ML20207F605
| ML20207F605 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 12/29/1986 |
| From: | Williams J TOLEDO EDISON CO. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.K.3.05, TASK-TM 1339, GL-83-10, GL-86-05, GL-86-5, TAC-49662, NUDOCS 8701060154 | |
| Download: ML20207F605 (6) | |
Text
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TOLEDO EDISON Docket No. 50-346 JOE WILUAMS. JR.
Sernur Vce FYesdet Puxar (4191249 2300 License No. NPF-3 s,spem Serial No. 1339 December 29, 1986 Director of Nuclear Reactor Regulation Attention:
Mr. John F. Stolz PWR Project Directorate No. 6 Division of PWR Licensing - B United States Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Stolz:
On May 29, 1986 (Log No. 1994), the NRC issued Generic Letter 86-05:
Implementation of TMI Action Item II.K.3.5, " Automatic Trip of Reactor Coolant Pumps." The Generic Letter transmitted the staff's conclusions regarding the B&W Owners Group submittals on reactor coolant pump trip in response to items e and f of Generic Letter 83-10, and provided guidance concerning implementation of the reactor coolant pump (RCP) trip criteria.
Also enclosed in Generic Letter 86-05 was the NRC Safety Evaluation for the B&W Owners Group submittals concerning Reactor Coolant Pump trip criteria.
The NRC has deteunined that the information provided by the 3&W Owners Group in support of loss-of-subcooling RCP trip criteria is acceptable.
However, the information provided did not address plant specific concerns about instrumentation uncertainties, potential RCP problems and operator training and procedures. The attachment provides Toledo Edison's response to the requested information contained in Section IV of the Safety Evalua-tion for Davis-Besse Nuclear Power Station Unit 1.
In accordance with Generic Letter 86-05, Toledo Edison's response to this Generic Letter is not subject to the fee under provisions of 10CFR170.
Therefore, no fee is submitted with this response.
Very truly yours,
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JW:DRB:plf 8701060154 861229 A
CK 05000346 Attachment P
PDR cc: DB-1 Resident Inspector d(a 0 0 THE TOLEDO EDISON COMPANY ED: SON PLAZA 300 MADISON AVENUE TOLEDO. CH'O 43652
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' Docket N3. 50'-346
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License No. NPF "F
' Serial'No. 1339 i
O,; December?25, 1986 g,iAttschmentf-y s n
,fj Response to Generic Letter 86-05;
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Implementation of TMI Action Item II.K.3.5, g
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<dAutomatic Trip of Reactor Coolant Pumps"
.For Davis-Besse Nuclear Power Station Unit 1 A.
Determination d RCP Trip Criteria 1.
Identify the instrumentation to be used to determine the RCP trip setpoint, including the degree of redundancy of each parameter signal needed for the criterion chosen.
Response
Reactor Coolant Pump (RCP) trip is based on the degree of subcooling.raargin in the Reactor Coolant System (RCS). Several methods of determining RCS subcooling margin exist. The primary method is through use of the T meters in the control room.
There are twe independent T eters, each of which has multiple, sat operator.selectable RCS temperature inputs. The inputs to the T
meters are the RCS wide range pressure instruments, located onkteRCShotlegs,viatheSafetyFeaturesActuationSystem (SFAS), the wide range hot leg temperature instruments, and sixteen selected core exit thermocouples (T/C). The normal inputs to.the meters, on which the RCP trip criteria is based, are nuclear safety grade and environmentally qualified. The selected incore T/C system is planned to be upgraded to nuclear safety grade and environmentally qualified.
In addition to the T meters, the operators can determine RCS subcoolingfromthe$$fetyParameterDisplaySystem(SFDS) system, the plant process computer, and multiple RCS pressure and temperature indicators in conjunction with a saturation curve (which is an approved figure in the Emergency Procedure EP 1202 01, "RPS, SFAS, SFRCS Trip or SG Tube Rupture").
Therefore multiple, redandant methods of determining the need for RCP trip exist.
2.
Identify the instrumentation uncertainties for both normal and adverse containment conditions. Describe the basis for the selection of the adverse containment parameters. Address, as appropriate, local conditions such as fluid jets or pipe whip which might influence the instrumentation reliability.
