ML17254A522
| ML17254A522 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 08/19/1985 |
| From: | Kober R ROCHESTER GAS & ELECTRIC CORP. |
| To: | Zwolinski J Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.K.3.05, TASK-TM GL-85-12, NUDOCS 8508230161 | |
| Download: ML17254A522 (18) | |
Text
REGULATORY INNATION DISTRIBUTION SYSTE(RIBS)
ACE-'ESS1'ON NBR: 8508230161 DOC ~ DATE: 85/08/19 NOTARIZED:
NO DOCKET FACIAL:50<<244 Robert Emmet Ginna Nuclear Plant< Unit 1E Rochester G
05000204 AUTH,NAME AUTHOR AFFILIATION KOBER R ~ N ~
Rochester Gas 8 Electric Corp.
RECIP,NAME RECIPIENT AFFILIATION Zl'JOLINSKIr J sA ~
Oper ating Reactor s Branch 5
SUBJECT:
Responds to Gener ic Ltr 85-12'Implementation of TMI Action Item II.K,3'EAutomatic Trip of Reactor Coolant Pumps."
Reactor coolant pump trip criterion, based on primary 8 secondary sys pressureiimplemented at facility, DISTRIBUTION CODE:
AOASD COPIES RECEIVED:LTR ) ENCL j SiZe:
TITLE:
OR Submittal:
TMI Action Plan Rgmt NUREG-0737 L NUREG-0660 NOTES:NRR/DL/SEP icy.
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/I//IE I///I ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, Lkl~ ONIJA I/f,v 3P::::
ROCHESTER, N.Y. 14649-0001 ROGER W. K08ER VICE PRESIDENT ELECTRIC Si STEAM PIIODUCTION TELEPHONE AREA CODE 7IE 546.2700 August 197 1985 Director of Nuclear Reactor Regulation Attention:
Nr. John A. ZwolinskiE Chief Operating Reactors Branch No.
5 U.S. Nuclear Regulatory Commission Washington>
D.C.
20555
Subject:
Generic Letter No.85-12E "Implementation of THI Action Item II.K.3.5E Automatic Trip of Reactor Coolant Pumps" R. E. Ginna Nuclear Power Plant Docket No. 50-244
Dear Hr. Zwolinski:
This letter is in response to Generic Letter 85-127 dated June 28<
1985.
Consistent with the Ginna Steam Generator Tube Rupture SER>
a reactor coolant pump (RCP) trip criterion has been implemented at Ginna.
This criterion is based on primary and secondary system pressure and assures RCP trip for all losses of primary coolant for which trip is considered necessary but also permits RCP operation to continue during most non-LOCA> including steam generator tube rupture events up to the design basis double-ended tube rupture.
Our previous letters dated April 227 19837 December 217 1983 and April 107 1984 provided a majority of the plant specific information'urrently being requested by Generic Letter 85-12.
For completeness>
the attachment to this letter contains the requested information to aid the Staff in the plant-specific review.
V truly yours, R
er W. Kober Attachment 85082301&i 850819 PDR-ADQCK 05000244 "-
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Attachment Generic Letter 85-12 NRC Request:
A.l.
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
NRC Request:
A.2.
The instrumentation used to determine if RCPs should be tripped is the primary system wide-range pressure indication and the steam generator pressure indication.
There are two primary system-wide-range transmitters and six steam generator pressure transmitters.
Therefore>
there are three steam generator pressure signals per steam generator and>two 'wide-'range primary system pressure signals.'
'1 I
j Identify the instrumentation uncertainties for both normal and adverse containment conditions.
Describe the'asis for the selection of the adverse containment parameters.
Address>'as appropriate>, local conditions such as fluid jets or pipe whip which might influence the instru-mentation reliability.
Response
The instrument uncertainties used to determine the RCP trip setpoint are listed below.
Since the steam generator pressure transmitters are located outside containment>
adverse containment conditions do not apply to steam generator pressure.
Instrument Channel Uncertainty normal adverse containment containment psi ps1 SG pressure RCS pressure 45 85 45 327 The basis for selection of adverse containment parameters is the, Westinghouse Owners Group (WOG)
Emergency Operating Procedures (EOPs).
The EOPs define adverse containment conditions as a
pressure of approximately55 psig or a radiation level of approximately lO R/hr.
