ML19338F565

Special Low Power Tests Safety Evaluation.

ML19338F565
Person / Time
Site:	Mcguire, McGuire
Issue date:	10/10/1980
From:	DUKE POWER CO.
To:
Shared Package
ML19338F555	List: ML19338F554 ML19338F557 ML19338F558 ML19338F561 ML19338F564 ML19338F565
References
NUDOCS 8010200585
Download: ML19338F565 (47)

Text

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l MCGUIRE UNIT 71 l l

l SPECIAL LOW-POWER TESTS i l

SAFETY EVALUATION j l

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1 OCTOBER 1980 1

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1~. 0 INTRODUCTION AND SLTIARY- .

E In l an :eff ort' to meet ithe -NRCJ-regulatory requirements of NUREG-0694, "

"TMI-Related: Requirements'for New 0perating Licenses," special tests

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similar to those performediat'Sequoyah for reactor power levels at Lor below15% of Rated Thermal' Power are proposed. ' These tests would demon-strate the plant's ' capability in. several simulated degraded modes of-Joperation and.would provide opportunities for operator training. Th e

. basic' mode 'of ' operation to be ' demons trated is i natural ^ circulation with various pertions . of ' dae plant equipment not operating, e.g. , pressurizer.

. heaters,: loss of off site power (simulated), and steam generators isolated.

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' Westinghouse has reviewed die proposed tests and has determined diat with. close ~ operator surveillance of: parameters and suitable operator action points in the event of significant deviation from te'st condi-tions, - the tests as outlined in the McGuira -Special Test ' procedures - are acceptable and -can be perforned with minimal risk. It is recognized 4

th a t in order to perform' these tests some automatic saf ety functions ,

.. reactor trips and safety injection, will be defeated. Wes tingtouse h as

' determined a set of operator action points which 'should replace these automatic actuations. It is also- recognized diat several technical

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, specification requireyents wil'11not be net while either preparing for or

- performing th ese _ tests. - Again Westinghouse has determined diat the low power levels and operator action will suf fice during these time periods.

Wes tinghouse has reviewed the ef fect of the proposed test conditions on die. incidents and f aults . Waich were discussed in the_ Accident Analysis ,

secti'on ofi the McGuire Final Safety Analysis Report. In most cases, th e FSAR ; discussion 'was found to boups the consequences of -such events R loccurring'under1testingscon'itions' d . Consequences of an ejected RCCA have net been ' analyzed lbecause of -the -low probabilities. For~some l

. incidents, :because of the _ far-of f-normal conditions, the analysis methods availhble have 'not shown th at , with reliance on automatic

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protection system action alone, the FSAR analyses are bounding. In

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those cases reliance is placed on expeditious operator action. Th e

. operator action points as defined will provide ~ protection for such events.

After performance of Spacial Low Power Test Programs at North Anna and Sequoyah, Westinghouse has determined that use of : ., exit th ermo-

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couples and wide range loop RTDs are acceptable for determination of margin to saturation-temperature under natural circulation flow condi-tions. This determination was based on comparison of the average of die core exit diermocouples to the average of the wide range loop RTD's T. H It was found in bcch cases th at cne comparison resulted in agree-ment t o wi th in 1 F. A further comparison was made between full core, incore flux map assembly F 3g values and die core exit th e rmocoup le readings, fRiis comparison resulted in the conclusion that the tempera-ture distribution indicated by diermoccupies agreed reasonably well with the power distribution indicated by - the flux map. Based on the above, Wes tinghouse has concluded that core exit thermocouples and wide range RTDs are reliable-means of determining margin to saturation temperature, die diermocouples for transient and equilibrium condicions and die RIDS

'or equilibrium and slow transient conditions.

f During performance of cooldown with the reactor critical, data was taken to determine the excore detector response as a function of vessel down-comer temperature. In both plants the error in indicated power, intro-duced by die decreasing temperature, was les s th an 0.5%/1'F. This is less dian half the error assumed in the Special Test accident analyses.

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2.0' DESCRIPTION'0F TESTS:

J2.12' NATURAL CIRCULATION TEST'(TEST-l'~- TP/1/A/2150/20)

.0bjective To : demonstrate the capability 'to~ remove decay heat by

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. natural! circulation.

Method - Dae - reactor , is f at approximatel'y.3% power and all Reactor Cool-ant Pumps (RCP's) are operating. All RCP's are tripped. simultaneously with die establishment of natural circulation indicated -by the core exit thermoccuples ' and ' the wide range RTD's.

2.2 NATURAL CIRCULATION i'ITH LOSS OF PRESSURIZER HEATERS (TEST TP/1/A/2150/20)

Objective 'To' demonstrate the ability 'to maintain natural circulation and saturation margin' with the loss of pressurizer heaters.

Method

• Establish' natural. circulation as in Test 1 and turn of f die

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pressurizer heaters at the main control- board. ' Monitor . the sys tem ' pres-

.sures to determine; die effect on saturation margin and the depressur-

-ization rate.

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-2.3' NATURAL CIRCULATION AT REDUCED PRESSURE (TEST 3 - TP/1/A/2150/20)

Objective .To'demonstrcte the ability to maintain natural circulation

-at.. reduced pressure and saturation margin.- Use accuracy of die saturs- l tion. meter will also be verified.

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Meth od - Th e ' te s t method is . die same as for Tes t 2, with die exception

'th at. the pressure 1 decrease can- be- accelerated with the use . of auxiliary-pressurizer! sprays; The' saturation margin will be decreased to approxi-mately 20 F. D'emonstrate the effects of charging / letdown .flov' and

steam generator} p'ressure1 on? the saturation margin.

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.2.41 NATURAL'CIRCUIhTION WITH'SIhULATED LOSS OF OFFSITE'

'AC' POWER'(TEST 41- TP/1/A/2150/23)'

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/ Objective ~ -JTo' demonstrate th'at' following a loss' of zoffsite AC power, J natural ~ circulation can be established and maintained while beicg-powered from' the emergencyi diesel > generators.-

' Method - The' reactor' is at approximately.1% power and all RCP's are i . . <

L - ' operating. All_RCP'slare tripped and a station blackout is simulated -

( AC' powerlis returned.by. the diesel generators and natural circulation is

' verified.

2.5 EFFECT-OF STEAM GE'NERATOR' SECONDARY SIDE ISOLATION s

0N NATURAL / CIRCULATION (TEST 5~ TP/1/A/2150/21) i Objective' - To . determine . the ef f ects of ' steam generator secondary side isolation on natural circulation.

. Method - Establish natural' circulation conditions as in Test 1 but at 1%

power. Isolate the feed rater and steam line for one steam generator and .

p J es tablish . equilibrium. Repeat this ' for one. more steam generator so th at

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. two' are isolated and e'stablish equilibrium. - Return the steam generators to i se rvice, in~ ~ reve rs e . order. .

. 2.6'. SIMULATED-l'OSS-~0F ALL'ONSITE AND.0FFSITE AC POWER

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- (TEST'6 - TP/1/A/2150/26) 1 Objective - To demonstratefthat'. following - s loss of all ~ onsite and'

'o f f site : AC power, including the' emergencyg diesel generators, th e . decay

' heat : can be removed by. using ' the t auxiliary feedwater ' system in the.

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c manual; mode.

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J Method = -lTaeI reactorfisi shut: down"and ; alllRCP's 'are running. 'Seleeted equipment;w'illibeitripped to" simul' ate a station blackout. Instrument,

, ; power is?providedTby: the. backup l.

batteries ~ lsince .the - diesels are Lshutdown.

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l 2. 7 - l ESTABLISHMENT OF NATURAL CIRCULATION FROM STACNANT CONDITIONS I =

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Westinghouse does n'ot believe that it is advisable to perform this - test

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-as ' noted 'in a l letter from .T. M." Anderson, Wes tinghouse, to H. Denton,- .

.NRC, NS-TMA-2242,-April 29, 1980.

' 2 .' 8 FORCED-CIRCULATION COOLDOWN

- This test is performed as preparation for th'e Boron Mixing and Cooldoin Test. Since Westinghouse does not believe it is advisable to perform "th e ' Boron Mixing Tes t as define'd using core heat , it is not necessary to l perform the Forced' Circul'ation Cool'down Test.

2.9~'30RON MIXING AND COOLDOWN Wes tinghouse does ' not believe that' it is ' advisable to perform this test utilizing core heat -as noted in NS-TMA-2242, T. M. Anderson,

.Wes tingh ouse , to H. Denton,_NRC.

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3.0 IMPACT ON PL\NT TECHNICAL SPECIFICATIONS In the evaluation of the proposed tests Westinghouse has determined that twelve technical specitications will be violated, and dius require exceptions, during the performance of the tests. Table 3-1 lists th e technical specifications diat ,will require exceptions and die tests for 4tich diey will not be met. The following notes the reasons th e se specifications must be excepted and die basis for continued operation during the tests.

