ML20008D874

Special Low Power Tests Safety Evaluation.

ML20008D874
Person / Time
Site:	McGuire
Issue date:	10/30/1980
From:	DUKE POWER CO.
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ML20008D873	List: ML20008D872 ML20008D874
References
NUDOCS 8010230412
Download: ML20008D874 (47)

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MCGUIRE UNIT pl l SPECIAL LOy POWER TES'IS SAFETY EVALUATION l

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OCTOBER 1980

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1.0 INTRODUCTION

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SUMMARY

In an effort to meet the NRC regulatory requirements of NUREG-0694, "TMI-Related Requirements for New Operating Licenses," special tests similar to those performed at Sequoyah for reactor power levels at or below 5% of Rated Thermal Power are proposed. These tests would demon-strate d2e plant's capability in several simulated degraded modes of operation and would provide opportunities for operator training. The basic mode of operation to be -temonstrated is natural circulation with various portions of the plant equipment not operating, e.g. , pressurizer heaters, loss of offsite power (simulated), and steam generators isolated.

f, Westinghouse has reviewed die proposed tests and has determined diat with close operator surveillance of parameters and suitable operator action points in dse event of significant deviation from test condi-tions, the tests as outlined in the McGuire Special Test procedures are acceptable and can be performed with minimal risk. It is recognized diat in order to perform these tests some automatic safety functions, reactor trips and safety injection, will be defeated. Westinghouse has determined a set of operator action points which should replace these

automatic actuations. It is also recognized diat several technical specification requirements will not be met while either preparing for or performing these tests. Again Westinghouse has determined diat die low

} power levels and operator action will suffice during these time periods.

Westinghouse has reviewed the effect of the pronosed test conditions on the incidents and f aults 42ich were discussed it the Accident Analysis section of die McGuire Final Safety Analysis Report. In most cases, th e FSAR discussion was found to bound the consequences of such events occurring under testing eenditions. Consequences of an ejected RCCA have not been analyzed because of the low probabilities. For some incidents, because of the far-off-normal conditions, die analysis methods available have not shown diat, with reliance on automatic 1

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protection syscem action alone, die FSAR analyses are bounding. In those cases reliance is placed on expeditious operator action. Th e

-/^i operator action points as defined will provide protection for such events.

After performance of Special Low Power Test Programs at North Anna and Sequoyah, Westinghouse has determined d2at use of core exit die rmo-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 die average of the core exit diermocouples to the average of the wide range loop RTD's T.H It was found in both cases that the comparison resulted in agree-ment to within 1 7. A furdaer comparison was made between full core, incore flux map assembly FAH values and the core exit diermocouple readings. D2is comparison resulted in the conclusion that the tempera-ture distribution indicated by diermocouples agreed reasonably well with the power distribution indicated by the flux map. Based on the above, Westinghouse has concluded that core exit thermocouples and wide range RTDs are reliable means of determining margin to saturation temperature,

%i die daermocouples for transient and equilibrium conditions and the RTDs d for equilibrium and slow transient conditions.

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 decreaaing temperature, was less dian 0.5%/1 F. This is less dsan half the e or assumed in the Special Test accideit analyses.

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2.0 DESCRIPTION

OF TESTS 2.1 $ URAL CIRCULATION TEST (TEST 1 - TP/1/A/2150/20)

Objective - To demonstrate the capability to remove decay heat by natural circulation.

Method - The reactor is at approximately 3% power and all Reactor Cool-anc Pumps (RCP's) are operating. All RCP's are tripped simultaneously with die esta'blishment of natural circulation indicated by die core exit thermocouples and the wide range RTD's.

2.2 NATURAL CIRCULATION WITH LOSS OF PRESSURIZER HEATERS (TEST 2 - TP/1/A/2150/20)

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

Mediod - Establish natural circulation as in Test 1 and turn off the pressurizer heaters at the main control board. Monitor the system pres-( sures to determine; the effect on saturation margin and die depressur-t ization rate.

