ML20058A229

Requests Lic Amend No Later than 781215 Re Deletion of Reactor Trip on Turbine Trip Below 50% Pwr & Encl Requests for Tech Spec Changes & Safety Evaluation Supporting Changes

ML20058A229
Person / Time
Site:	Prairie Island
Issue date:	11/03/1978
From:	Mayer L NORTHERN STATES POWER CO.
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 7811150115
Download: ML20058A229 (34)

Text

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f NORTHERN STATES POWER COMPANY M I N N ft A r* C L i e. M I N N E r5 0 T A 55401 November 3, 1978

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Director of Nuclear Reactor Regulation g y U S Nuclear Regulatory Corrission

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Washington, DC 20555 w

i PRAIRIE ISLAND NUCLEAR GENERATING PLA';T

'3LJr30'6'/ License No.

Docket No.

50-282 DPR-42 DPR-60 License Anendnent Request Dated November 3, 1978 Deletion of Reactor Trip on Turbine Trip Below 50% Power Attached are three originals and 37 conformed copies of a request for a change of Technical Specifications, Appendix A, of Operating Licenses DPR '.2 5 60.

Also attached is one copy of the license amendment class determination and a check in the amount of $4,400.00 for the amendment fee.

This change is required to permit us to complete a design change to delete the reactor trip on turbine trip below 50 percent power in both units.

Since each unit has a 50 percent load rejectica capability, an automatic reactor trip is unnecessary below 50 percent power.

Deletion of this trip can lead to an increase in plant availability by reducing the length cf time needed to restart a unit fo11 ewing a readily correct-able turbine trip at low powe r.

!c permit us to implement this change on Unit ':o. 2 during the Auturn 19 M refueling cutage, we ask that the necessary license amendment be 1, sued no later than Decenber 15, 1978.

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Q/. s Va /f? W V L 0 Mayer, PE Manac,er of Nuclear Support Services LN /032!/deh

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cc J G Keppler G Charneff MPCA Attn: J W Fe rman

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i UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHEEN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket No. 50-282 50-306 REQUEST FOR AMENDMENT TO OPERATING LICENSE NO. DPR-42 & DPR-60

( License Acendment Reques t Dated Novembe r 3, 19 73)

Northern States Power Company, a Minnesota corporation, requests authorization for changes to the Technical Specifications as shown on the attachrents labeled Exhibit A, Exhibi t B, and Exhibit C.

Exhibit A describes the proposed changes along with reasons for the change.

Exhibit 6 is a set of Technical Specification pages incorporating the proposed changes. Exhibit C is a safety evaluation supporting the changes.

NORTHERN STATES POWER COMPANY By 6

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1. 7 L J Uachter Vice President, Power Production

& Sys tem Opera tion On this 3rd day of November, 1978, before ne a notary public in and for said County, pe rsonally appeared L J Wachter, Vice President, Power Production and Systen Operation, and being firs t duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Ccapany, that he knows the contents thereof and that to the best of his know'.ed ge, infornation and belief, the statements made in it are t ru e and t ha t it is not interposed for delay.

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Exhibit A PRAIRIE ISLAND NUCLEAR GENERATING PIANT Docket Nos. 50-282 & 50-306 LICENSE AMENDMENT REQUEST DATED NOVEMBER 3,19 78 PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS APPENDIX A 0F OPERATING LICENSE DPR-42 & DPR-60 Pursuant to 10CFR50.59, the holde rs of Operating Licenses DPR-42 and DPR-60 hereby propose the following changes to Appendix A, Technical Spe ci f ica tions.

PRCPOSED CHANGES:

a.

Add new Specification 2.3.B.5 as follows:

5.

Reactor trip on turbine trip shall be unb!acked whenever power range neutron flux is250% of rated power.

b.

Revise the first paragraph of the section 2.3 Bases on Page TS.2.3-6 to read:

The other reactor trips specified in A.3. above provide additional protection. The trip initiated by steam /feedwater flow misettch in coincidence with low steam generator water level is designed for protection from a sudden loss of the reactor's heat sink.