3.1 IMPACT

SUMMARY

3.1.1 T.S. 2.1.1 REACTOR CORE SAFETY LIMITS The core limits restrict RCS T,y as a function of power, RCS pressure (pressurizer pressure) and loops operable. Th ese limits provide protec-tion by insuring that die plant is not operated at higher temperatures or lower pressures than those previously analyzed. The core limits in die McGuire tech specs are for four loop operation. Obviously Mien in natural cibculation with no RCP's running diese limits would not be

' met. However, it should be noted that d' e tests will be performed with limits on core exit temperature (< 610 7), T avg ( < 590*F) and Loop AT (< 65 F) such dia t no boilidg will be experienced in the core and the limits of specification 2.1.1 f or temperature will be met.

The limits . will not be met simply because less dian four RCP's would be running.

3.1.2 T.S. 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip System provides protection ~ from various transients and f aulted conditions by tripping the plaa.; When various process parameters exceed their analyzed values. When in natural circulation two trip

. functions will be rendered inoperable, Overtemperature aT and Over-power ST. There is a temperature input to these functions wh ich. ori-ginates f rom the .RTD' bypass loops. Due to the low flow conditions, 5%

or less, de temperature indications fres these loops will be highly 3-1 w- - _

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suspecti To prevent the inadvertent tripping of the plant den in the

> natural circul'ation mode ths.se functions will be bypassed. Their pro-

tection functic ns will be performat by- the ' operator verifying that Pressurizer Pre ssure and Level,, Steam Generator Level, and subcooling margin.(T,,gl are above the operator action points for Reactor Trip and> Safety Injection.

Steam Generator, Level-Low -Low' is the' third trip function that can be affected. Wh en at low power . levels it is.- not uncommon for this function

.co be difficult to maintain .above the trip setpoint. This function assures that there is some volume of water in the steam generators above the tops' of' the U-tubes to maintain a secondary side heat sink. Th e amount' of water is based on the decay heat present in the core and to prevent dryout of 'the. s team generators. With- the plant limited to 3:

RTP or less'and being at BOL on Cycle 1 there will be little or no decay heat present. The heat source will be the core operating at the limited

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. power level. Tripping'the reactor on any of the different operable trip

- functions or' the operator action points will assure that this require-ment will be met. Thus, Westinghouse finds that it is. acceptable to lower the tirip setpoint from 12% span. to 5% span' for all of the special tests. In addition, the ' steam generator low-level setpoint sich is part et th e steam / f eedwater , mismatch alarm may be lowered to 5% span.

3.1.3 T.S. 3.1.1.3- MODERATOR TEMPERATURE COEFFICIENT

The Moderator Temperature Coef ficient is limited to O pcm/ F or more negative. When _ performing tests with the plant critical below 551 F this coef ficient may be slightly positive. However, it is expected that the .Isoth ermal: Temperature -Coef ficient will remain negative or approxi-mately zero. The: tests will be performed such that' this is the case and thus minimiting. any i= pact from rapid heacups 'or cooldowns. In addi-I tion, the' ef fect of fa~s~ mall positive Moderator ' Temperature Coe.f ficient-has; been considered in the ,accidentf analyses performed - for the test cond i tions'. I i9:

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3.1.4 T.S. 3.1.1 ~. 4 J MINIMUM TEMPERATURE ~FOR CRITICALITY

' The-Minimum Temperature for Criticality is limited to 551oF by spec.

321.1.5 and 5410F by spec. 3.10.3. To perform test 5 it is, expected .

that. the -RCS average temperature will, drop below 5410F. Wes tingh ouse

- has ' determined that . operation: with Tayg as low as 485 F'is accept-able assuming that: '

1. ' Control B'ank' D i~s ' inserted ' to no draper than 114' s teps with' drawn, and

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2.- Power . Range Neutron. Flux Low Setpoint and Intermediate Range Neutron Flux reactor trip setpoints are reduced from 25% RTP to 7% RTP.

This~ will considerably reduce the consequences of possible transients by 1)' reducing individual' control rod worths (Bank D) on unplanned:with-drawal, 2) reducing bank worth (Bank D) on unplanned ' withdrawal, 3) maximizing reactivity insertion capability consistent with operational

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requirements, 4) limiting maximum power - to a very Icw value on - an

-unplanned power . excursion, and 5) allowing the use of -the "at pcwer" reactor trips as back-up trips - rather than as primary trips.

3.1.5 T.S. 3.3.15 REACTOR TRIP SYSTEM INSTRUMENTAT13N The reactor trips noted in Section 3.1.2 will 'not meet the operability requirements'of spec. 3.3.1. Specification 3.3.1 can be excepted for L the reasons : noted in Section 3.1.2 of this evaluation.

3.'1.6 T.S. 3.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION

'To' prevent inadvertent Safety Injection and to allow performance -of the

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speciali tests, all automatic -Safety-Injection functions will be U blocked. Indication of partial Safety Injection logic trips. for the non- defeated' channels ' and manual initiation will be operable, however, the: automaticisafety Injection. actuation functicns. will-Lb'e made

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inoperable by forcing'the logic to see that-the reactor trip breakers i

are lopen.y Westinghouse believes that this mode of' operation is accep-table:for the short peri'd o' oO time these tests will be. carried out based

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on ~ the ' f ollowing:

-1. Close observation of the partial trip' indication by the operator, 2.' Rigid adherence to the operator action points as defined by t West-inghouse,'see Section 3.2.

3 .~ Little or no-decay heat is present 26 the system .thus Safety Injec-t

' tion serves primarily as a pressurization function.

' Blocking these fu'nctions will allow the performance of these tests at low power, pressure, or temperature and close operator surveillance will assure initiation of Safety Injection,-if required, within a short time period.

Lowering the automatic auxiliary feedwater start-will have little effect, since there is little or no ' decay heat present. Close operator surveillance will insu're auxiliary feedwater addition if necessary.

3.1.7 T.S. 3.4.4- PRESSURIZER The. Pressurizer provides the means of' maintaining pressure control for the L pla'nt . . Norma 11y' this isL accomplished through the use of' pressuri:er heaters andJspray. In sever'a1 tests; the pressurizer heaters will be either. turned off or rendered-inoperable by. loss'of power. 1 mis mode of' operation is acceptablerin that pressure' control-will=be maintained through the use of pressurizer Lleve11and. charging / letdown ' flow, i3;1'.8~T.S. 3.7.1.2 LAUXILIARY FEEDWATER SYSTEM.

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The auxiliary' feedwater' system will,be rendered partially inoperable. for j two tests. IIh'e1 two testsisimulate some form of loss of AC : power, i.e. ,-

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. motor driven auxiliary feedwater pumps inoperable.. Wes tinghouse h as determined tnat this is acceptable for these - two tests because of ths -

little or no decay heat' present -allowing suf ficient time (e 30 =in-utes) for operating personnel to rack in the pump power supplies and regain steam generator level.

3.1'.9..T.S.'3.8.1.1, 3 3.2.1, 3.8.'2.3' POWER SOURCES These . specifications are outside Westinghouse control, however it is

acceptable to alter power source availability as long as manual Safety Injection is operable and safety related equipment will function d en required..

3.1.'10 T.S. 3.10.3 S~tECIAL' TEST EXCEPTIONS -1 PHYSICS TESTS This specification allows the ~ minimum temperature for criticality to be

as' low as 541 F. Since it is expected that RCS T avg will be taken

as low as-'485 F t$is specification will be excepted. See Section ~

3.1.4 for basis of acceptability.

-3.1'.11 TECHNICAL' SPECIFICATIONS NOT EXCEPTED

'ahile not' applicable at. power levels below 3% RTP the following tech- ~

- nical specification limits can.be expected to be. exceeded:

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3.2.2 : HEAT FLUX HOT' CHANNEL FACTOR - qF (Z).

At low' temperatures and flows Fq (Z)'can be expected to be above normal ' for 5%l RTP with ' RCPs running. However at such a low power levelano significant deviations in burhup or Xe peaks are expected.

2. 3.2.3 gNUCLEARLENTHALPY HOT. CHANNEL FACTOR ~- F 6H At: low temperaturesLand flow .F6H can. be expected' to be higher m

'trian if pumps - are running. However, no significant consequences for

full power operacion are expected.

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3. _3.'2.4'- QUADRANT POWER TILT RATIO With no, onei 'two, or three' pumps running and critical, core power distributions' .resulting in quadrant power tilt may form. . AtLlow power levels and ' for .short periods ' of times these tilts will not significantly influence core burn-up.

4. 3.2.5 DNB PARAMETERS' In the performance of several tests the plant will be depressurized below 2230 psia. At low operating power, levels dt is .depressur-i:ation is not significant as- long as subcooling margin is main-tained.