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

Objective - To demonstrate the ability to maintain natural circulation at reduced pressure and saturation margin. The accuracy of die satura-tion meter will also be verified.

Method - The test method is the same as for Test 2, with the exception th at the pressure decrease can be accelerated with the use of auxiliary pressurizer. sprays. The saturation margin will be decreased to approxi-mately 20 7. Demonstrate the effects of charging / letdown flow and steam generator pressure on the saturation margin.

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2.4 NATURAL CIRCULATION WITH SIMULATED LOSS OF 0FFSITE AC POWER (TEST 4 - TP/1/A/2150/21)

/'\~-')i Objective - To demonstrate diat following a loss of offsite AC power, natural circulation can be established and maintained while being now* red from die emergency diesel generators.

Method - The reactor is at approximately 1% power and all RCP's are cerating. All RCP's are tripped and a station blackout is simulated.

AC power is returned by the diesel generators and natural circulation is verified.

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2.5 EFFECT OF STEAM GENERATOR SECONDARY SIDE ISOLATION ON NATURAL CIRCULATION (TEST 5 - TP/1/A/ 2150/21)

Objective - To determine die effects of steam generator secondary side isolation on natural circulation.

s Method - Establish natural circulation conditions as in Test i but at 1%

power. Isolate the feedwater and steam line for one steam generator and establish equilibrium. Repeat this for one more steam generator so diat two are isolated and establish equilibrium. Return the steam generators to service in reverse order.

2.6 SIMULATED LOSS OF ALL ONSITE AND OFFSITE AC POWER (TEST 6 - TP/1/A/2150/26)

Objective - To demonstrate diat following a loss of all onsite and offsite AC power, including the emergency diesal generators, the decay heat can be removed by using the auxiliary feedwater system in die manual mode.

Method - The reactor is shut down and all RCP's are running. Selected equipment will be tripped to simulate a station blackout. Instrument power is- provided by the backup batteries since die diesels are shutdown.

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2.7 ESTABLISHMENT OF PATURAL CIRCUIATION FROM STAGNANT CONDITIONS Westinghouse does not believe that it is advisable to pen'orm this test as noted in a letter from T. M. Anderson, Westinghouse, to H. Denton, NRC, NS-TMA-2242, April 29, 1980.

2.8 FORCED CIRCULATION COOLDOWN This test is performed as preparation for the Boron Mixing and Cooldown Test. Since Westinghouse does not believe it is advisable to perform tite Boron Mixing Test as defined using core heat , it is not necessary to perform the Forced Circulation Cooldown Test.

2.9 BORON MIXING A**D C00LDOWN 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 tinghouse , to H. Denton, NRC.

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'3.0 IMPACT ON PLANT TECHNICAL SPECIFICATIONS 1

In the evaluation of the proposed tests Westinghouse has determined that D) cwelve technical speci.fications will be violated, and thus require exceptions, during the performance of the tests. Table 3-1 lists the technical specifications that will require exceptions and the tests for tAich they will not be met. The following notes the reasons these

.specificati1ns must be excepted and the 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 e.s a function of power, RCS pressure (pressurizer pressure) and loops operable. These limits provide protec-tion by insuring titat the plant is not operated at higher temperatures or lower pressures than those previously analyzed. The core limits in the McGuire tech specs are for four loop operation. Obviously t en in natural circulation with no RCP's running these limits would not be met. However, it should be noted that the tests will be performed with limits on core exit temperature (< 610 F), T,yg (< 590 F) and Loop AT (< 65 F) such diat no boiling will be experienced in the

' core and the limits of specification 2.1.1 for temperature will be met.