The safety injection signal trips the reactor to decrease the seve rity of the accident condition.

The reactor is tripped when the turbine generator trips above a power level equivalent to the load rejection capacity of the steam dump valves. This reduces the severity of the los s-of load t ransient.

REASON FOR CHANGES:

Presently the reactor protection system provides for a reactor trip upon a turbine trip when above 10% power. Since our units will accept a 50%

load rejection, a reactor trip is not necessary because of a turbine trip until at or above 50% rower. A pending design change will initiate a reactor trip upon a turbine trip when at or ateve 50%.

SAFETY EVALUATION:

Ref er to Exhibit C for the Safety Evaluation prepared by Westinghouse Corporation in support of this proposed change.

Two-of-four channels logic will be used to detect 50% power and yield permissive P-9.

This permissive will be developed in the reactor protection logic racks, and will be wired into the reactor trip matrices. The P-9 logic will be tes table, using the same design techniques as for other protection logic relays. An associated logic change wil? modify the steam dunp logic.

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Exhibit B j

' CENSE AMENDMENT REQUEST DATED NOVEMBER 3,19 78 Exh ibi t B consists of revised pages of Appendix A Technical Specifications as listed below:

Pages TS.2.3-4 TS.2.3-6 l

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TS.2 3-4 REV 2.

Low power block of single loop loss of flow is permitted whenever power range neutron flux is 610% of rated power.

3.

Power range high flux low setpoint trip and intermediate range high flux trip shall be unblocked whenever power range neutron flux is 6 9% of rated power.

4.

Source range high flux trip shall b-Y0 unblocked whenever inter-mediate range neutron flux is 6 10 amperes.

s 5.

Reactor trip on tarbine trip shall be unblocked whenever power range neutron flux is350% of rated power.

C.

Control Rod Withdrawal Stops 1.

Block automatic rod withdrawal:

a.

Turbine load 615% of full load turbine impulse pressure.

Basis The power range high flux reactor trips (low set point) provides redundant protection in the power range for a power This trip was used in the safety analysis.((gursion beginning f rom low power.

The intermediate and source range high flux reactor trips provide additional protection against uncontrolled startup excursions. As power level increases, during startup, these trips are manually blocked to prevent unnecessary plant trips.

The power range high flux (high set point) reactor trip protects the reactor core against reactivity excursions which are too rapid to be protected by temperature and pressure protective ci rcu i t ry. The prescribed set point, with allowa nce is consistent with the trip point assumed in the accident analysis.jgrerrors, The high and low pressure reactor trips limit the pressure range in which reactor operation is permitted. The high pressurizer pressure reactor trip setting is lower than the set pressure for the safety valves (2485 psig) sach that the reactor is tripped before the safety valves actuate. The low pressur-1:er pressure reactor trig frips the reactor in the unlikely event of a loss-of-coolant ac cide nt.

The ove rt empe ra tur e A T reactor trip provides core protection against DN3 for all combinations of pressure, power, coolant temperature, and axial power distribution, provided only that (1) the transient is slow with respect to piping transit 4 seconds),pg}ays from the core to the temp-erature detectors (about and (2) pressure is within the range between the high and low pressure reactor trips. With normal axial power distribution, the reactor trip limit, with allowance

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s TS.2.3-6 REV The other reactor trips specified in A.3. above provide additional protection.

The trip initiated by steam /feedwater flow mismatch in coincidence with low steam generator water level is designed for protection f rom a sudden loss of the reactor's heat sink.

The safety injection signal trips the reactor to decrease the severity of the accident condition. The reactor is tripped when the turbine generator trips above a power level equivalent to the load rejection capacity of the steam dump valves. This reduces the severity of the loss-of load transient.

The positive power range rate trip provides protection against rapid flux increases which are characteristic of rod ejection events f rom any power level. S pe ci f ica l ly, this trip complements the power range nuclear flux high and low trip to assure th a t the criteria are met for rod ejection from partial power.