3.1.12 SPECIAL TEST EXCEPTIONS

1. Special Test Exception Specification 3.10.3 allows limited excep-tions for the following:

3.1.1.3 Moderator Temperature Coefficient

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3 .1.1,. 4 Minimum Temperature for Criticality 3.1.3.1 Movable Control Assemblies 3.1.3.5 Shutdown Rod Insertion Limits 3.1.3.6 Control Rod Insertion Limits

2. Special Test Exception Specification 3.10.4 allows limited exception for-3.4.1.1 Reactor Ce:? ant Loops - Normal Operation.

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t, 3.2 ' OPERATIONAL SAFETY CRITERIA

.During the performance of these . tests:. the . operator mus t ' meet the follow-

', ;ing-' set'of.. criteria for operation:

.1. : Maintain ForAll Tests

'a)IPrimary System,Sub-cooling.(T ,g Margin) > 20 F' b) Steam' Generator Water... Level > 25% Narrow' Range Span c)-Prescurizer Water Level "'

(1) With RCPs running- > 22% Span

'(2) Natural Circulation > Value wh en. RCPs tripped -

d). Loop 6T <

_. 65F '

e) T avg -

< 590 F f)-Core Exit Temperature (highest) < 610 F g) Pcwer Range Neutron Flux Low Setpoint' and Intermediate Range Neutron Flux

. Reactor. Trip:Setpoints _< 7% RTP h) Control Bank D 114 steps withdrawn or higher i)'RCS cold temperature ,435

> F

2. Reactor. Trip and Test Termination'must occur .if an'y of the following condi-tions are met:

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's) Primary System Sub-cooling (T,,e Margin) <-15 F b) Steam Generator Water Level. < 5% Narrow Range Span 3

t or. Equivalent Wide Range Level- ,-

c). NIS[ Power Range, 2 channels ' > 10%'RTP

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~ d)-Pressurizer Wate.r Level < 17% Span or an unexplained decrease of more than 5% not

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- e)fAnyLoop-AT >;s5 7 '

0 "f) T avg- -

> : 590 F. l Lg)l Core Exit Temperature (highest) >:610 F-

".h ) Uncontrolled : rod motion -

ib Control: 3ank' D. ?less y th an 11' -

4 L s teps' withdrawn -

l[j ), RCS!'. cold; temperature - . 48'S F~

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(3..SafetyLInjectionLaust'be manually iniciated if any ofL the following condi-

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ti' ens are met:- .

Ta) Primary. System Sub'-cooling--(Tg,g Margin) <-10 F b)" Steam Generat'or-Water Level' < 0% Narrow' Range Span:

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or Equivalent Wide Range Level

' c) Containment Pressure > 1.1 psig I. Ed) Pressurizer Water Levelf

. <:10% Span or an unexplained decrease of more than 10% not' concurrent-with a T,yg _

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-e) Pressuriser! Pressure Decreases by 200 psi. or more in an unplanned or unexplained

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' Safety Injection must not,be termin.sted until'the Westinghouse criteria

as defined in~E0I:E-2, Loss of Secondary Coolant are met.

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' These operating and function: initiating conditions'are selected to j' ~~ assure that the. base conditions for safe operation are met, i e., .

~1. Sufficient marginLto' saturation temperature at system pressure to assure adequate core' cooling (no boiling-in the hot channel),

' 2. : sufficient steam generator level to assure'an adequate secondary l -

.~ side' heat sink,'

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3. . sufficient level.in the pressurizer to assure coverage;of the.

heaters to. maintain pressure; control,

4. -sufficient controljrod'. worth-to ensure adequate shutdown margin'and

minimize

impact iof: uncontrolled bank withdrawal, and'-

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15. .l'imit maximum possible powerLlevel'.in.the event of an' uncontrolled

.. power . increas e .

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J TABLEc3-1 2 TECHNICAL SPECIFICATION IMPACT l

Test Techni-cal :Specificacion - 1 2 3 4- 5 .- 6 2.1.1 . Core! Safety. Limits ,X :X X X X

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l-L , 2. 2.1 ~. -Various;. Reactor Trips Overtemperature AT X X X' X X X

!! 0verpower AT X X- X X X X Steam Generator Level -X X X- X -X X

~ 3'.121.3. Moderator Temperature Coef-- X

~ficient 3.'1.1.E -Minimum Temperature for. X-

Crit *cality.

3. 3 '.1 ' Various' Reactor-Trips.

Overtemperature AT X X: X X X X 0verpower AT- X X X- X X X

. Steam Generator Level X X X X X X 3'. 3 . I Sa'fety Injection LAll X' X X X X X sutematic functions Auxiliary feedwater i automatic; start X' X_ X X X X 3.4.4  ! ressuricer X X X

- 3.7.1.2 Auxiliary Feedwater X X-

- 3'.8'.1.1- AC Power Sources- X X-

-3.8.2.1 'AC OnsiteLPower Distribu- X X tion ' Sys tem '

3.8.2.3 . DC Dis tribution - System X X 3.'10.3 I Special Test Exceptions - X- l

' Physics Tests i l

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.4.0 . SAFETY EVAL.UATION

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. In this ~ section ;the: safety effects of those'special test conditions D

4tich 'are: outside the bounds. of conditions ' assumed in the FSAR 'are evaluated. The interaction of these conditions with 'the transient analyses'i'n the FSAR' are discussed. 7 '

L-t '4.1 EVALUATION'OF TRANSIENTS ,

The effect of the unusual operating conditions on the transients

! analyzed ;in the FSAR are evaluated.

~

-4.1i1 CONDITION II: -FAULTS OF. MODERATE FREQUENCY-t p

'4.1.1.1 Uncontrolle'd' Rod Cluster Control Assembiv Bank Withdrawal f ros

!. ~ a'Suberitical Condition

-Restriction of controljrod operation to manual control, and constant

! operator monitoring of.. rod position, nuclear power and-temperatures greatly red'uces the: likelihood .of an' uncontrolled RCCA withdrawal, i

l

. Operation without : reastor _ coolant ; pumps, and in some cases with ' a posi-

~

L tive moderator temperature reactivity coefficient, , tend to make the consequences.of'RCCA withdrawal worse compared to .the operating condi-tions- assumed in the FSAR. For these reasons- the operating procedures require ? that following'any reactor trip at least one reactor coolant pump will be restarted' and the reactor boron concer tration will be -such-

~

that it. will not go; critical with less Jthan 114~ ste's withdrawal'on D Bank. An. analysis of thisievent is presented.in Section 4.2.1. For.

Tes t :6, Ithis : transient ,is . bounded ,by' th e' FSAR ' analysis ,. since all reac-tor.coola~t;pompsnar'n sperating'.

i'

4il .'l. 2. Uncontrolled - Rod" Control ' Cluster Assembly Bank '4ithdrawal. at -

' Power-

- Thebsame considerations ! discussed in' Paragraph 4.1.l'. l~ apply here. .In

[, s addidion,f the.itow- operating, power iand th'e'. Power .Ranbe Neu tron Flux Lov m

.and(Intermediatef Range.; Neutron Flu'x !. trip -setpoints act to mitigate this 4 $ '

, 5 }

g is .

W - , :/

w- . -

n . - . :.-

-~

. ,_ , __ l- '_,

_, = c.--

3 incident', while lack ~of the Overtemperature LT trip removes some of the protection provided in the FSAR case. An ' analysis is discussed in Paragrsph 4.2.2.

4.1.1.3 fred Control Cluster Assembly Misalignment Thel FSAR discussion concerning static RCCA misaligement applies to the test conditions.. The consequences of a dropped RCCA would be a decrease in power. t.us no increase in probability or severity of this incident -

7 is introduced by the test conditions.

l 4.1.1.4- Uncontrolled 3eron Dilution The consequences of, and operator action time requirements for, an uncontrolled boren. dilution under the test conditions are bounded by

, those discussed in the FSAR. The fact that the centrol-rods will never be in's erted to the insertion linics, as well as the Power Range Neutron Flux Low Setpoint and the constant operator monitoring of reactor power, i temperature and charging system operation, provides added protection.

4.1.1.5 ' Partial Loss of forced Reactor Coolant Flow 4

3ecause of the low pcwer Ibnits the consequences of losalof t '*ctor coolant pump power are trivial; indeed they are' bounded by normai opera-ting conditions for.these tests.

4.1.-l . 6 'Startup of an Inactive Reactor Coolant Loop i

When at least one reactor coolant pump is operating, the power limit for these-tests results.in-such:small teoperature : differences in the reactor coolant system that startup of another loop caunot' introduce a signifi-

~

' tant peactivity disturbance. Inl natural circulation-operation,-inadver-tentiscartup of a. pump would-- reduce the core water. temperature and thus provide a : change in. reactivity and power. Because of the small modera-

'cor: reactivity coefficient at beginning 'of . life .the power increase in.

.the worst condition would

be small and gradual and the flow-to-power m

.~.._, _ . _ - -_.---..~m-----__~.w.--.