Response
The Emergency Procedure EP 1202.01 requires RCP trip when RCS subcooling, as indicated on the T meters, is less than 20*F subcooled.
Based on calculatioa N Stv-64.00-001, Revision 1, this margin ensures that the RCS is utill, subcooled when RCP trip is perforn:ed. The margin, including uncertainties, ranges
~Dockst No. 50-346 License No. NPF-3
- Serial No. 1339
. December 29, 1986 Attachment-
'from a minimum margin of 1.6*F at 470 psig RCS' pressure to a maximum margin of 11.4'F at 2470 psig RCS pressure. The 20*F subcooling margin is alco sufficiently low to ensure that-undesirable.RCP trip does not occur during non-LOCA events such asLa steam generator tube rupture. As noted in the Safety Evaluation included in the NRC's May 29, 1986 Generic Letter 86-05, the FSAR assumes the worst case RCP status (off or on) in non-LOCA accidents where subcooling could approach the 20'F subcooling margin for RCP trips.
In all cases safe operation of the plant has been demonstrated.
It is not necessary to include consideration of adverse contain-ment conditions or local conditions that might affect instru-mentation reliability even though the normal input instrumentation is environmentally qualified for all design basis accident environments and the selected core exit T/C's are planned for upgrading to be fully qualified.
In the time frame in which RCP trip is required,_and the spectrum of breaks for I
which such trip is required, significant adverse containment conditions will not have developed inside containment. There-fore, only normal containment condition instrument uncertainties need.be considered. Should a pipe break occur that could expose the instruments to fluid jets or pipe whip, it could only be in the line which the instrument is monitoring, as there is suffi-cient separation between instrument locations. This would make the measurement meaningless, which would be obvious to operators in the control room due to the redundancy of equipment. There-fore, use of normal containment condition instrument inaccura-cies is justified.
3.
In addressing the selection of the criterion, consideration of uncertainties associated with the BWOG or plant specific sup-plied analyses values must be provided. These uncertainties include both uncertainties in the computer program results and uncertainties resulting from plant specific features not repre-sentative of the BWOG generic data group.
Response
The B&W analysis summarized in " Analytical Justification for the Treatment of Reactor Coolant Pumps during Accident Conditions" (Reference 1 of Generic Letter 86-05's Safety Evaluation) is inherently conservative; therefore, it can be concluded that the results represent a conservative estimate of plant behavior.
Due to the built-in conservatisms, we hold that the final numbers represent worst-case values and actual transients will be less severe. This is typical of all safety related calcula-tions, so that determination of calculation uncertainties are not necessary or meaningful. The generic model used in the BWOG analysis has been reviewed to ensure it envelopes Davis-Besse specifics, and since it does, it is concluded that its results represent worst case behavior.
- >J Docket Ns. 50-346 License No. NPF Serial No. 1339 December 29,.1986~
Attachment B.
Potential Reactor Coolant Pump Problems 1.
Assure that containment isolation, including inadvertent isola-tion,-will not cause problems-if it occurs for non-LOCA tran-sients and accidents.
a.
Demonstrate that if water services needed for RCP opera-tions are terminated, then they can be restored fast enough.
to prevent-seal damage or failure once a non-LOCA situation is confirmed.
b.
Confirm that containment isolation with continued pump operation will not lead to seal or pump damage or failure.
Response
Containment isolation is initiated by the Safety Features Actuation System (SFAS) which is divided into five incident levels and which cause automatic actions based on the types of accidents that could cause that incident level's setpoint to be reached. Levels 1, 2, and 5 do not affect the~RCP's or their auxiliary systems. A Level 3 signal, caused by a high contain-ment vessel-(CV) pressure (set at less than or equal to 18.4 psia CV pressure) or a low-low RCS pressure (set at greater than or equal to 420 psig RCS pressure), will cause the RCP seal injec-
-tion and seal return lines to be isolated. A Level 4 signal, which occurs only on a CV high-high pressure condition (set at less than or equal to 38.4 psia CV pressure), will cause compo-nent cooling water to the RCP seal cooler and bearing cooler to be isolated.
