated radio-active gases and liquids out of the primary containment include:
main steam system, clean up system, drywell equipment and floor drain systems, recirculation loop sample line, drywell and suppression chamber vent systems.
These systems are designed to isolate on either low reactor water level or high drywell pressure. Procedurally, a sample is taken for airborne in the primary containment before venting. The drywell sumps are procedurally operated in manual and thus the possibility of inadvertent pumping is minimal. While no installed radiation monitoring exists for these sump systems, their discharge lines could be monitored with portable instrumentation if the potential for pumping highly contaminated water was present.
The main steam system and clean up systems are equipped with process or area radiation monitors to protect against inadvertent high level releases by these paths.
Each of these protective features is routinely calibrated and/or tested.
8.
a.
The administrative procedures have been revised to specify that prior to removal of safety related systems from service the redundant system will be verified operable.
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Mr. Boyce H. Grier,,
For those equipments which Technical Specifications require specific surveillance, that testing will be completed prior to removing the system from service, b.
Procedures for maintenance and testing of safety related systems have been reviewed and changes have been made to strengthen the requirement to verify operability of safety (related systems prior to taking credit for the system s) to satisfy Technical Specification requirements.
c.
A licensed operator is required to authorize all maintenance, tests, or surveillance which affect plant systems.
Prior to releasing the controlling document, the operator ensures he is aware of the effect of the activity on the system or equipment. Upon completion of the item, the document is returned to the operator for acceptance or for the purpose of returning the system to service. The administrative procedures which control these evolutions provide the required explicit notification of operational personnel whenever a safety related system is removed from and returned to service.
9.
A revision to the administrative procedure on communications and outside assistance has been approved. This revision incorporates the required notifications and establishment of comunication channels requested in the bulletin.
The wording of the reason for innediate notification ("The reactor is not in a controlled or expected condition of operation")
is general in that many different circumstances may or may not fit the definition, depending on who is interpreting the situation. We request more specific guidance on this point in order to provide more explicit instructions to our operators and duty officers.
10.
Hydrogen gas generation is not a problem for Millstone 1.
During nonnal operation, the reactor pressure vessel dome is filled with steam, which flows to the turbine.
During reactor isolation, the dome may be automatically vented through the SRV's to the suppression pool.
In addition, the reactor pressure vessel head has a vent line with a valve remotely operated from the control room.
The primary containment is nitrogen inerted per Technical Specification requirements and thus, hydrogen flammability is precluded.
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Mr. Boyce H. Grier..
Any required Technical Specification changes required as a result of the implementation of the above items will be forwarded within thirty days as requested.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY W. G.
ounsil Vice President cc:
Director, Office of Inspection and Enforcement Division of Reactor Operations Inspection, Washington, D.C.
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