1. _" . _ - . . ..

I ratio in the core vculd be' increasing.

The Power RangeINeutraa Flux ~ Law.

~

l

' Setpoint< reactor - trip provides . an upper. bound on power. Because of the i increase in flow-to power ratio -and -because of 'the low setpoint on the reactor trip,-- DNB is precluded . in' this transient.

4.1.1.7 Loss of External Load and/or ' Turbine Trip -

3ecause of the -lowLpower level,~ the disturbance caused by any Ivas of

. load is small.. The FSAR case is bounding.

l 4'. l .1. 8 Less of Normal Feedwater L

3ecause of the low power leve'1, the consequences of'afloss of feedwater are bounded by the FSAR case. . In the case of loss of all feedwater sources, if the reactor is not shutdown manually, it would be' tripped on Low-Low Steam Generator Water L'evel. Ample tLae is available to rein-stitute auxiliary feedwater sources.

4.1.1.9 ' Loss of Offsite Power to the Station's Auxiliaries (Station l

' Blackout)

Because:of the' low power. level, the consequences of a loss of off-site power'are bounded by th'e FSAR case.

~

r 4.1.1.10 Excessive Heat Removal Due to Feedwater System Malfunctions 1

The main feedwater control valves will not be used 'while the reactor is

~

at pover n or:near criticality on'these tests' . Thus, _the potential water.

flew .is restric ted' to the' auxiliary Efeedwater flow, about 6% of normal t ~

~-ilow. . The transient _ is further mitigated lyr the low operating power

-level, small moderator temperature reactivity coefficient. the low set-points Ton the' Intermediate 'and Power - Range Neutron Flux Low setpoint-trips,1and'close? operator surveitL3nce of feed flow, RCS temperatures, RCS pressurei and - nuclear; power. The case of excess heat removsl.due to

'feedwater system malfunctions with very low rsaccor coolant flow is among .the Jcooldown transients ; discussed in more detail. in Section 4.2.3.

4 1

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3. . . .

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4.1.l.11 Excessive Load Increase Incident

- The turbine will not be'in use during1 he t performance of these tests,

-and load. control will be' limited co operation of a; single steam dump or steam ' relief valve.1 The small moderator temperature reactivity coeffi-cient- also reduces the consequences 'of this transient. Close operator surveillance o.. f steam pressure,' cold leg temperature, pressurizer. pres-sure, . and reactor power, with specific initiation riteria for manual

reactor trip, protect

against~ an unvanted reactor power increase. In

- addition,: the ' low setpoints for Power-Range and Intermediate-Range Neu-

- tron Flux reactor trips limit any power transient. Analyses are discussed in Section 4.2.3.

4.1.1.12 Accidental Depressurization of the Reactor Coolant System CloseToperator surveillance of pressu'rizer pressure and of hot leg sub-cooling, with s'p ecific initiation pointe for manual reactor trip, pro-

. vides sprotection 'against DNB'in the event of an accidental depressur-

~

1 ization of ' the RCS'. ;In addition, automatic reactor trip caused by the -

Low' Pressurizer Pressure Safety Injection signal veuld occur when core outlet. subcooling readied ap'proxiaately 250F as an automatic backup for manual trip.- ' During test 2. and ' 3,- when this ~ trip is bypassed to allow. deliberate operation'at low pressure, th'e pressurizer PORV block valves will be closed to remove the major credible' source of rapid inadvertent depressurization. (Th'e Low Pressure trip is automatically L

reinstated 'when; pressure goes above -1955 psig and the PORV block valves will . be : reopened 'at . that' time. )

~

4.1.1.13 ' ' Accidental' Depressurization of the Main S team System The FSAR analysis, for . accidental' steam system depressurization ' indicates thatlif the transient starts at. hot shutdown ~ conditions with the worst RCCA stuck out of the core, the ' negative" reactivity introduced by. Safety.

.. Injection prevents .the core from' going. critical. - Because of the small moderator' te=perature reactivity coefficient which will--exist 'during1the-f r w

Y i

^ '

j-4}

ii a il

. . . . -- g,- 7

.: . .- g. ,

l l

test per,ied,-the reactor-vould remain'suberitical even if it were cooled co. room tenperature without Saf ety Inj ection. -.Thus the SAR analysis is

. . bounding.

- 4'l.1.14 -Sourious Operation of the Safety Injection System at ?ower

. In order co reduce the possibility of ~ unnecessary thermal f atigue cycling of the reactor coolant system components, the ac scion of high head charging in. the safety injection mode, and of the safety injection pumps, by any source except sanual action will be disabled. Thus, the mos' . likely sources of spurious Safety Injection, i.e. , spurious or

" spike" pressure or pressure-difference signals from the primary or secondary systems, have been eliminated.

4.1.2 ' CONDITION III - INFREQUENT FAULTS 4.1.2.1 Loss of Reactor Coolant from Small Ruotured Pipes or from Cracks in Large Pipes-%~hich Actuates Emergency Core Cooling

<~

A review of the plant- loss of coolant accident behavior during the lov power. testing-sequenc's indicates that without automatic Safety Injection there is sufficient cooling water readily available to prevent the fuel rod cladding fran overheating on a short term basis. The systen inven- l 1

tory and normal charging' flow provide the short tera cooling for the  !

l smalltbreak transient. Aisample calculation for a 2 inch break shows  ;

l that the core remains covered for at- least 6000 seconds. This is suf- 1 4

ficient time for the operator to manually initiate SI and align the system for long cers coolin'g.

4 It nus t be noted that the =sgnitude of the resulting clad heacup tran-sient during a LOCA event' from these conditions is significantly reduced

-from~.the-FSAR basis scenario.-by the' low decay heat and' core stored energy resulting: from the low power level and short operating history.

< l i

4-5 '

l  !

NOS h _ -

I x .,s- , , , , .,

~

_@_ ____~ _. _ _ _ . .

. -e 4.1.2.2 . Minor Secondary System Pipe 3reaks The consequences of- sinor secondary system pipe breaks .are within the bounds discussed in Paragraph 4.2.3.

4.1.2.3 . Single Rod Cluster Control Assembly Withdrawal at Power T'

The FSAR ' analysis shovs that assuming' limiting parameters for normal operation a maximum of 5 percent of the fuel rods could experience a DNBR of less than 1.3 following a single RCCA withdrawal. As the 75AR

~

points out, Jno single electrical or mechanical failure in the control system coul'd cause such an event. .The probability of such an event happening during the . test period is further reduced by the short dura-tion of this period,-by the restriction to manual conrrol, and by the close operator surveillance of reactor power, rod operation, and het leg temperature.

4.1.2.4 Other Infrequent Faults The consequences of an inadvertant loading of a fuel. assembly into an

improper position, complete loss of forced reactor coolant flow, and waste gas - decay' tank' rupture, as described in the ?SAR, have been reviewed and found . co bound the consequences of

such events occurring during test. Operation.

14.1.3 ' CONDITIOP: IV - LIMITING FAULTS

-4.1.3.1. Major Reactor' Coolant Pipe Ruptures'(Loss of Ccsir.4t Accident)

A review of the plant loss of' coolant' accident behavior during'the low powerftesting' sequence indicates that without automatic safety injection there is ~ sufficient' cooling water . readily ' available to prevent- the fuel rod : cladding 4 from over heating- on - a sher't term basis. During the large l

. break event the ' system: inventory - and cold - leg accumulators will' have I

t removed enough:. energy cof have : filled theJ reactor vessel t to ' the bottem 'of

~the no::les. : Following the' system depressuri:ation there is enough 6 _

a;. _ .---  :

ggg., _

'Q g. 2 m ~- .e ea . .. v* -m

+

- water _ in~- the react'or :ves'sel below the nozzles to keep : the core . covered -

~

ifor ov,er 'one hour .using conservative assumptions'. . This is suf ficient

-  : time for the' operator cosmanually initiate SI :and align the'sys tem -for-long term?co'oling. At no -time during Lthis transient ~will the core be uncovered.

~It must! be' noted;that ' the magnitude 'of the resulting clad heacup tran-

sient during a LOCA ' event f rom these conditions is - significantly reduced from. the FSAR' basis scenario by tne low decay. heat and core stored energy resulting from 'the low power level and short operating history.

(4.1.3.2- Major Secondary System Pipe Rupture The small1 moderator egmperature reactiv' ty coef ficient, close operator

. surveillance of. pressurizer-pressure, cold leg temperature, and. reactor power, with l specific initiation criteria for reactor trip; low trip setpoints on die Intermediate-Range and Power-Range Neutron Flux trips; MSIV closure on I'ow' Steam Pressure: and Low Pressurizer Pressure trip

~

(S.I. .initi Ation); assure a Reactor Trip. without excessive reactor pcwer

'following a1cooldown' transient caused ;by- the secondary system.

