ML20133C592
| ML20133C592 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 10/01/1985 |
| From: | Musolf D NORTHERN STATES POWER CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.K.3.05, TASK-TM GL-85-12, TAC-49663, TAC-49664, NUDOCS 8510070449 | |
| Download: ML20133C592 (6) | |
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Northem States Power Company 414 Ncollet Mall Minneapohs. Minnesota $5401 Telephone (612) 330-5500 October 1, 1985 Director Office of Nuclear Reactor Regulation U S Nuclear Regulatory Commission Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Response to Generic Letter 85-12, Implementation of TMI Action Item II.K.3.5, Automatic Trip of Reactor Coolant Pumps This letter is provided in response to Generic Letter 85-12, Implementation of TMI Action Item II.K.3.5,
" Automatic Trip of Reactor Coolant Pumps" Generic Letter 85-12 asked NSP to provide the information requested in Section IV of the Office of Nuclear Reactor Regulation's safety evaluation of the Westinghouse Owners Group submittals on reactor coolant pump trip.
The response to that reqt.est is provided as an attachment.
Please contact us if you have any questions related to the information we have provided.
O b.
David Musolf Manager - Nuclear Suppor Services DMM/EFE c: Regional Administrator-III, NRC NRR Project Manager, NBC Resident Inspector, NRC MPCA Attn:
F W Ferman G Charnoff Attachment f.f 0t i
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8510070449 Bh00002B2 PDR ADOCK O PDR P
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RESPONSE TO GENERIC LETTER No. 85-12 A.
Determination of RCP Trip Criteria 1.
Identify the instrumentation to be used to determine the RCP trip set point, including the degree of redundancy of each parameter signal needed for the criterion chosen.
Response
The RCP trip criteria chosen for Prairie Island is based on Reactor Coolant System (RCS) pressure.
Two channels of wide range RCS pressure are available to the operator for monitoring RCS pressure. Readout of pressure channels are available on the main control board via the subcooling monitor display.
Additionally, dedicated RCS wide range pressure indicators for each channel will be installed in the near future.
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
a.
Instrument uncertainty under normal containment conditions is approximately 43 psi.
This value is based on a square root sum of the square combination of transmitter, current repeater and indicator errors.
b.
Instrument uncertainty under adverse containment conditions is approximately 283 psi. Assuming the adverse containment condition error is in addition to normal instrument errors, the total pressure error becomes 326 psi.
Adverse containment conditions (radiation and steam) were applied only to the transmitter because only the transmitter is exposed to the containment environment.
Adverse containment parameters used for the calculation of the instrument uncertainties are the same as those developed for Prairie Island for demonstrating compliance with 10 CFR 50.49, " Environmental Qualification of Electric Equipment Important to Safety For Nuclear Power Plants".
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3.
In addressing the selection of the criterion, consideration of uncertainties associated with the WOG supplied analyses values must be provided. These uncertainties include both uncertainties in the computer program results and uncertainties resulting from plant specific features not representative of the generic data group.
EIf a licensee determines that the WOG alternative criteria are marginal for preventing unneeded RCP trip, it is recommended that a more discriminating plant-specific procedure be developed.
For example, use of the NRC-required inadequate-core-cooling instrumentation may be useful to indicate the need for RCP trip.
Licensees should take credit for all equipment (instrumentation) available to the operators for which the licensee has sufficient confidence that it will be operable during the expected conditions.
Response
The Reactor Coolant Pump trip criteria selected for use at Prairie Island is a RCS pressure of 1200 psig (normal contain-ment conditions) and 1500 psig (adverse containment conditions).
Adverse containment conditions are defined in the procedures as acontainmentpressuregreaterghan5psigorcontainment radiation level greater than 10 R/hr. The values of RCS pressure were calculeted using the Westinghouse Owners' Group methodology and Prairie Island specific design parameters.
Westinghouse analysis shows that the minimum RCS pressure due to a Steam Generator Tube Rupture or other non-LOCA events is 1348 psig. The calculated trip pressure of 1200 psig therefore affords adequate assurance the pumps will not be tripped for non-LOCA events.
B.
Potential Reactor Coolant Pump Problems 1.
Assure that containment isolation, including inadvertent isolation, will not cause problems if it occurs for non-LOCA transients and accidents.
a.
Demonstrate that, if water services needed for RCP opera-tions are terminated, they can be restored fast enough once a non-LOCA situation is confirmed to prevent seal damage or failure.
b.
Confirm that containment isolation with continued pump operation will not lead to seal or pump damage or failure.
Response
At Prairie Island, containment isolation, including inadvertent isolation, does not isolate either component cooling water or seal injection water to the Reactor Coolant Pumps. Therefore, water services needed for RCP operations are not terminated for either LOCA or non-LOCA situations.
J 2.
Identify the components required to trip the RCPs, 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 reliability.
Describe the basis for the adverse containment parameters selected.
