ML20003D692
| ML20003D692 | |
| Person / Time | |
|---|---|
| Site: | 05000000 |
| Issue date: | 04/23/1981 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| References | |
| IEIN-81-16, NUDOCS 8103300372 | |
| Download: ML20003D692 (2) | |
Text
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SSINS No.: 6835 Accession No.:
8103300372 IN 81-16 UNITED STATES NUCLEAR REGULATORY COMMISSION OFF'CE OF INSPECTION AND ENFORCEMENT WASHINGTON,.D.C.
- 20555, IE Information Notice No. 81-16 April 23, 1981 Page 1 of 3 o
CONTROL ROD DRIVE SYSTEM MALFUNCTIONS Description of Circumstances:
Continued NRC evaluation of BWR control rod drive (CRD) systems operating experience has identified several operating events which highlight the need for timely operator action if certain CRD system malfunctions occur during specific modes of reactor operation.
In each event, operator action was taken when needed and scram capability was maintained at all times.
However, if timely operator action were not taken (or if other circumstances existed) scram capability might have been degraded.
This notice is provided to inform reactor operators of these events and re-emphasize the reliance on timely operator action (IEB 80-17 Supplement 4, Confirmatory Order dated October 2, 1980, and. Safety Evaluation Report dated December 1, 1980.)
On February 24, 1981, at Brunswick Unit 2, the reactor was manually scrammed from 1.5% power after the group 4 control rods had received three scram signals.
(Group 4 contains 33 control rods and is one of four control rod groups.) The first scram signal for group 4 occurred when surveillance testing caused a trip of RPS "B" channel.
Reactor power decreased from 7% to 1.5% and the RPS "B" trip was reset by the operator.
Another group 4 scram signal, received when an intermediate range monitor (IRM) drifted upscale, was reset by the operator.
A third scram signal, received when the IRM drifted upscale i
i again, caused the operator to initiate a manual scram.
Subsequent investi-gation revealed that a relay contact (K14c) in group 4 RPS "A" had failed open.
Thus, group 4 rods received a scram signal each time RPS "B" was tripped.
We note that rod group scrams of this type have been previously addressed by the NRC (December 1, 1980 Safety Evaluation Report, pages 22-24).
For plants like Brunswick with good communication between the SDV and instrument volume (IV), operator action is not net Jed to maintain scram capability.
- However, for those BWRs with poor communication between the 50V and IV, CRD seal leakage from the scrammed control rods (with open scram outlet valves) could potentially result in filling the SDV before level switches in the IV initiate an automatic scram.
In this case, timely operator action is needed to prevent a temporary loss of scram capability.
Indications are available to alert the operator to i
scrammed CRDs and accumulation of water in the SDV.
These indications include control rod position indication, rod drift indication (with annunciator), high b
bb
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IN 81-16 April 23, 1981 Page 2 of 3 level in SDV (with annunciator), high level in IV (with rod block and annunci-ator), and change in reactor power with attendant indications.
With current equipment and requirements (IEB 80-17 and Confirmatory Orders) we expect the operator would initiate a manual scram while sufficient capacity remains in the SDV.
Two other operating events involved operator action following CRD. system malfunctions not related to SDV performance.
At Brunswick 1 on August 27, 1980, both CRD pumps became inoperable due to low suction pressure caused by high pressure drop across the tuction. filter coincident with low level in the condensate storage tank.
In accordance with procedures the operator manually scrammed the reactor which was in startup, subtritical with some control rods not fully inserted, and at approximately 5 psig pressure.
At Oyster Creek on November 30, 1980, operability of both CRD pumps was challenged by seil water piping leaks on each pump.
This condition was detected and corrected by operators during routine power operation.
There was no direct threat to loss of scram capability in this event since the reactor was pressurized; however, this event is of interest since similar failures affected both pumps.
Scram capability was maintained at all times during both events.
Evaluation of these two events anti possible CRD system failure modes show the need for operator action to raintain scram capability.
Under conditions of reactor low pressure, such as those encountered during startup, control rod scram capability could be lost in an event in which complete failure of CRD hydraulic flow occurred simultaneously with gross leakage from the. scram accumulators. The CRD pumps maintain the pressure on the accumulators and provide motive force for single rod drive operations.
Failure of CRD hydraulic flow can be caused by (1) inoperability of both CRD pumps caused by power failure; (2) plugging of CRD pump suction strainers; (3) lack of an adequate condensate storage tank supply; or (4) other failures in the CRD hydraulic system.
Scram capability under these conditions is designed to be provided by the scram accumulators.
Extensive deterioration of the accumulator charging line check valves could cause a sufficient number of accumulators to discharge and result in a loss of scram capability if the operator does not take appro-priate action.
In the event of such multiple failures, reactor shutdown would have to be accomplished by use of the liquid control system.
This information is provided as a notification of a possibly significant matter that is still under review by the NRC staff.
In case the continuing NRC review finds that specific licensee actions would be appropriate, a bulletin or circular may be issued.
In the interim, we expect that licensees will review this information for applicability to their facilities paying l
particular attention to their operating procedures.
The operating procedures should include specific actions (i.e., initiation of full scram) to be taken by the operator in response to a scram of a portion of tha cont.rni rods.
Procedures should also include the required response (i.e., again to initiate a scram) on recognition of loss of operability of both CPD pumps, especially r
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IN 81-16 April 23, 1981 Page 3 of 3 during the initial stages of cient to accomplish a scram. plant startup when reactor pressure is insuff t-as reflected in BWR Standard Technical Specifications, include testing at least once every eighteen months to check the leak tightness of scram accumulators to hold pressure for at least 20 minutes.
No written response to this IE Information Notice is required.
of the appropriate NRC Regional Offi,ce. additional informat If you need d
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IE Information Notice No. 81-16 April 23, 1981 LISTING OF RECENTLY ISSUED IE INFORMATION NOTICES Information Subject Date Issued To Notice No.
Issued 81-07 Potential Problem with 3/16/81 All power reactor Water-Soluble Purge Dam facilities with an Materials Used During Operating License (OL)
Inert Gas Welding or Cons,truction Permit (CP) i 81-08 Repetitive Failures of
?/20/81 All power reactor Limitorque Operator SMB-4 facilities with an Operating Motor-to-Shaft Key License (OL) or Constructicn Permit (CP) 81-10 Inadvertent Containment 3/25/81 All power reactor facilities Spray Due to Personnel with an Operating License Error (OL) or Construction Permit j
(CP) 81-09 Degradation of Residual 3/26/81 All power reactor facilities Heat Removal (RHR) System with an Operating License i
(OL) or Construction Permit (CP) 81-11 Alternate Rod Insertion 3/30/81 All BWR facilities for BWR Scram Represents with an Operating License a Potential Path for loss of Primary Coolant (OL) or Construction Permit i
(CP) 81 12 Guidance on Order Issued 3/31/81 All BWR facilities January 9, 1981 Regarding with an Operating License Automatic Control Rod (OL) or Construction Insertion on Low Control Permit (CP)
-Air Pressure 81-13 Jammed Source Rack in a 4/14/81 Specified Irradiator Gamma Irradiator licensees 81-14 Potential Overstress of 4/17/81 All power reactor Shafts on Fisher Series facilities with an 9200 Butterfly Valves Operating License (OL) with Expandable T Rings 81-15 Degradation of Automatic 4/22/81 All power reactor ECCS Actuation Capability facilities with an by Isolation of Instrument Operating License (OL)
Lines or Construction. Permit (CP)
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