.-Follcwing reactorr trip,. assuming 'the worst RCCA stuck out of the core, othe: reac tor would : remain subcritical- even if it were cooled to room temperature.- Transient anslyses for a steam pipe rupture ar'e provided .

in Section 4.2.3. . .Tae consequences of. a main feedline rupture are b'ounded in the cooldown di'rection.by the steam pipe rupture discussion.

~

.Because of the Elew operating power, the heatup aspects -of a feedline rupture 'are . bounded by the :FSAR discussion'.

4.1L3.3il Steam ' Generator Tube Rupture The1steamigeneratorLtube rupture' event may-be-categorized by two dis-

~

' tinc t < ph as e s ' . The Jinitial? phase of : the " event is analogous to 'a 'small __

LOCA event. Prior ~ co operator-controlled. system Ldepressurization',- .the

, _ steam' generator; tube rupture isfa.special class!of small break-LOCA 2

s

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k - 4

, ~

L, .

.  ;;4 7 '

. 19 1 ,

1,,n d 3 2. ;

9,

. w_ :,,.. . . , , .,

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...,--n

... ,.x ..- . -. . , .,--. , . : ;. . _  ; -m? e.__:.- . , _ a ,.

j- ._ ,: . - . . . _

< t f

transients, and the operator actions required co f deal 'with' this' ' situ- '

~

aci~on,during;.this phase are identical; to those required for mitigation ofIa smalllLOCA.= ' Hence, evaluation'of-ithe' steam generator tube rupture -

. '.during ' thi's ' phase! is wholly covered by the' safety evalua tion of "the-

. small LOCA..  !

L

' Af ter the ippropriate' operator actions have taken place to deal- with the

initial'LOCA phase of ' the: event, L the remainder of the steam generacor-

tube. rupture laccidentimitiga' tion would consist of .those operator actions i required ' to isolate the faulted steam generator, . cooldown the RCS, and

~

i

.depressuri e the RCS :to . equilibrate primary RCS' pressure with the f aul ted -.'s t'e'am generator - secondary pressure. These actions require util-

'izati~on of ' the' following 'sys tems:

e.

Auxiliary' feedwater control to the f aulted steam generator.

1.

2. Steam line . isolation of the faulted steam generator.

.3.. Steam relief-capability of at least one non-faulted steam generator.

t 4 RCS-depressurization capability.

1" Evaluation of the McGuire special t'est -procedures has verified that all of the 'above . systems are immediately' available for operator control from

~

the . control' room. Th e re f ore , f. i t is concluded that the ability to miti-gate the1 steam generator' tube rupture' event is.not compromised by the 2

- modificationsirequired for operation at 5% power' during' th'e proposed -

- l tests,: and . that' th'e Lanalyse.s' performed for the S'AR regardingi this event remain; bounding.~-

4.153.4L'5inglelReactor' Coolant Pump Locked Rotor

, . .. . i

+

- 3ecauseJof L the -low powerf. level, the locking ~.of a- single' _ reactor coolant - .i y ' pump rot'or is" inconsequential. .

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4.1.3.5 Fuel Handling Accidents 21e'.FS'AR ; analysis lo'f fuel handling accidents is bounding. -

g .4.1.3.6. Rupture of' a Control Rod Drive 'Medsanism Housing (Rod" Cluster Control'-Assembly Ejection)

Ihe control rod bank. insertion will be so . limited (i.e' , only Bank D .

inserted,' with at least 114 steps withdrawn) ' that the' worth of an ej ec-ted. rod will be substantially less than the delayed neutron fraction.

Thu s , thel power rise -following a control rod ejection would be rela-tively gradual and' terminated by the Power Range and Intermediate Range Ne'utron Flux reactor trips. While the. core' power transient and power

~

distribution following an-RCCA ejection at this time would be less severe than those shown in -the FSAR, the result' of combining these ameliorating effects with the 'effect of the natural circulation flow rate' ~en clad-to-water h eat transfer and RCS pressure have not been

~

analyzed. The extremely low probability of an RCCA ejection during this brief period in the test sequence does 'not warrant such an analysis.

4.2~ ANALYSIS OF TRANSIENTS 4.2.1 ANALYSIS 0F RCCA BANK WITHDRAWAL FROM SUBCRITICAL CONDITION s

l' An analysis was . performed to bound the test transients. The methods and assumptions used 'in the FSAR, Section '15.2.1 were used with the follow-

-ing exceptions:

1. Reactor coolant flow was.0.1% ofinominal.

2. = Control ' rod incremental worth and totallwords were upper. bound values for Ethe .D bank initially ~114 stepsL withdrawn.

c 1

~

'3. ' Moderator : temperature reactivity coaf ficient was an upper bound 1 (positive). for anyDcore average temperature -at or- above 4850F. .

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4-9 705 2'A ,

..~,..a. . .L - . , . - . . .: .. - .. - .. _. " - , .

.,- .~ .., i.L. ..l I. ?

~ ^

4. Th'ellower beund foricotal delayed neueren-fraction for.the beginning 1

of life for. Cycle l'was use'd..

5. LReactor. trip ~ was' initiated at 10% of full power.

6. DN3 was . assumed .co occur spontaneously at the hot spot, at the Lbeginning.of the transient.

' The resulting nuclear power peaked at 65% of full' power, as is shown in Figure'4.2.1.' .The peak clad temperature reached was under 13000F, as

-is shown in Figure 4.2.2. No ' clad failure is expected as a result of this transient. .

J 2

4.2.2 ANALYSIS OF RCCA BANK WITHDRAWAL AT POWER 1

Analyses of RCCA. bank withdrawal transients were performed for natural ciret'.ation conditions. The transients were assumed to start from steady-state operating conditions at either 1% or 5% of full power, and

. with either all steamline isolation valves open 'or. two of those valves

. closed. A range .of - reactivity insertion rates up to the maximum f or two banks' moving was assumed for' cases with all steamlines open, and up'to

't he: maximum f or. one bank moving for the cases ~ with two steamlines iso-

~

l laced. Both maximum and minimum bounds . on reactivity feedback coeffi-cients f or. beginning of life, Cycle - 1, we're investigated. - In all cases ,

.-reactor-trip:vas initiated at 10% nuclear power.

R Reactor. conditions at:the-tLae-of. maximum core heat flux are shown in Figures 4.2.3 -and 4~.2.4 as. functions of thel reactivity insertion rate

for three four-loop active cases. . For high reactivity insertion races, aha -inimum; reactivity-coefficient-cases'give the greatest heat flux trip aetpoint' M is reached, ..and .have the lowest . coolant flow

~

j' 1 rate:st1the-time of peak. heat flex. 'For these cases even the slowest h

insertionlrates. studied did noti result .in ~ any increase in core inlet

~

temperature Jat . the : time fofL peak, heat . flux.' For maximum 1 feedback cases,.

however,Jthe transients . for

very -low insertion rates go on f or so long E

4

+ 4 4 4-10' .

fr- he -r 1 y w -

g g  ? v- -Y f 6 *y9y a M'- $di f' T N $

a. ,._._,

r ch'at he. core' inlet temperature finally increases before trip, i.e.,

af ter ' approximately one and one-half minutes of . continuous withdrawal.

Thus, the cases shown bound :he vorst cases; 4.2.3 ANALYSIS OF C00LDOWN TRANSIENTS Cooldown transients include. feedwater system malfunctions, excessive scess load increase, accidental'depressurization of the main stesa sys-tem, and minor and major secondary system pipe ruptures Attention has been focused en the possibility and magnitude of . core power transients resulting from such cooldowns before reactor trip would occur. (Follow-ing reactor trip, no cooldown . event would- return the reactor :o a. cri- ,

tical condition.)

During natural circulation operation, approximately one to two sinutes vould elapse follcwing a. secondary side event before cold water from the steam generator reached :he core; thus, considering~the close and con-stant surveillance during these tests, time vould be available for the operator to' respond to such an event. Analyses were also performed to determine the excent ofl protection provided by automatic protection systems under. trip condi: ions.

M 4.2.3.1 Load Increases A. load increase or a small pipe break, equivalent to the opening of a single. power-operated steam pressure relief valve, a dump' valve., or a safety valve', would cauce an increase of less than four percent in reac-ter' power, with a corresponding increase in core flow with natural cir-culation, assuming :he-bounding negativa moderator :emperature coeffi-eient for the beginning of life, Cycle 1. ~Thus no automatic protection

. Lis; required,:and ample time.is available to the' operator to trip.the

' reac:or, isolate feedvater to the 'f aulted steam generstor, and isolate

.the break-to :he extentLpossible. Calculated results for the sudden opening of-a. single steam. valve, assuming the most negative SOL Cycle

.one moderator reactivity coefficient and 57. initial power are shown in Figure s . 4. 2.5 ' and 4. 2. 6.  ;

i

)

1 F

7- .4-11'

~_ r - , y e

n.~.._.... . . . _ . . . - _ _ , . . . .. . . . . _ . . , _ __ ___ _ _ _ . . _..._%..._._._,.