Response
Components required to trip a Reactor Coolant Pump at Prairie Island include:
- 1) Control Switch - This switch is located in the main control room. One switch is provided for each pump.
- 2) Breaker - Each RCP breaker is located in the turbine building in areas not affected by any adverse containment or auxiliary building environments.
3)
D.C. Control Power - D.C. control power is provided from the station D.C. power system (batteries or battery charger).
This equipment is also not affected by any adverse contain-ment or auxiliary building environments.
C.
Operator Training and Procedures (RCP Trip) 1.
Describe the operator training program for RCP trip. Include the general philosophy regarding the need to trip pumps versus the desire to keep pumps running.
-Response:
Criteria and philosophy of tripping reactor coolant pumps is covered in three (3) separate parts of License Operator Training and License Requalification Training. They are:
reactor coolant pump system training, emergency operating procedures training and simulator training. The reactor coolant pump system lesson plan includes a discussion of how reactor,
coolant pump trip criteria was selected. Advantages and dis-advantages of tripping the reactor coolant pumps are discussed.
Emergency operating procedures training includes a review and discussion of the basis document of each step in the emergency operating procedures.
Simulator training involves practice in recognizing that RCP trip criteria has been reached and prac-tice_in tripping the RCPs. Training on when the RCP criteria applies is also covered during simulator training.
The general philosophy taught is that it is desirable to keep the RCPs running during accident conditions unless the trip criteria is met.
For some larger break LOCAs fuel centerline temperatures may exceed USAR allowable values if the RCPs are lost later in the accident; therefore, under certain conditions it is better to trip the RCPs.
In general, strict adherence to the emergency operations procedures is emphasized. However, we do teach that if the RCS is being depressurized by operator action and in a controllable manner then the RCPs should not be
. tripped even though RCS pressure has been reduced to less than 1200 psig.
2.
Identify those procedures which include RCP trip related operations:
(a) RCP trip using WOG alternate criteria (b) RCP restart (c) Decay heat removal by natural circulation (d) Primary system void removal (e) Use of' steam generators with and without RCPs operating (f) RCP trip for other reasons
Response
a.
RCP trip using WOG alternate criteria E-0 Reactor Trip of Safety Injection E-1 Loss of Reactor of Secondary Coolant E-3 Steam Generator Tube Rupture ECA-2.1 Uncontrolled Depressurization of All Steam Generators b.
RCP restart ES-0.1 Reactor Trip Recovery ES-0.2 SI Termination ES-0.3 Natural Circulation Cooldown ES-0.4 Natural Circulation Cooldown (with RVLIS)
ES-0.5 Natural Circulation Cooldown (without RVLIS)
ES-1,1 Post LOCA Cooldown and Depressurization E-3 Steam Generator Tube Rupture e. --
ECA-2.1 Uncontrolled Depressurization of All Steam Generators ECA-3.1 SGTR With Loss of Reactor Coolant:
i Subcooled Recovery ECA-3.2 SGTR With Loss of Reactor Coolant:
Saturated Recovery ECA-3.3 SGTR Without Pressurizer Pressure Control FR-P.1 Response to Imminent Pressurized Thermal Shock FR-I.3 Response to Voids in Reactor Vessel c.
Decay heat removal by natural circulation E-3 Steam Generator Tube Rupture ES-0.1 Reactor Trip Recovery ES-0.2 SI Termination ES-0.3 Natural Circulation Cooldown ES-0.4 Natural Circulation Cooldown (with RVLIS)
ES-0.5 Natural Circulation Cooldown (without RVLIS)
ES-1.1 Post LOCA Cooldown and Depressurization ECA-2.1 Uncontrolled Depressurization of Both Steam Generators ECA-3.1 SGTR With Loss of Reactor Coolant:
Subcooled Recovery ECA-3.2 SGTR with Loss of Reactor Coolant:
Saturated Recovery d.
Primary system void removal.
FR-I.3 Response to Voids in Reactor Vessel e.
Use of Steam Generators with and without RCPs operating.
All of the above listed procedure deal, in some degree, with the use of steam generators.
f.' RCP Trip for other reasons.
ES-1.1 Post LOCA Cooldown & Depressurization E-3 Steam Generator Tube Rupture ES-3.1 Post SGTR Cooldown Using Backfill ES-3.2 Post SGTR Cooldown Using Blowdown ES-3.3 Post SGTR Cooldown Using Steam Dump ECA-1.1 Loss of Emergency Coolant Recirculation ECA-3.1 SGTR With Loss of Reactor Coolant:
Subcooled Recovery ECA-3.2 SGTR With Loss of Reactor Coolant:
Saturated Recovery ECA-3.3 SGTR Without Pressurizer Pressure Control FR-C.1 Response to Inadequate Core Cooling FR-C.2 Response to Degraded Core Cooling FR-H.1 Response to Loss of Secondary Heat Sink
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