The limits will not be met simply because less than 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 faulted conditionc by tripping the plant den various process parameters exceed their analyzed values. Mien in natural circulation two trip functions will be rendered inoperable, Overtemperature AT and Over-power AT. There is a temperature input to these functions Aich ori-3i nates frem the RTD bypass loops. Due to the low flow conditions, 5%

or less, the temperature indications from these loops will be highly

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a suspect. To prevent the inadvertent tripping of the plant Aen in the natural circul'acion mode these functions will be bypassed. Their pro-

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g tection functions will be performed by the operator verifying that Pressurizer Pressure and Level, Steam Generator Level, and subcooling margin (T,,g) 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. When at low power levels it is not unconunon for this function to be difficult to maintain above the trip setpoint. This function i assures that there is some volume of water in rte neam generators above the tops of the U-tubes to maintain a secondary side heat sink. Le amount of water is based on the decay heat present in the core and to prevent dryout of the ' steam generators. With the plant limited to 5%

RTP or less and being at BOL on Cycle 1 there will be little or no decay ,

nuat present. We heat source will be the core operating at the limited 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 trip setpoint from 12% span to 5% span for all o' the special tests. In addition, the steam generator low-level setpoint thich is part of the steam /feedwater mismatch alarm may be lowered tc 5% span.

3.1.3 T.S. 3.1.l.3 MODERATOR TEMPERATURE COEFFICIENT j The Moder t.cor Temperature Coefficient is limited to O pcm/ F or more

negative. When performing tests with the plant critical below 551 F this coefficient may be slightly positive. However, it is expected that the Isothermal Temperature Coefficient will remain negative or approxi-

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mately zero. The tests will be performed such that this is the case and thus minimizing any impact from rapid heacups or cooldewns. In addi-tion, the effect of a small positive Moderator Temperature Coefficient has been considered in the accident analyses performed for the test conditions.

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i 3.1.4 T.S. 3.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY fN The Minimum Temperature for Criticality is limited to 5510F by spec.

(} 3.1.1.5 and 5410F by spec. 3.10.3. To perform test 5 it is expected th at the RCS average ' temperature will drop below 541oF. Westinghouse has determined that operation with Tavg as low as 435 F is accept-able assuming that:

1. Control Bank D is inserted to no deeper than 114 steps withdrawn, and

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 requirements, 4) limiting maximum power to a very low value on an unplanned power excursion, and 5) allowing the use of the "at power"

() reactor trips as back-up trips rather than as primary trips.

3.1.5 T.S. 3.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION 1

1 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 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 24 special tests, all automatic Safety Injection functions will be blocked. Indication of partial Safety Injection logic trips for the non-defeated channels and manual initiation will be operable, however, the automatic Safety Injection actuation functions will be made v

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inoperable by forcing the logic to see that the reactor trip breakers

, are open. Westinghouse believes that this mode of operation is accep-table for the short period of time these tests will be carried out based on the following:

1. Close observation of the partial trip indication by the operator,

2. Rigid adh*rence to the operator cetion points as defined by West-inghouse, see Section 3.2.

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3. Little or no decay heat is present in the system, thus Safety Injec-tion serves primarily as a pressurization function.

Blocking these functions will allow the performance of these tests at low power, pressure, or temperature and close operator surveillaners will ,

assure initiation of Safety Injection, if required, within a short time period.

i Lowering the automatic auxiliary feedwater start will have little p

effect, since there is little or no decay heat present. Cicse operator surveillance will insure 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 plant. Normally this is accomplished through the use of pressurizer heaters and spray. In several tests the. pressurizer heaters will be either turned off or rendered inoperable by loss of power. This mode of operation is acceptable in that pressure control will be maintained through the use of pressurizer level and charging / letdown flow.

3.1.8 T.S. 3.7.1.2 AUZILIARY FEEDWATER SYSTEM The auxiliary feedwater system will be rendered partially inoperable for two tests. The two tests simulate some form of loss of AC-power, i.e.,

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notor drivan cuxilicry fesdwater pumps inoparable. Wastinghouse has determined that this is acceptable for these two tests because of the

( little or no decay heat present allowing sufficient time (# 30 min-utes) for operating personnel to tack in the pump power supplies and regain steam generstor level.

3.1.9 T.S. 3.8.1.1, 3.8.2.1, 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 sa hty related equipment will function d en required.