The negative power range rate trip provides protection satisfying all IEEE criteria to assure that minimum DNBR is maintained above 1.30 for all multiple control rod drop accidents. Analysis indicates (Section 14.1.3) that in the case of a single rod drop, a return to f ull power will be indicated by the automatic reactor control system in response to a continued full power turbine load demand and it will not result in a DNBR of less than 1.30.

Thus, automatic protection f or a single rod drop is not required. Adminis trat ive limits in Specification 3.10 require a power reduction if design power distribution limits are exceeded by a singic misaligned or dropped rod.

References:

(1)

FSAR 14.1.1 (2)

FSAR Page 14-3 (3) FS AR 14.2.6 (4) FSAR 14.3.1 (5)

FS AR 14.1.2 (6) FSAR 7.2, 7.3 (7) FSAR 3.2.1 (8) FSAR 14.1.9 (9) FS AR 14.1.11

r LICENSE AMENp5ENT REQUEST DATED ~ NOVEMBER 3,1978 EXIIIBIT C i

DELETI0li 0F REACTOR TRIP ON TURBINE TRIP I

DELO 50 PERCEi!T POWER l'ORTilER:i STATES P05.'ER CO.

PRAIRIE ISL/J'D !?UCLEAR GEt;ERATIliG STATIOtl s

Prepared by:

J. P. Cunninghara Reactor rctection Analysis I PWRSD - fiuclear Safety e

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TABLE Of CD: Tft:lS SECT 10rt TITLE 1.0 Introduction 2.0 Identification of Causes & Accident Description 3.0 Analysis of Ef fects & Consequences 4.0 Results 5.0 Conclusion I

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_t.lST Of FIGURI.S FIGURE TI1LE

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1.

Turbine Trip with Pressure Control, Minimum Feedback 2.

Turbine Trip with Pressure Control, Mininum Feedback 3.

Turbine Trip with Pressure Control, t aximum Feedback 4.

Turbine Trip with Pressure Control, !!aximum Feedback 5.

Turbir.e Trip withsut Prec.sure Control,Itinimum Feedback 6.

Turbine Trip without Pressure Control, i:inimum Feedback 7.

Turbina Trip without Pressure Control, !!aximum Feedback 8.

Turbine Trip without Pressure Control, Maximum Feedback e

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LIST OF 1AGLES TA3LE TITI.E 1.

Initial Conditions o

2.

Time Sequence of Events for a Turbine Trip with Pressure Control 3.

Time Sequence of Events for a Turbine Trip Without Pressure Control s

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_1_. 0 l!!ilt0DI'CT10!i The present protection for a turbine trip autom tically results in a reactor trip.

llouever for plants with a 50 percent load rejection capability this trip is unnecessary if the cause of the turbine trip is readily correctable.

Deletion of the reactor trip following tur-

,bine trip for these cases would'significantly reduce the down time required to restart the plant.

Thereby an increase in plant avail-ability could be achieved.

Several safety considerations must be evaluated in order to ~ implement a new system that eliminates recctor trip below 50 percent power.

First the results of a loss-of-external-electrical-load trans.ient initiated from 50 percent pcwer must be shown to be acceptchle.

Second the results of a loss of reactor coolant flow occurring 30

' seconds after a turbine trip must be shown to be acceptable.

This analysis is required to demnnstrate recctor safety should the fast bus transfer fail folic-aing the generator rotoring delay on turbine trip.

Section 2. discusses the plant transient behavior follouing a loss of external electrical load without a subsequent turbir.e. trip and following a loss of load resulting from a" turbine trip.

In section 3 an analysis is presented for a loss of load from 52%

po.ter resulting from a turbine trip but without a direct reactor trip.

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2.0 IDDITirlCATinN OF CIE ES AND ACCIDutT DESCRIPTION A major load loss on the plant can result from loss of external electrical load die to some electrical system disturbance.

Offsite a-c power remains available to operate plant components such as the

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reactor coolant pumps.