- Davis-Besse has addressed loss of seal injection and seal return and loss of. component cooling water (CCW) to RCP's, initiated by any cause, in Abnormal Procedure AB 1203.21, " Reactor Coolant Pump and Motor Abnormal Operation". This procedure states,
. based on Vendor Technical Manuals, that RCP operation can continue essentially indefinitely without seal injection and seal return, as long as CCW is available to the RCP. Since there are no non-LOCA events which can cause a Level 4 actuation of SFAS, CCW would be available to the RCP's at all times for non-LOCA events and tripping of the RCP's will be tripped only if a loss of subcooling margin occurs. During LOCA's which cause a Level 4 isolation, loss of RCS subcooling margin will also occur, so that RCP's would be tripped anyway.
Consequent-ly, use of subcooling margin as the sole RCP trip criteria
. parameter is justified.
Dock:t No. 50-346 License No. NPF-3 Serial No. 1339 December 29,.1986 Attachment-Response to inadvertent isolation of seal injection and seal return is adequately addressed by AB 1203.21.
If RCP's could not be restarted after shutdown, the plant can be cooled down by natural circulation in the RCS. Adequate instrumentation and controls exist in the control room for operators to promptly identify and safely control inadvertent loss of CCW and seal injection / seal return to RCP's.
During events that require controlled cooldown of the plant, containment isolation of RCP seal injection and seal return is prevented by blocking the appropriate SFAS signal just prior to reaching the respective set point.
2.
Identify the components required to trip the RCP's, including relays, power supplies and breakers. Assure that RCP trip, when
' determined to be necessary, will occur.
If necessary, as a result of the location of any critical component, include the effects of adverse containment conditions on RCP trip reliabil-ity. Describe the basis for the adverse containment parameters selected.
Response
There are four methods of tripping the RCP's, all of which are based on manual action. Two of the four can be accomplished in the control room. The other methods are performed at the RCP switchgear.
The primary method of tripping the RCP's is from the individual RCP control switches in the control room. Each pump's control logic is the same. To trip the pump breaker, the pump control switch in the control room is placed to "stop".
This closes contacts in the switch which complete the circuit to energize the trip coil in the breaker, thereby tripping the breaker. There is also a control switch at the breaker which can trip the breaker in a similar method. These control switches receive 125 VDC power from the switchgear DC busses. These busses are supplied with redundant, diverse sources to ensure control power is of the highest reliability.
The RCP's can also be tripped manually at the individual pump breaker's switchgear, located in the high voltage switchgear room, by use of a manual trip mechanism. This requires no electrical power.
Another method of stopping the RCP's is to trip the supply breakers to the "A" and "B" 13.8 KV busses using controls in the control room. These busses supply power to the RCP's via breakers HA01 (RCP 1-L-2), HA03 (RCP 1-1-1), HB01 (RCP (1-2-2),
and HB03 (RCP 1-1-2).
The "A" and "B" bus supply breakers receive DC control power from different sources than the RCP breaker control power supplies. This provides an independent, diverse method of stopping RCPs.
4 s
Docket Ns. 50-346 License No. NPF-3 Serial No. 1339 December.29, 1986 Attachment It is not-necessary to consider adverse environmental effects on l.
'RCP trip capability since the only portion of the system exposed l-to harsh environments is the RCP motor and its power cables.
The. remainder of.the system is located in mild environment areas.
There are adequate controls and indication available to the operators in the control room to allow them to determine RCP status and to stop the RCP's as described above.
Davis-Besse performs routine maintenance and' surveillance activities on the equipment required for RCP trip to ensure reliability.
It should be noted that the RCP breakers have never failed to trip at Davis-Besse, which illustrates their reliability.
C.
Operator Training and Procedures (RCP Trip)
In response to NRC questions concerning the identification and management of primary system voids, the BWOG response identified potential changes to the ATOG procedures to incorporate proposed detection and management schemes.
Each licensee should endorse this program as described, and provide an implementation schedule for the revised ATOG.
If a licensee does not endorse the provided proposal, then a suitable alternate proposal must be provided including an implementation schedule.
Response
Toledo Edison, by prior commitment to the NRC, has endorsed the BWOG response to the RCP trip criteria. The Davis-Besse ATOG and EP 1202.01 have been revised to incorporate the criteria; and, operators have been trained on this criteria. Therefore, no implementation schedule is required.
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