, .s .

4.2.3.2 Migh Flux ?rotection Reactor trip on high nuclear flux provides backup protection for larger pipe breaks or load increases. Analyses were. performed to detensine the worst core conditions-that could prevail at the time of high-flux trip, independent of the cause. The following assumptions were used:

1. Upper-bound negative moderatcr isothermal temperature coefficient, vs. core average temperature, for beginning of life, Cycle 1.

i l

2. . Lower-bound fuel temperature - power reactivity coefficient.

i i

3. Initial operation vich core inlet temperature 555o7.

~

l-

4. ' Initial powers of 0% and 5% of full power were analyzed.

I

5. Hot leg coolant ac incipient boiling at the time of reactor trip.

This results in some boiling in the reactor. The negative reactiv-icy intioduced by core boiling would effectively limic power; this negative reactivity was - conservatively neglected.

l i-

! 6... Uniform core inlet temperature and flow.

7. Reactor trip equivalent to 10% of full power at the initial inlet temperature. The power as measured by the NIS.is assumed to be diminished frem the,true power by 1% for each lor decrease in l

reactor inlet temperature, resulting in a true power of greater than I

10%.at the. time of trip.

y 8. - Core flow race as a function of' core power was assumed equal to the r

predicted flew under' steady-state operating conditions.

4.121

%- 4 _

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,_,.a ., _. _

.... , _ m .-

m_ . - - .

%..,%.~..._...__.<_..,u.._,

.y .

• a 4

! Analyses.of core conditions. based on these assumptions indicate that.the-DNB criterion of.the FSAR'is' met.

~ '

?4.'2.3.3; Secondary ~ Pressure Trip' Protection Large steamline ruptures.which' affect all loops' uniformly will actuate reactor trip and steamline. isolation'on Low Steamline Pressure signals lin'any two lines. Low Pressurizer Pressure'and. Power. Range Neutron Flux low setpointLtrips-serve as further backups. An example is the double-ended. rupture of a main steamline downstream of che isolation

~

valves, with all isolation valves initially open. Figures 4.2.7 and 4.2.8 show theLresponse to such an event, with an initial power of.5%'

and natural circulation. :The Low Steamline Pressure crip occurs almost Licmediately. In the example shown, the main steamline isolation valve on loop one was assumed to fail to close. No power excursien resulted,

'and the reactor. remained suberitical after.the trip.

4.2 ADDITIONAL CONSIDERATIONS I

In the great majority of cases it was concluded, either by reanalysia or l

L

.by comparison with previously analy:ed FSAR conditions, that fuel clad I; integrity would be maintained without need-for operator mitigating action. For the LOCA or steambre k eveats, it vas concluded that the L .-

. ope'rator would have more than' ample time (> 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) to respond by l

L manual action,-e.g., manually initiate safety injection, to preclude

~

L l

fuel damage.

i Tinally, in certain other: cases,'primarily associated with certain

i'nadvertent RCCA withdrawal events, the postulated accident conditions b ivare'neither< amenable to direct analysis ~-nor'eredit for operator. inter-l u

vention., min particular,'the postulated accident conditicas were outside the. bounds of acceptedfanalysis techniques so that fuel damage was not l

-l prec1'uded either!by analysis.or idsntified. operator-action. .For these j cases;1the.-basis for acceptabilityLwas primarily ansociated with the low l 1

iprobability :of :an . inadvertent rod'withdrawalievent during the limited duration'offthebspeciall tests.

~

yy, t,

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This section provides an additional assessmend relative to the potential for andiconsequences'of' fuel fail'ure for these "unanalyzed" accident conditionsJassociated.with certain rod withdrawal events. This assess-ment is partially based'upon an attempt to bound certain effects'which-may.. exist for-conditions removed _'from.the range'of' direct model applica-bility.. Additional information (attached) is' provided f or four- areas t l'. ' Thermal' margin associated with normal test conditions.

~

2. The' potential for DNB during accident c'onditions.

The clad temperature response assuming that DNB occurs.

3'.

4. Radiological consequences; associated with presumed gross fuel failure.

.The conclusions of this' assessment are as follows:

1. - DNB is not expected'for the limiting' thermal condition associa-ted with any RCCA withdrawal event. )

i

2. JEven assuming DNB,.there should be adequate heat transfer to

. prevent clad overheating.

3. Fuel clad failure is not expected.

4. :Even assuming 100% clad-failure and other extreme conservatisms, the.resulting offsite' dose would be.small.

4.3.1 ' DESIGN CONSIDERATIONS A

Margin to hotichannel, boiling'has beenzincorporated with all normal test

~

conditions by? establishing a lower bound requirement on'the degree of

. reactor'coolantLsubcooling.- This test requirement assurec that postula-

~

' tediaccidentsl are initiatedifrom a' condition of excess thermal margin.

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7 1., 4~.3.2{<DNB CONSIDERATIONS '

+ '. ~

3

~

(ForJcertain cooldown transients,: the: conclusion that DNB is-precluded ,

was ' drawn b'ased on use; of ithe' W-3 critical heat flux correlation.

~

I :llthough th's: analyses for the cooldown events discussed in section

~

l ^i2.3.2: result in mass. vel'ocity below che range of ~ direct applicability uof the correlation, the' reactor heat flux was so low relative to the predicted l critical'.heatfluithat.even-afactorof2wouldnot result in serious:.concernforb(3forthis' event.

7 For the'non-cooldownLtransients the limiting conditions, with respect to

'DNB, are farther'away from the W-3 range'of applicability because the coolant ~ temperature is higher-and the power-to-flow ratio is larger.

~ ~

t; Comparison A' c'he'W-3 DNB correlation to low flow'DNB test data and correlations -(ref erences 1. and 2) indicate that it will conservatively predict ~ critical heat- flux at- low pressure (# 1000-psi) conditions ,

with low coolant-flow. Pool boiling critical heat' flux values (ref er-t ~

[ ence 3) at these l pressures: are : higher than those predicted by the low-i

. fl ow"c orreh tions . - Further-review of the data.in reference 1 indicates.

that the critic'al heat flux at higher pressure.is'significantly lower than;the'above data at'1000i psi. The minimum critical heat flux of the data set is .16'x 106BTU /hr-ft2 for a' data point at 2200 psia at a -

0

. mass velocity ofL.2 x 10 . Ibm /hr-ft2,_

- 1 Since.-the exit quality for this data point was.64%, it is unlikely that 1

(c 'thelreactor would.be able' to maintain 'a' heat . flux of that level due to i l the' nuclear feedback fromrvoiding. The' power distrib~ution would tend to )

L peak towards the bottom 1thus further reducing the local quality at the I.

. peak;fluxilocati~ons.A '

i' ,

s.

' -  ;,f F , )

+ l

\

" = , -

. L.l;; i ,

l

~

y E I - -

" .a f

T + ., lJ ; g} u e~~ + , .. , ,4v - v+ ,. ,i- - ,3 .- , - - . ,

' ~

..--,-n.- - . . . . - , .

-- --~ L ~_.. _ ~.i.

j- -:-- , * , .J

_7 .,. ,

v- 'Also th'e pool'holling correlations in reference 3 show some decrease in'

^ '

critical: heat -flux. above 1000. psia to the maximum pressure ' of applica-

'bility of 2000 psia. 'However extrapolation oflthe correlations-to a value of zero' critical heat flux at the critical pressure (3206.2 psia)

- would not' result in lover' critical heat fluxes than shown in the' data ,

set.from reference 1 : Since'the core average heat flux at 10% of nom-inalpower(Sighestexpectedpower'forheacupevents)isonlyon.the .

or'erd of'.02 x1106? 3TU/hr-ft2 a large peaking factor would be required.co put the reactor heat. flux as high as the critical heat flux.

For the transients considere'd, the only ones that lead to significant off norma 1 ' peaking f actors .are. rod motion transients. The rod with-i drawal from suberitical is a power burst concern. As such, it is expec-Ced that even if DN3 occurred, the rod surface would revet. For the rod ,

bank withdrawal, the combination of maxi =um power and peak'ing factor would result in a peak power lower than the data referenced ~4'uove.

L Given the lack of data, it is difficult to completely preclude DNB, although a prudent judgementLindicates that it is indeed remote.

! 4. 3'. 3 Cl.AC TEMPERATURE CONSIDERATIONS

l. Should DNB occur, the peak clad temperature reached would depend pri-marily on the local nuclear. transient following DNB and on the behavior. l

of the' post-DNB heat transfer coefficient.

l For a rapid power transient, as is illustrated by the SER analysis for RCCA bank- withdrawal from a subcritical condition, the fuel temperature.

reactivity feedback and reactor trip on a nuclear flux signal would shut down the reactor..before sufficient energy could be generated to cause a damaging: rise.in clad temperature. In that case, the maximum clad tem-1perature. calculated was under-1300 7 even assuming an extremely low

-heat transf er coefficient (# 2 BTU /hr-f 2t - F).