Following the loss of generator load, an immediate. fast closure of the turbine control valves will occur.

This will cause a sudden reduction in steam flow, resulting in an increase in pressure and tc,rperature in the steam generator shell.

As a result, the heat transfer rate in the steam generator is reduced, causing the reactor coolant temperature to rise, which in turn causes coolant expansion, pressurizer insurge, and RCS pressure rise.

For a loss of external electrical load uithout subscquent turbine i

trip, no direct reactor trip signal would be generated.

The plant would be expected to trip from the Reactor Protection System.

In the event that a safety limit is approached, protection would be provided by the high pressurizer pressure and overtemperature AT trips.

In the event the steam dump valves fail to open following a loss of loadorturbinetrip,thesteamgeneratorsafetyvaivesmayliftand the reactor may be tripped by the high pressurizer pressure signal, t

the high pressurizer water level signal, or the overtemperature AT signal.

The steem generator shell side pressure and reactor coolant temperatures will increase rapidly.

The pressurizer safety valves and steam, generator safety valves are, howevcr, sized to protect the Reactor Ccolant System (RCS) and steam ger.crator against overpressure I

for all load losses without assuming the operation of the steam dump system, pressurizer spray, pressurizcr power-opcrated relief valves, '

automatic rod cluster control assembly control or direct reactor trip on turbine trip.

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'The steam generator safety valve capacity is si.'ed to remove the steam flow at the caximun calcul..'ed turbin load (105 percent of s'. cam flow at rated power) fro:n the steam generator without exceeding 110 percent of the stean sy stem design pressure.

The pressurizer safety valve capacity is sized based on a complete loss of heat sink

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with the plant initially crerating at the maximum caleviated turbine load along with operation of the steam generator safety valves.

The

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pressurizer safety valves are then able to relieve sufficient steam to maintain the RCS pressure within 110 percent of the RCS design a

' pressure.

g For a. turbine trip event, the reactor would be tripped directly (unless below approximately 50 percent power) from a signal derived from the turbina auto stop emergency trip fluid pressure and turbine stop valves.

The turbine step valves close rapidly on loss of trip-fluid pressure actuated by one of a number cf possible turbine trip signals.

Turbir.e trip initiation signals include:

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Generator Trip 2.

Low Condenser Vacuum 3.

Loss'of Lubricatir.g Oil 4.

Turbine Thrust Bearing failure 5.

Turbine Overspend 6.

!!anual 1 rip Upon initiation of step valve closure, steam flow to the turbine steps abruptly.

Sensors on the step valves detect the turbine trip and initiate steam daTp and, if above 50% pc',.or, a reactor trip.

The loss of steam fic: results in an almost immediate rise in secondary system temperature and pressure with a resultant primary system tran.

sient.'

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lhe automatic steam duus !.ystem would normally accommodate the excess steam generation. Reactor coolant temperatures and pressure do nnt significantly increate if the stem dump system and pressurizer pres-sure control systera are functioning properly.

If the turbine con-denser were not available, the excess steam generaticn would be' dumped to the atmosphere and main feedwater flow would be lost.

For this situation fcedwater flow would be maintained by the Auxiliary Fec'dwater System to insure adequale residual and decay heat removal capability.

Should the steam dump system fail to operate, the steam generator safety valves may lift to provide pressure control, as j

discussed previously.

Normal po.ccr for the reactor coolant rumps is supplied through busses fro.n a transformcr connected to the generator.

!! hen a generator trip occurs, the busses are automatically transferred to a transformer supplied frca external power lines, and the pumps will continue to supply coolant flow to the core.

Following any turbine trip where there are no electrical f aults which rcquire tripping th,e generator l

from the natuork, the generator remains connected to the network for approximately 30 seconds.

The reactor coolant pumps remain connected to the generator, thus ensuring flow for 30 seconds before any transfer is made.

Should the net.ork bus transfer f ail at 30 seconds, a ccmplete loss of forced reactor coolant ficw would result.