The transmitter uncertainty is based on the generic qualification of Foxboro transmitters for post LOCA environ-ment.
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The transmitter locations were selected such that fluid jets or pipe whip would have negligible effects on instrumentation reliability.
See SEP Topic III-5.A.
NRC Request:
A.3.
In addressing the selection of the criterion>
consideration to 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.
If a licensee determines that the WOG alternative criteria are marginal for preventing unneeded RCP trip> it is recommended that a more discriminat-ing plant-specific procedure be developed.
For exam'pie<
use of the NRC-required inadequate
'core cooling instrumentation may be usef'ul'~to indicate the"need for RCP trip".'icensees:should take credit for all equipment (instrumentation) available to the operators for which the licensee has sufficient confi'dence th'at it will be operable during the expected conditions.
Response
The WOG supplied analysis is presented in Reference 1.
The Westinghouse plants were grouped into categories with similar plants (generic groups) and accident analyses were performed for representative plants in each category.
The resulting generic group analysis is conservative for Ginna.
The LOFTRAN computer code was used to perform the alternate RCP trip criteria analyses.
Both Steam Generator Tube Rupture (SGTR) and non-LOCA event were simulated in these analyses.
Results from the SGTR analyses were used to obtain all but three of the trip parameters.
LOFTRAN is a Westinghouse licensed code used for FSAR SGTR and non-LOCA analyses.
The code has been validated against the January 1982 SGTR event at the Ginna plant.
The results of this validation show that LOFTRAN can accurately predict RCS pressure>
RCS temperatures and secondary pressures especially in the first ten minutes of the transient.
This is the critical time period when minimum pressure and subcooling is determined.
The major causes of uncertainties and conservatism in the computer program results>
assuming no changes in the initial plant conditions (i.e.< full power< pressurizer level>
all SI and AFW pumps run) are due to either
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models or inputs to LOFTRAN.
The following are considered to have the most impact on the determination of the RCP trip.criteria:
l.
Break flow 2.
SI flow 3.
Decay heat 4.
Auxiliary feedwater flow The following sections provide an evaluation of the uncertainties associated with each of these items.
To conservatively simulate a double-ended tube rupture in safety analyses>
the break flow model used in LOFTRAN includes substantial amount of conservatism (i.e.< predicts higher break flow than actually expected).
Westinghouse has performed analyses and developed a more realistic break flow model than has been validated against the Ginna SGTR tube rupture data.
The break flow model used in the WOG analyses has been shown to be approximately 30% conservative when the effect of the higher predicted break flow is compared to the more realistic model.
The consequence of the higher predicted break flow is a lower than expected predicted minimum pressure.
The SI flow input used was derived from best estimate calculations<
assuming all SI trains operating.
An evaluation of the calculational methodology shows that these inputs have a
maximum uncertainty of + 10%.
The decay heat model used in the WOG analyses was based on the 1971 ANS 5.1 standard.
When compared with the more recent 1979 ANS 5.1 decay heat inputs, the values used in the WOG analyses is higher by about 5%.
To determine the effect of the uncertainty due to the decay heat model>
a sensitivity study was conducted for SGTR.
The results of this 'study shop,that a
20% decrease in
'"decay,, heat resulted, in only a.",1( decrease in RCS pressure for,the'"first. 10 minutes of'he transient.
Since RCS temperature is controlled by the steam dump< it is not affected by the decay heat model, uncertainty.
The AFW flow rate input used in the WOG analyses are best estimate values>
assuming that all auxiliary feed pumps are running>
minimum pump start delay<
and no throttling.
To evaluate the
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uncertainties with AFW flow rate>
a sensitivity study was performed.
Results from the two loop plant study show that>
a 64% increase in AFW flow resulted in only an 8% decrease in minimum RCS pressure>
a 3% decrease in minimum RCS subcool-ing< and an 8% decrease in minimum pressure differential.
Results from the 3 loop plant study show that<
a 27% increase in AFW flow resulted in only a 3% decrease in minimum RCS pressure>
a 2% decrease in minimum RCS subcool-ing> and a
2% decrease in pressure differential.
The effects of all these uncertainties with the models and input parameters were evaluated and it was concluded that the contributions from the break flow conservatism and the SI uncertainty dominate.
The calculated overall uncertainty in the WOG analyses as a result of these considera-tions for R.