-n A possibly more' limiting condition for RCCA withdrawal would be the case

in.which/aLpowerl increase causes DN3 but would either not result in-reactor . trip on;highLnuclear- flux or the trip is. delayed.. In the former a

m 14-16

[' Wi . ; ,

p _ . . _ _ . a. . -

_ % g. a _ -

7. _. . . ,

o. . - *, - . .

case,' a steady state ' condition with' hot spot DN3 could be postulated.

-In this state the clad temperature'could be calculated given only the total-core power, local-heat flux channel factor, heat _ transfer coeffi-cient_and-saturation temperature.-

The core power is postulated to be' essentially at the power which would cause a reactor trip on high power Range Neutron riux low setpoint. The ,

trip setpoint is at 7% for these tests. To~ allow for calorimetric errors and normal _ system errors,-trip is assumed to occur at 10% of rarad thermal' power (RTP), unless a large decrease in downcomer coolant.

temperature occurs during the test. In tests 2 and 3, depressuri:ation

_to less than approximately 1450 psia could require temperature reduc tion, as is indicated in Figure 4.3.1; however, such low pressures are not expected.

Figure 4.3.2 shows the relationship of peak clad temperature, local heat transfer coefficient, and~the product of heat flux hot channel factor (Fq) times core p'over (fraction of RTP). For the event of an_uncon-trolled RCCA bank or single RCCA the upper bound of this heat flux product-is approximate 1y'O.34. Using this value, the heat transfer coefficient required to keep'the peak clad temperature below 1300 7, the. threshold of significant heat flux increases due to circonium-water reaction, can be found from Figure 4.3.2.

Various film boiling heat transfer correlations have been reviewed to evaluate.the heat transfer coefficient for post-DNB conditions.

Although_no correlations were found'which' cover the complete range of conditions being tested, some data exist which can be extrapolated to obtain representative heat . transfer coef ficients. The. Westinghouse UHI fil's boiling correlation (reference 4), was developed at low flow condi-tions similar to those postulated for incidents occurring during the l McGuire tests. 'This correlation was extrapolated to the higher pressure  !

l

conditions.of the tests to obtain. representative film _ boiling coeffi- ]

.cients. This resulted in a heat transfer coefficient in excess of R

(100-3TU/hr-ft2_op )a,e at 2200 psia and 5". flow with quality l 1

between 10-50%.. Other -- film' boiling heat - transfer. correlations, devel- 1 oped at higher pressures,fwere'also examined. These correlations were