The immediate effect of.

loss of cool:nt flow is a rapid increcsc in the coolant temperature in additicn to the in:reased coolant tcrperatL

.s a result of the turbine trip.

This increase could result in L.

.ith subsequent fuel daaage if the rea:ter were not tripp:d prcmptly.

The follcuing signals provide the necessary protecticn against a complete loss of flow accident:

1.

Reactor ccolant purrp pcwer supply undervoltage.

2.

Low reactor coolant loop flow.

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I lhe reactor trip on reactor coolant pu:rp undervoltoge is provided to protect against conditions which can cause a' loss of voltage to both reactor coolant pu:nps, i.e., station blackout.

This function is blocked below approximately 10 percent power (Permissive 7).

Thereactortriponlowprimarycoolantloopflowisprovidedto protect against loss of flew conditions which affect only one reactor coolant leop.

This functico is generated by tuo out of three 10w flow signals per reactor coolant loop.

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3.0 fMALYSIS or ErFECTS_ Alm CON 5i:0yl!;CES flethod of Analysiq In this analysis, the behavior of tie unit is evaluated for a complete loss of stecm load fro.n 52 percent of full po,wer without direct reactor trip.

This shows the adequacy of the pressure reliev-ing devices and also downstrates the core p otection margins; that is, the turbine is assue d to trip without actuating all the sensors

• for reactor trip on the turbine step valves.

The assumption delays reactor trip until conditions in the RCS result in a trip due to other signals.

Thus, the analysis cssumes a worst transient.

In addition, no credit is taken for steam dump.

Itain feeducter flow is terminated at the time of turbina trip, with no credit taken for auxiliary feeduater to mitigate the consequences of the transient.

A fast bus transfer is attempted 30 seconds following the loss of steam lead.

Tl:t transfer to an external power source is. assumed to f ail which results in a complete loss of flod transient initiated fro a the loss of load conditions.

The loss of floa transient coincident with turbine trip transients

.are analyzed by employing the datailed d'igital ccTputer codes II), LOFTRAri(2), FACTRAU(3), and Ti!INC.

The PHOEltlX PHOENIX Code calculates the loop and core flows following the failure of the fast b;s transfer.

The LOFTRAN 'ccde shulates the neutron kinetics, RCS, pressurizer, pressurizer relief anu safety valves, pressurizer spray, st eam generator, and ste n generator safety valves.

The program ccmputes pertincnt plant variables including temperatures, pressures, and power level from PiiOENIX.

The FACTRAN Code is then used to calculate the heat flux transient based on the nuclear power and flow from 1.0FTRAN and PHOEf!!X.

Finally, the THIffC Code is used to calculate the DNBR during the transicnt based on the heat flux from FACTRAN and flow from LOFTRAM.

The DNDR transients presented represent the minimum of the typical or thimble cell.

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flajor assuir.pticus cre sunraarized below:

1.

Initial._Orraf inn Condit. inns - the initial reactor power and RCS terrperatures are assumed at their maximum values consistent with the steady state 52 percent power operation including allowances for calibration and instrument errors The initial RCS pressure is assumed at a minimum value consistent with the steady state 52

' percent po,cer operation including allowances for calibration and instrument errors.

The initi31 RCS flow is assumed to be Con-sistent for two loop operation. This results in the maximum

, power dif ference f or the load toss and the minimum margin to core protection limits at the initiation of' the accident.

Table 1 summarizes the initial conditions assumed.

2.

lloderator and Doppler _ Coefficient of Reactivity,_ - the turbine trip is analyzed with both a least negative moderator temperature coefficicnt and a large negative moderator temperature coef-ficient.

Doppler power coefficients are adjusted to, provide con-sistent inaximum and minimum reactivity feedback cases.

3.

Reactor Con 2p_l, - f rom the standpoint of the maximum pressures attained, it is conservative to assume that the reactor is in manual control.

If the reactor were in automatic control, the control rod banks would move prior to trip and reduce the severity of the transient.

4.

