W3P88-0320, Forwards Responses to 880307 & 0818 Requests for Addl Info Re Reactor Coolant Pump Trip Strategy
| ML20195D494 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 10/31/1988 |
| From: | Burski R LOUISIANA POWER & LIGHT CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| W3P88-0320, W3P88-320, NUDOCS 8811070030 | |
| Download: ML20195D494 (7) | |
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LO UISI AN A / si7 84RONNe sinter. p.O.e,0x e0340 POWER & LIGHT NEW ORLEANS. LOUISIANA 70160 * (504) 595-3100 IitTOIsdONtU October 31, 1988 W3P88-0320 A4.05 QA U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
SUBJECT:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 Reactor Coolant Pump Trip Strategy
REFERENCES:
- 1) NRC Request for Additional Information - Reactor Coolant Puup Trip, dated March 7, 1988.
- 2) NRC Request for Additional Information - Reactor Coolant Pump Trip, dated August 18, 1988.
Centlement The NRC via references one and two submitted questions to LP&L regarding the reactor coolant pump trip strategy for Waterford 3.
Provided as attachment one are the responses to the referenced questions.
Please contact me or Robert J. Murillo should there be any questions concerning this information.
Yours very truly,
(
d R.F. Burski Manager Nuclear Safety & Regulatory Affairs RFB/RJM/gsp Attachment cc E.L. Blake, W.M. Stevenson, J.A. Calvo, D.L. Wigginton, R.D. Martin, NRC Resident Inspector's Office (W3) haj\\
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Attcchment On2 To W3P88-0320 i
Questions from March 7, 1988, letter from NRC (D.L. Vigginton) to LP&L (J.G. Dewease):
QUESTI6N 1:
Combustion Engineering report CEN-268 uses a combination of subcooling margin (20*F) and containment activity or absence of secondary activity to assess the need to trip the last two reactor coolant pumps (RCPs). However, the Licensee's response to Question 1 (of NRC Generic Letter 86-06) did not mention subcooling margin in the diagnostic flow chart.
(a) Verify that the Waterford 3 RCP trip strategy is representative of the generic approach used in CEN-268. Justify any deviations that are not conservative.
(b) Express the second RCP trip setpoint in terms of the subcooling margin (*F) and secondary activity limits. Verify that the instrumentation uncertainties are acceptable.
l (c) Do the instruments, which indicated the RCP trip conditions, serve other functions such as pump restart or primary roid detection and management? Verify that the instruments remain qualified when needed.
RESPONSE (a) & (b):
The intent of the RCP trip strategy developed in CEN-268 is to ensure all RCP's are tripped in the event of a LOCA. The Waterford 3 RCP trip strategy is con-sistent with this intent. The trip strategy is also consistent with the NRC approved Emergency Procedure Guidelines (EPG's) given in CEN-152 Rev. 2, which does not include a reference to subcooling margin.
The Waterford 3 strategy requires tripping two RCP's if Reactor Coolant System (RCS) pressure drops below a setpoint value, consistent with CEN-268. The remaining two RCP's are tripped if a LOCA is diagnosed. The Waterford 3 procedures (and CEN-152) use tren/.s of pressurizer pressure and level, steam generator pressure, containment pressure and temperature, and steam generator secondary activity to diagnose a LOCA. This strategy ensures the RCP's are tripped in a timely manner during a LOCA consistent with CEN-268.
The trip strategy does not require specific trip setpoints based on subcooling margin and secondary activity limits.
Since the decision to trip the remaining two RCP's during a LOCA is based on parameter trends rather taan explicit setpoints, instrument uncertainties have a smaller impact and are acceptable.
RESPONSE (c):
1 The instruments used to diagnose a LOCA at Waterford 3 include instruments for pressurizer pressure and level, reactor containment pressure and temperature, and i
steam generator pressure and secondary activity. There are envitonmentally qualified instruments for monitoring all these parameters.
These instruments are used by operators for control and monitoring of plant conditions during the accident.
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Attachment One Is W3P88-0320 QUESTION 2:
CEOG Report CEN-268 Section 5.4 evaluated the effectiveness of the RCP trip setpoints for an Increased Heat Removal (IHR) event, namely the inadvertent increase in turbine power from zero load to full power.
The results of the analysis indicated that the pressurizer pressure decreased to 1700 psig. Since the 1300 psia trip setpoint used in the generic analysis is below this value, no reactor coolant pumps were tripped for the analyzed case.
However, the 1700 psig value does approach the Waterford 3 RCP trip setpoint (1621 psig) when it is combined with the pressurizer pressure instrument uncer-tainty (150 paia tn 1100 psia). See response to Question 2, Reference 3 (LP&L letter W3P86-2829 dated February 5, 1987). Describe the consequences on the plant if RCP trip indication is reached.