E. Ginna is +90 to +100 psi for the minimum RCS/secondary differential pressure RCP trip setpoint.
Due to the minimal effects from the decay heat model and AFW input< these results include only the effects of the uncertainties due to the break flow model and SI flow inputs.
The RCP Trip Criterion based on primary and secondary system pressure has been implemented at Ginna.
The setpoint assuming normal containment conditions is acceptable based on Table 1 of Reference l.
NRC Request:
B.l.
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 operations 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
Essential services for RCP operation are available during a Containment Isolation signal at Ginna unless a Safety Injection (SI) signal occurs with a loss of offsite power.
Seal injection from the Chemical Volume Control System is terminated by a charging pump trip upon Reference 1:
Westinghouse Owners Group> Letter OG-110<
"Evaluation,.
of Alternate RCP Trip'Criteria>"'December 1> '1983.'
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receipt of an SI signal.
Howeveri Component Cooling Water (CCW) services to the RCP remain in operation independent of the SI and/or.,
Containment Isolation signalsi unless offsite power is lost.
A loss of offsite power coincident with, an SI signal will trip the CCW pumpsi thereby terminating CCW flow to the RCPs.
Since, the RCPs oper'ate.from offsite poweri,the",
RCPs 'will al'so be tripped and wall no't be avail-able while offsite power is lost'.
As stated above<, water. services needed for RCP operations are supplied by two independent sources.
Therefore> it is highly unlikely that both services would be terminated.
If termina-tion did occur<
the RCPs would be tripped.
NRC Request:
B.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
The components associated with tripping the RCPs are all located outside containment.
The RCPs are hard wired to the breakers which are located in the Turbine Building.
The relays associated with tripping the breakers are located at the breakers in the Turbine Building.
Control room switches are hard wired to these relays.
The relays and switches are powered from the l25V DC power system.
Therefore>
adverse
'containment conditions do not effect RCP trip reliability.
NRC Request:
C.l.
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
The RCP Trip Criterion based on primary and secondary system pressure has been implemented and the Operators have been trained in the use of this criteria.
The general philosophy regarding RCP trip is to trip the RCPs only when the trip criterion is reached.
NRC Request:
C.2. Identify those procedures which include RCP'rip related operations:
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(a)
RCP trip using WOG alternate criteria (b)
RCP res tart (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
Since the requested information is scattered throughout several procedures<
only a partial list is presented.
This will provide a guideline as to where the information is located.
Since the procedures are periodically reviewed and updated as necessary<
the information presented herein may be revised in the future.
The procedures are always available for inspection.
RCP trip using the Trip Criterion is addressed in the Emergency Series (E-Series) of procedures which are based on the Westinghouse Owners Group guidelines.
These procedures are:
E-0 E-1 "Reactor Trip or Safety Injection" "Loss of Reactor or Secondary Coolant" "Steam Generator Tube Rupture" ES-0.4 "Natural Circulation Cooldown with Steam Void in Vessel" ECA-2.1 "Uncontrolled Depressurization of All Steam Generators".
RCP restart is addressed in the E and Operating (0-Series) of procedures.
Some of these procedures are:
ES-0.1 "Reactor Trip Response" ES-1.1 ES-1.2 "SI Termination"
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"Post LOCA Cooldown and Depressuriza-tion" ECA-2.1 "Uncontrolled Depressurization of All Steam Generators" E-3 "Steam Generator Tube Rupture"
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ECA-3.1 "Post-SGTR Cooldown Using Backfill" ECA-3.2 "Post-SGTR Cooldown Using Blowdown" ECA-3.3 "Post-SGTR Cooldown Using Steam Dump" FR-C.l "Response to Inadequate Core Cooling" FR-C.2 "Response to Degraded Core Cooling" FR-P.1 "Response to Imminent Pressurized Thermal Shock Condition" FR-I.3 "Response to Voids in Reactor Vessel".
Decay heat removal by natural circulation is addressed in the 0 Series of procedures.
Primary system void removal is addressed in the E
Series of procedures.
Use of steam generators with and without RCPs operating is not specifically addressed.
It appears that this subject is inherent in the cooldown guidance given for normal and natural circulation cooldown.
RCP trip for other reasons will be addressed in the Abnormal Procedures (AP) Series of procedures when the Revision 1
E procedures are implemented later this year.
RCP trip for other reasons're currently addressed in the Ginna E procedures.
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