He,3_m Release - no credit is taken for the operation of the steam dump sysica o steam generator power operated relief valves.

The steam generatcr pressure rises to the safety valve setpoint where steam relene through safety valves limits secondary steam pres-sure at the setpoint value.

5.

Pressurizer Spray and ' Power Operated Relief Valve! - two cases for both the minimum and maximum reactivity feedback cases are analyzed:

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a.

Full credit is taken for the eff ect of pressurizer spray and power-operated relief valves in reducing or limiting the coolant pressure.

Safety valves are also availabic.

b.

No credit is taken for the effect of pressurizer spray and power-operated relief valves in reducing or limiting the coolant pressure.

Safety valves arc operablel 6.

Feeductcr Ficu - main fecdaater flow to the st' cam generators in assumed to be lost at the time of turbine trip.

No credit is taken for auxiliary feeduater flcu since a stabilized plant con-dition will be reached before auxiliary feeduater initiation is normally assumed to occur; however, the auxiliary feedwater pu.mps would be expccted to start on a trip of the main feedwater pumps.

The auxiliary feedvater flow would remove core decay heat follcuing plant stabilization.

7.

' Reactor trip is actuated by the first Reactor Protection System trip setpciot reached with no credit taken for the direct reactor trip on the turbine trip.

Trip signals are expected due to high pressurizer pressure, overtemperature AT, high pressurizer wate'r level, icw reactor coolant ic0p flow, and reactor coolant pump poner supply undervoltage.

Except as discussed above, ncrmal reacter control system and Enginecred Safety Systems are not required to function.

The Reactor Pro.ecticn System may be required to function following a turbine trip.

Pressurizer safety valves and/or steam generator safety valves may be required to cpen to maintain system pressures below allowable limits.

No single active failure will prevent opera; tien of any system required to function.

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4.0 RESULTS

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The transient response for a turbine trip from 52 percent of full power operation are shown for four cases:

two cases for minimum reactivity feedback and two cases for maximum reactivity feedback (Figure 1 through 8).

The calculated scquence of events for the accident is shown in Tables 2 and 3.

Figures 1 and 2 show the transient responses'for the tctal loss of steam load with a least negative moderatcr temperature coefficient assuming full credit for the pressurizer spray and pressurizer power-operated relief valves.

No credit is taken for the steam dump.

The fast bus transfer is attempted and failed 30 seconds after the total loss of steam load.

The tra'nsfer failure results in an undervoltage trip of the reactor and the initiation of the loss of flow transient.

The minimum DNBR ren3 ins weil above the 1.30 limit.

The steam generator safety valves limit the secondary steam condi-tions to satura; ion at the safety valve setpoint.

Figures 3 and 4 show the responses fer the total loss of stean load with large negative moderator and fuel temperature coefficients, low power coefficient and leu delayed neutron fractions.

All other plant parameters are the same as the above.

The minimum DN3R remains Well

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above the 1.30 limit throughout the transient.

Pressurizer spray and, steam generatcr safety valves prevent overpressurization in primary and secondary systens, respectively.

The turbine trip accident was also studicd assuming the plant to be initially crerating at.52 percent of full power with no credit taken for the pressurizer spray, pressurizer pcuer-operated relief valves, or steam dump.

For the case with a least negative moderator coef-ficient, the reactor is tripped on the high pressurizer pressure signal.

The fast bus transfer is assumed to fail at 30 seconds af ter the total loss of load.

Figbres 5 and 6 shew the transients with a 4-1.

Icast negative moderator coefficient.

The neutron-flux remains essentially constant at 52 percent of full po.ver until the reactor is tripped.

The DIiCR remains above 1.30 throughout the transient.

In this case the pressurizer safety valves are actuated and maintain system pressure below 110% of the design value.

Figures 7 and 8 show the transients with maximum reactivity feedback with the other assumptions being the same as in the preceding case.

j Again, the minimum Df'BR re:uins above 1.30 throug'iout the transient.

In this case the pressurizer safety valves are not actuated.