Confirm that the CE00 evaluation of RCP trip setpoints for the other sample transients (see Section 5, CEN-263) is conservative compared to Waterford 3.
RESPONSE
If the pressure setpoint for tripping the first two RCP's is reached for a non-LOCA transient, tripping these RCP's will not significantly impact the ability to cope with the event. The two remaining RCP's provide sufficient forced circu-lation flow in each steam generator loop so that operator actions in response to the event are not significantly affected. The analyses and conclusions provided in Section 5 of CEN-268 are bounding for Waterford 3.
QUESTION 3:
As cited in the Licensee's response to Question 3 (of Generic Letter 86-06),
CEN-268 and Supplement 1-NP describe the major assumptions and conservatisms used in the generic CE00 analyses. These reports have been reviewed and accepted by the NRC staff as documented by Generic Letter 86-06.
The Licensee has indicated that the CEOG reports and corresponding RCP trip strategy are conservative with respect to Waterford 3.
We find that this response alone is insufficient to identify and evaluate the specific areas of cons e rva tism. The Licensee needs to describe the technical basis for the con-i servatism by ecmparing the generic assumptions with the Waterford 3 features.
(a) Core power (b) High Pressure injection capacity and number of pumps available (c) Steam generator heat transfer characteristics (d) Make up flows (c) Setpoints for reactor trip, safety injection, and accumulator injection.
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Attechment One To
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W3P88 0320
RESPONSE
From discussions held with the NRC on July 28, 1988, LP&L determined that the intent of this question is to provide justification for the RCS pressure set-point for tripping the first two RCP's.
As stated in Appendix A of CEN-268, a specific calculation was performed for the 3410 MWt plants (Waterford 3 and SONGS 2 & 3) which determined a nominal RCS pressure setpoint of 1361 psia for tripping l
the first two RCP's.
This calculation used Waterford 3 values for core power, 3
safety injection flow, steam generators heat transfer, makeup flow, and set-J points for reactor trip, safety injection and accumulator pressure to determine the RCS pressure setpoint.
Since Waterford 3 specific values were used to determine this setpoint, the "generic" assumptions are identical to Waterford 3.
Adding a Waterford 3 specific uncertainty of 260 psia results in a setpoint of l
1621 psia fcr tripping the first two RCP's.
QUESTION 4:
It is not clear that the Licensee has identified the procedures and operator training that involve use of single steam generators.
(See Response to Question 3
4, Reference 3.) The Licensee is requested to provide this information.
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RESPONSE
The Waterford 3 E0P's provide explicit instructions for accommodating RCS cooldowns using only a single steam generator. These instructions are incor-porated into the specific event E0P's including OP-902-004, Excess Steam Demand Event Recovery, and OP-902-007, Steam Generator Tube Rupture Recovery.
Operators undergo continuing training on the E0P's including situations that involve use of l
single steam generators.
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I Attachment One i
To~
I W3P88 0320 The following additional questions from the NRC are related to the "Response to Generic Letter 86-06, Implementation of TMI Action Item II.K.3.5."
QtLESTION 5:
P In CEN-268, Rev. 1, the CEOG stated that containment isolation could cause the loss of pump cooling water and that continued operation under these conditions could result in Reactor Coolant Pump (RCP) damage. Timely operator action could restore water service and prevent pump damage. Each utility was to review its plant specific pump cooling service system requirements and make any changes necessary. On April 23, 1985, the Licensees with CE plants were sent a request for additional information on this area.
In Generic Letter 86-06, the staff i
stated that the response to this request for information was not available for Waterford Unit 3 and that the review of this item would be completed as part of the plant specific implementation SE review. The submitta.is for Waterford 3 have
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not addressed this issue. Therefore, provide the following information.
Iden-tify the systems which are essential for the continued operation of the reactor coolant pumps. Does any containment isolation signal or any other signal (euch as SIAS) result in the termination of these systems that is essential for the continued operation of the RCP's? Identify the signals and the systems affected l
during containment isolation.
RESPONSE
LP&L has previously addressed the issue of reactor coolant pump cooling water isolation. The NRC via a letter dated July 16, 1985, from G.W. Knighton to i
R.S. Leddick requested that LP&L address the issue of RCP cooling water isola-tion. LP&L via letter W3P85-2611 dated August 12, 1985, addressed the issue.
LP&L conveyed in this letter that the original design for Waterford 3 provided for interruption of Component Cooling Water (CCW) to the RCP seal coolers upon receipt of either a Safety Injection Actuation Signal (SIAS) or a Containment Isolation Actuation Signal (CIAS).
Both SIAS and CIAS are actuated via s low pressurizer pressure or a high containment pressure signal.