~

J i

I l

17 l i

7052. A .  ;

W v Y

n . -

. .+2~ 9a , ; -

., w -- _ _

_mg3- .

~~~---------

" ~ ~ ~ " ' * ~

. 4 y* ~. - - ~ ':~ 7 ~T " ' ~ " w "

b, ~ ,

extrapolated 'down.to-the' lower'flowLeonditions of the McGuire tests as-U canother approach to obtain representative film boiling coefficients.

Using;b'oth the.Mattson ettal (reference 5) and the Tongp(reference 6) cfilm boiling correlations resulted in post-DNB. heat transfer coeffi-

~

"cients:in~ excess of 150LBTU/hr-ft2 _oF at the conditions given above.

- These results : indicate : that' ia clad' temperature excursion resulting in

. fuel' damage'is;not likely to occur even'if DNB is assumed.

4.3.4:.: DOSE ANALYSIS CONSIDERATIONS-The dose analyses: vere performed for,a hypothetical accident senario-using conservative assumptions sotas to determine an extreme upper bound on postulated accident consequences. The analysis assumed a reactor accident involving no pipe-break.with a coincident: loss of condenser

. vacuum. This accident scenario is representative of the Condition II type. events analyzed in the FSAR. The bounding assumptione made in the analysis include:

170 Mwe (5% power)'

1.0 dose-equivalent I-131 RCS activity (tech spec. limit).

500 spd steam generator leak in each Sc (tech spec' limit)

~

100%-' clad damage and gap activity release .

-10% iodine / noble' gas in gap space 100 DF in steam generators 500. iodine: spike f actor over steady state :

509,000 lb. atmospheric steam dump ~over 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

'1.7 x'10 ~3 sec/m3 X/Q percentile value *

-)

~The results of-lttie. analysis' show that the two hour site boundary-doses would be 5 rem-l thyroid, 0.9; rem total body and 0.4 rem to the skin.  ;

I

The_ analysis of-the accidents has incorporated some;very conservative l

~

2 assumptionsJwhich;goes beyond the normal degree of' conservatism used in F SAR '~analys es . .The,most prominent of~-these assumptions and a brief-

^

Jdescription'ofJthe extreme conservatism includes-

, , ~

I

- 7052A. -

c. - .- . - . .

r._._.-.._

-m.,a  : . - --# % . _ . ._ , - , s,. 4. _ . .-

y ,

t c

1) 'EquilibriumiradionuclideLinventories' established at 5% power. For iodines,7this: requires'r;l month af steady state operation at 5%

t Juninterrupted.

2)1 . Fuel clad;gapsinventories st 10% of core' inventory, this is a time.

dependent,-temperature-dependent phenomena. At 5% power, very o .

.little. diffusion to. gap. space is expected for the short test' period.

~

3) 100% fuel-ro'd clad' damage.

I4) . Primary .to secondaryfleakage - to tech spec values. Since McGuire is

a. new plant, -no primary. to secondary leakage is expec ted. If leakage _ vere present, it vould most likely slowly increase- in steps

~

up to tech spec, levels.

5) Percentile' meteorology, there is 95% probability of better diffusion characteristics and thus-lower =offsite doses. Additionally, the

.fifth percentile X/Q for McGuire isfsignificantly less than the

~

-generic value'used in this analysis.

For these reasons,-inLthe unlikely event of a potential. accident'during the tes ts , the resulting dose .is small, 'even assuming 100% clad damage and other extreme'conservatisms.

• .This is a generic conservative value. representing the worst meteor-ologica1' dispersion characteristics of any Westinghouse nuclear

? plant: site.in the United States.

i 4 y .'L ,

, =

,a 4-19 d c .7 & C ? * ' . ,

+

u._ ~ . . _ ~ .

. ,.. ~_,,. n

-.-- ._.- .; .; - ,.,.. < .y m g ;,.rp:. w :- -

r

-,y_ . . . . , . _ .

p,m .m. . .

m

-4.3.5 ~0THERLCONCERNS- ,

The[LOdA analyses presented indicate that there are^over 6,000 seconds for>the operat'or to take action. This'is'more than sufficient time for the operator: to t take : corrective action. - Some: transients were not analyzed or discussed..in_this supplement dueLto the combi' nation of the

' low probability of the_ transient occurring and the very short time period of the'special. tests. This is true for the rod ejection acci-

~ dent. TheLeombination of the low probability of, occurring and the bounding doselevaluation'for a condition II transient given here indi-cate that.these events do not need.to be analyzed.. Similar dose calcu-lations have been done for the steamline break accidents which resulta in somewhat _ high ar doses i than the condition II analysis. These dose results indicate that the- f act that the NIS channels are not completely

' qualified does notfalter the conclusion that the results are bounded.

fi

, 4-20

" IO S 2 A'- -

1. _ .- y .. ,__.._m---;,

_ i__  ; __. _ _ _ _ :d.@r ,..7.,_4_ _

3 , - _ -

.;-- ~ ; ,, ,.. -

.-l.,*

/TABLEi1 0

% 1

SUMMARY

J 0F SAFETY' EVALUATION,-SECTIONf 4'.0*:

!Section. Transient < ' Test:- 1- 2' .3: 4 5 :6

1. l' . RCCA Bank With. , Suberit 2,4 . 2,4. 2,4 2,4 2,4 -1 11.2- RCCA Bank' With . ,_ at Power '

4; '4 4 -4 4 1 11.'31 . RCCA Misalignment -l~ 1- l' .1 l' l

. l '. 4 : Boron Dilution 11 1 1- 1 1 1

-1.5 Partial-L'osstof Flow- -l ~1, 1 1 1 li

-1.6f . Start: Inactive Loop .1 l'- l1 1 1 1 11.7- Loss of Load' l- 1 1 1 1 1-1.8 Loss'of Feedwater -1 1 1 1 1 -- 3 1.9' ' Los s' Of fsite Power

-1 1 1 1 1 3 1.10  : Excessive Feedwater 2 ~2 2' 2 2 2

-1.11. Excessive Load? 2 2 2 2- 2 2

.l.12: RCS Depressurization. 1- -4 4' 1 1 1

.l.13' - Steam Depressuri:iation 1 1 -1 1 1 1 1.14- Spurious Safety' Injection- 1 1 1- 1 1 1

-2.1 = Small' .'LOCA 3 3 3 -3 3 1 2.2 :Small Secondary Breaks 2' 2 2 2. 2 1

~2.3 Single: RCCA Withdrawal ~ 4 4 4 4 4' 1

,- '2.4 Mi's loaded Fuel' Assembly- 1 1 1 1 1 1 l Comp'lete Los's'of Flow _.

I l- 'l- 1 1 1

. Waste Gas-Decay. Tank 3rk. ~ IL 1- 1- 1 1 'l 3.1 Major'LOCA. '3 '3' 3' 3 3 1 3_. _2 Major Secondary-3reaki ,

- 2,3 2,3' 2,3 2,3 2,3 1

.3.3 S/C Tube. Rupture. 1- l! 1 1 1- 1 3.4' RCP Locked Rotor. I l- 1 1 1 l' i 3.5- Fuel Handling 1 l' 1 1 1 1 3 .' 6 Ruptured CRDM-3,5 :3,5 3,5 3,5 3,5 1

• Bases of. Evaluation- _

e

11. 1 Bounded by FSARanalysi's-results-

- 2. LReanalysis shows fuel-clad ^ integrity is maintained-

' ~

3. : Operator action'is; required'for protection

!4. Probability of.6ccurrance reduced by restrictions ~on operation ~

' conditions' -

( 3.1 Probability';of loccurrance reduced by_ short: testing _ period solely

)

- f _[

w

- Y 14-211 2 Sui: ,

. V .- . , -

, . - - -. - . u. . ..-

- . - - . - . ~ - ~ - - -

---,-y-

, ~-

m. _g.m. y ,~. .-.

_ .. w- . .

-e REFERENCES

-le J. S. Cellerstedt, R. A. Lee, W. J. Oberjohn, R 9. Wilson, L. J.

~

Stanek, " Correlation of Critical Heat Flux in a hadle Cooled by Pressurized Water," Symposium.on Two-Phase Flow in Rod Bundles, Code H27, ASME Winter Annual Meeting, November, 1969.

2. . Hao,' 3. R. , Zielke, . L. A. , Parker, L 3. , " Low Flow Critical Heat Flux," ANS 22,'1975.

3. Lahey, R. T. , Moody, F. J. , "The Thermal-Hydraulics of Boiling Water Nuclear Reactor," American Nuclear Society, 1977.

4. WCAP-8582-P, Vol. II,'" Blowdown Experiments With Upper Head Injec-tion in ' G2 17x17 Rod Array," McIntyre, 3. A. , August, 1976.

(West-inghouse Proprietary) -

5. Mattson, R. J. , Condie, K. G. , Bengston, S. J. and Obenchain, C. F. ,

"Regres,sica Analysis of Post-CHF Flow Boiling Data," paper 33.8, vol. 4, Proc. of Sch Inc. Heat Transfer Conference, Tokyo, September

-(1974).

si

6. Tong, L~. S. , " Heat- Transfer in Water-cooled Nuclear Reactors," Nuc.

Engng. - and . Des ign - 6_, 301 (1967).

k i

4 4-22 4

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n u

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2.

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5 .c a O'

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u.

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w 0.~ 20

,- , , ,J , , , , i 0.0 10.0' '20.0 30.0 40.0 Time -(sec. ) -

s Ficure:4.2.1

~

Uncontrolled Rod Bank.ilithdrawal frcm a Subcritical Condition, Neutron Flux.

vs. Time-9 L , ,. ,,

,_,.m._,.,._...,,,__

1300.0  :  : ,

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o .. .

e o

e o O e

. o o m .e n w r- C o - cu t

TIME (SEC)

-' c' ',.a' a"?-'2 Uncontrolled Rod Bank '4 ddrawal ' rom a

<ubcritical Condition, not Spot Clad iemceraturevsTime,AssuminaCNBat Time = O.

24-

~ ' - " ^-

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a c., . vi u . , , i , , .. .

it , , ,i .

= m =:

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g j_

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a =

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>

y - - n .- -

- - - = ,

m > 3

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. u o

s e =,

s W _ _-__

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=

=. -  :

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I

= ; en ec r x & w m ~ - c w

~ ._

=

0

(;MW' EZyg ; 30 ",) xn(3- ;?aH ido'] 1eBd

-(md6.00E49'J0 ",) %0L3 STd.

- (3 ): santeaadmai i,vp.put_ aAcqy as tu aanteaacmaj 1stut

. D"*D '" 0 9[ $

_wwA-w A n ffh

4. 25 '.

[ ~ ,.

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D

~D

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100, L

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t .

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' '- '- r ! ' ! ' ' i - '- F ' + ' '

1~

-11 10 -100

. s

','. REACTIVITY INSERTION RATE-(PCM/SEC)

~

Fig'ureiK2.f (!ncontrolled Rod: Bank Withdrawal at Power.

iime or.1 Reactor Trip vrReactivity Insertion .

- -. ..-....-..u-.....-. .

... , . - . . -.  :.a. . . . - - . - - - . . ~ , . ..

- ' m.u ;n?.~::- - - - - -- -. -. -. - - - - - - - - - - - --

, c-. ... . - -

r~.~~

c . ..

.10000 E5 .08000 --

• w =., - .06000 -

4

< .o t

2 w .: u .

J.

- c. .04000 - -

w a _

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0. 0 ~  :

e

> .00050 -- t

~

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< 00 ,

-w m . 0010 -- , , ,

600.00  : .  : .

e Core Avg. -

w

. a o C 500.00 -- +

o U w a

e

=

e .. --

• m w W - 400.,00 - -

All Loops -

o

<eR

< Cold Leg

=

w -- ..

> w

< c.

x

.e o

W 300.00_-- --

u 200.00  :

1 m .15000 -

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g-g 10000 -- --

w -

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s a d

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a o

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o- e o o o o o o

-o o o o o cs c 'o c o- o es e o

e o

o

- o o o o a::s o o o i

o ~' .~ m a c w n m j

TIME (SEC) l l

'!IGURE t. 2.5 TRANSIENTS IN THE REACTOR CCRE .CD CCOLANT LOOPS l FOLLCWING THE OPE',*ING OF A STEAM DUMP VALVE FRCM cy one ; . .t r i rnnne a cT vt- j

_.. _ _ _ . _- . ..~ -

. . ._ ; y ,. .a ., .;, ..n. __.y.

-.a. . . . . . . . .

y 2400'0 2 ,

.v. -. --

1 v.

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=

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"f 2000.0 -N -

.= .. ..

2 e

vs w 1800.0 - - --

c C

1

, 1

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o

>- 1500.0 -- --

a- -

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s 0 1000 00 - - --

c y ... ..

w es

.~ m 500.00 - --

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x 1

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1

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r 1000 00 -- ,g e 800 00 - --

G w

c "o

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C L

Z

< 400 00 -- --

en 4

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0. 0 O C C- O C O O d C o o o --

. o

-o o o e- o e e e e

• O O- O o o o o o - ~ m O,

. o e ~ m TIME (SEC)

FIGURE .'. 2.6 TRANSIENTS IN THE PRESSURIZER AND STE.ui GENERATOR FOLLOWING-THE OPENING OF A STE.O! DU'? VALVE FROM

~5 P0b'ER. ALI. LOOPS ACTIVE L

__ _ .. , ._ sere = ,g _.---_.y,

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m

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e Core Avg D Loop 3&4

\

oC w f 500.00 -- -.

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Loop 3&4 e .O

, 00 .

e .c e- o a 9

g. g. 8 .S, S o

=- , ,

e d__ d

~

d a

< g o TIME (SEC)

FIGURE 4.2.7 TRANSIENTS ni THE REACTCR CORE AND CCOLANT LOOPS FOLLO' JING 'A DCU3LE ENDED RUPTURE. OF A MIN STM-LINE CCtOJSTRIXi 0F THE STER 4LINE ISOLATION k

a u :.

u, . -n.

._ . ;. ~ . . , - .

f; _: :.2

-- ._ a _.:.

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= ..

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T.

=

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$ W 1000.00 --

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l e S --

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i

< 1 e

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00 e o o = . .. =

=

o o e o o o- o o o o .c o.- -o

.o o

o e

o e a

.o -. .~ .m -

TIME .tsEC)

- FIGURE - ~ 4. 2 ; 8 - TRANSIC*TS IT THE PRESSURIZER AND SIE.O! GEiERATOR FOLLC'w'ING - A 00U3LE ENDED.' RUPTURE OF A !!.CN STE_OCINE

'4 29- ..--

.DCWS~? LAM OF THE STE.01INE ISOLATION

(._

~~

j1-

• ' :i5)i 2400 - f--l!<

.- j ij ..

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540 560 580 i 460 480 500 520 1:

Active Cold Leg Temperature, "F W e

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