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Pesults or the analyses sinw that l'io plant design is such that a turbine trip without a direct or imetedicte recctor trip presents no hazard to the interpily of the I!C3 cr the main stecn system.

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Presr.ure relieving devices incorporated in the tuo systet1s are adequate to limit the n.uieu pr.;asurer, to within ihe design limit.s.

The cnalysis also d:nnstrates that fcr a ceraplete loss of forced reactor coolant flo. initiated from the most adverse precenditions of a turbina trip, the Dil3R does not decrease belca 1.30 at any time during the transient.

Thus, no fuel or clad damacc is predicted, and all cpplicable ccceptatce criteria are rnet.

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LIST Of PITritr.f;CES 1.

R. K. Lind.! felt, T. f t. [ lord 21on, and T. J. !!and, " Calculation of Flav Cor.tdown Af ter I oss of Reactor Coolant Pump (Pl!0DlIX Code)", !! CAP-7551, August 1970.

2.

T. !!. T. !!urnett, C. J. f4cIntyre, and J. C. Baker, "l.0rTRAT1 Code

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Description," UCAP-7378, !! arch 19/2.

3.

C. Ilunin, "fAClRAi! - A FORTRtdl IV Code for Thermal Transients in a U0 Fuel Red", ll CAP-7337.

2 4.

Cnelemer, J. Weiseman end L. S. Tong, "Subchannel Thermal Analyscs of Rod Dundle Corc," ll CAP-7015, Rev.1, January,1959.

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Tant.C 1

!!t1 TI AI._ C0::DITIO*l5 fio, of Active Leops 2

Core Po ce, ftit 858 a

Thermal Design rio.: per Loop, GPM 178000 Reactor Coolant Tcaperature Vessel Outlet, OF 578.2 Vessel Inlet, F

544.2 Reactor Coolant Pressure, psia 2230 Steam Generator Steam Temperature, F

531.6 Pressure, PSIA 897.4 6

Flow, 10 lb/hr total 3.43 4

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lACLE 2

_T_IME SEgto:rE Or E'.T'iTS F01: A Tt'R U :E Trip 1l1111 PhiS3Uil!7Q PP.F55Uf!E C0?.'Ir.0L Ev en t, Time (sec.)_

1.

14inimtm feedback (i!0L)

Turbine Trip 0

Peak pressurizer prer.sure occurs 10.4 Initiation of sterm release from

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steca generator safety valves 16.8

,I Fast bus transfer failure, flow coastdoe.n begins, and undervoltage trip occurs 30.0 Rods begin to fall 31.6 liinimu:a DNCP. occurs 32.6 2.

Max imum Feedback (ECL)

Turbine Trip G

Peak pressurizer pressure occurs 15.3 Initiation of steam release from steam generator safety valves 17.5 fast bus transfer failure, flow I

coastdo<:n begins, and undervoltage' l

trip occurs 30.0 P.ods begin in fall 31.6 -

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Tli.it $f ND:7 0F I:Vl!!TS FOR A Titi:Pl!!F TRIP ll!'lH0t:T Pl? DSS!'R !ZLR Pl!f SSU:!E CD :!ROL

Event, Tirac (sec.)

1.

liin iu n feedbcck (P.01.)

Turbine Trip 0

liigh pressurizer pressure trip 12.4 occurs Rods begin to fall 13.4 Peak pressurizer pressure cccurs 15.9 Initiation of sicaat releasc from steam generater safety valves 16.6 Fast bus transfer f ailure, flua coastdy,,n begir.s '

30.0 2.

Maximua Feedback (EOL)

Turbine Trip 0

iligh pressurizcr pressure irip occurs 14.4 Rods begin to fall 15.4 Initiation of steam release from stram generator safety valves 17.3 e

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lAIPE 3 (Continued)

Event Tin:e (r.cc. ),

1 Peak pressurizer pressure occurs 17.5 fast bus trcnsftr failure, flow coast <fo.;n begins and undervoltage trip occurs 30.0 9

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