Ar, described in LP&L letter W3P88-0405 dated February 16, 1984, LP&L implemented hardware changes to substitute the Containmet-Spray Actuation Signal (CSAS) for the previous SIAS/CIAS in isolating CCW to the RCPs. The CSAS is actuated oc high-high containment pressii coincident with SIAS. As a result, CCW isolation to the RCPs does not occur a receipt of a low pressurizer pressure signal or I
CIAS. Non-spurious actuation of CSAS, and therefore isolation of CCW, would
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occur only for a large LOCA or main steam line break in containment.
l Systems essential for the continued operation of RCP's include the CCW system and the controlled bleedoff flow through the RCP seals. The normal controlled l
bleedoff path to the Volume Control Tank (VCT) is isolated upon a CIAS, which has a setpoint for high containtnnt pressure of 17.1 psia. When the normal bleed-off path is isolated, a pressure relief valve will open to the quench tank to provide sufficient bleedoff flow for continued RCP operation. The CCW system i
provides cooling water to the RCP seals and is isolated upon a CSAS, which has
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Attachment Ons l
To W3P88-0320
-RESPONSE to Qu3stion 5 Cont'd.
i a setpoint for high-high containment pressure of 17.7 psig. The operation of these systems are not terminated for other Emergency Safety Features Actuation Signals, (ESFAS), such as SIAS. No other systems essential for RCP operation are affected by a CiAS.
s QUESTION 6:
t If essential RCP cooling water services are terminated, provide a description of the operation guidelines, training, and procedures in place which assure that these services are restored in a timely manner to prevent seal damage or failure, once a non-LOCA condition has been confirmed.
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RESPONSE
The Waterford 3 E0P'a provide explicit instructions to operators to trip the l
RCP's if CCV flow to the RCP seals has been lost for greater than or equal to l
3 minutes. The E0P's also contain explicit statements that the RCP's should
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not be restarted if CCW flow to the RCP seals has been lost for greater than l
or equal to ten minutes. Operators will attempt to restore CCW flow within
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this 10 minute period.
QUESTION 7:
1 Usually RCP component cooling water is provided for both the seal injection i
and the component cooling water (such as motor bearings). Are both seal injection i
and component cooling water required for RCP operation? It is possible to operate the RCP with only one cooling service? Indicate the actions taken if all l
cooling water services are terminated.
Indicate the recovery procedure for l
restoring the preferred RCP cooling water service. !!ow much time is the operator l
allowed to perform this action?
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RESPONSE
o The Waterford 3 RCP design does not include seal injection. Only CCW is required for the necessary cooling of operating RCP's.
If CCV flow to the RCP's is lost I
for 3 minutes or longer, the operators are required to trip the RCP's.
I Waterford 3 procedures do not allow restarting RCP's if the CCV supply has been
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l lost for 10 minutes or loager. Operators are trained on the actions to be taken to restore CCW within ttas limited time. The time criteria are based on the j
l recoassendations of the pump s'anuf acturer, Byron Jackson.
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Attcchment On2 Ia W3P88-0320 QUESTION 8:
Provide confirmation, including the technical basis, that containment isolation with continued RCP operation will not lead to seal damage or pump damage or failure.
RESPONSE
Containment isolation on a CIAS with cons'nued RCP operation will not lead to RCP seal damage because CCW !
oot isolated on a CIAS. However, if containment pressure increases to the higr high pressure setpoint of 17.7 psia, a CSAS is generated and the CCW supply to the RCF scals is isolated. Upon CCW isolation, operators are trained to restore CCW flow or trip the RCPs if CCW flow is lost for three minutes or longer. This ensures that damage to the RCP acals or pump does not occur.
QUESTION 9:
Since RCP trip will be required for LOCA events, assurance must be provided that RCP trip, when required, will occur. To address this concern, provide the following information:
p (a) Identify the components required to trip the pumps.
Include the relays, power supplies, breakers or other equipment. Address re-liability, alternate trip methods, and operator training. How much time is the operator allowed to trip the RCP's from within the Control Room before using an alternate method? How much time is the operator allowed to travel to an alternate location to trip the pumps?
(b)
If necessary, as the result of the location of any critical com-ponent, include the effects of adverse conditions inside or outside containment on RCP trip reliability. Desc ibe the basis for the l
adverse conditions selected.
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RESPONSE
Various methods are available to trip RCP's.
A control switch for each pump located in the Control Room can be used to de-entrgize the RCP's.
The pump breakers can be opened to obtain a local trip of the RCP's.
It is also possible to trip the RCP's by deenergizing the (non-safeti/) electrical busses which supply RCP electrical power. Adverse conditions will 90t affect RCP trip reliability f
because all equipment needed to trip RCP's is located outside containment, o
Furthermore, the analyses in CEN-268 show that the timing of RCP trip is not important. A "most probable best estimate" analysis described in Section 7.2 of CEN-268 demonstrates that acceptable consequences are obtained even if the RCPs
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are not tripped or if they are tripped (or fail) at the worst time.
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