3.1.10 T.S. 3.10.3 SPECIAL TEST EXCEPTIONS - PHYSICS TESTS This specification allows the minimum temperature for criticality to be as low as $41 F. Since it is expected that RCS T,y will be taken as 1.v as 485 F this specification will be excepted. See Section 3.1.4 for basis of acceptability.

3.1.11 TECHNICAL SPECIFICATIONS NOT EXCEPTED Wile not applicable at power levels below 5% RTP the following tech-nical specification limits can be expected to be exceeded:

1. 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% RTP with RCPs running. However at such a low power ,

level no significant deviations in burnup or Xe peaks are expected.

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2. 3.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR - F AH I

At low temperatures and flow FAH can be expected to be higher than if pumps are running. However, no significant consequences for full power operation are expected, m

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3. 3.2.4 QUADRANT POWER TILT RATIO With no, one, two, or three pumps running and critical, core power distributions resulting in quadrant power tilt may form. At low power levels and for short periods of times these tilts will not significantly influence core burn-up.

3.2.5 DNB PARAMETERS In the performance of several rests the plant will be depressurized f below ?,230 psia. At low operating power levels this depressur-ization 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 3.1.1.4 Minimum Temperature for Criticality O

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 Coolant Loops - Normal Operation.

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3.2 OPERATIONAL SAFETY CRITERIA During the performance of these tests the operator must meet the follow-O, . . .

Ing set of criteria for operation:

1. Maintain For All Tests a) Primary System Sub-cooling (T3 ,g Margin) > 20 F b) Steam Generator Water Level > 25% Narrow Range Span c) Pressurizer Water Level (1) With RCPs running > 22% Span (2) Natural Circulation > Value d en RC?s tripped d) Loop AT $ 65 F e) T,yg 3 590 F f) Core Exit Temperature (highest) $ 610 F g) Power Range Neutron Flux Low Setpoint and Inter:nediate Range Neutron Flux Resctor Trip Setpoints 3 7% RTP

4) Control Bank D 114 steps withdrawn or higher i) RCS cold temperature > 485 F

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

a) Primary System Sub-cooling (T,,e Margin) 3 15 F b) Steam Generator Water Level < 5% Narrow Range Span or Equivalent Wide Range Level c) NIS Power Range, 2 channels > 10% RTP d) Pressurizer Water Level < 17% Span or an unexplained i decrease of more than 5% not 1

concurrent with a T,y ch ange '

e) Any Loop AT > 65 F f) T > 590 F

! ayg g) Core Exit Temperature (highest) > 610 F h) Uncontrolled rod motion (m) wJ i) Control 3ank D less than 114 steps withdrawn j) RCS cold temperature 485 F l 3-7 L 7052A l

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3. Safety Injcetion must b3 m:nually initiatsd if (ny of the following condi-tions are met:

a) Primary System Sub-cooling (T,,e Margin) < 10 F o) Steam Generator Water Level < 0% Narrow Range Span or Equivalent Wide Range Level c) Containment Pressure > 1.1 psig d) Pressurizer Vater Level < 10% Span or an unexplained decrease of more than 10% not concurrent with a T,yg change.

e) Pressurizer Pressure Decreases by 200 psi or more in an unplanned or traexplained manner.

Safety Injection must not be terminated until the Westinghouse criteria as defined in E0I:E-2, Loss of Secondary Coolant are met.

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

O 1. Sufficient margin to 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 i side heat sink,

3. sufficient level in t..e pressurizer to assure coverage of the heaters to maintain pressure control,

4. sufficient control rod worth to ensure adequate shutdown margin and minimize impact of uncontrolled bank withdrawal, and

5. limit maximum possible power level in the event of an uncontrolled power increase.

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TABLE 3-1 TECHNICAL SPECIFICATION IMPACT b

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Test Technical Specification 1 2 3 4 5 6 i

2.1.1 Core Safety Limits .X X X X X 2.2.1 Various Reactor Trips Overtemperature AT X X X X X X Overpower AT X X X X X X Steam Generator Level X X X X X X i 3.1.1.3 Moderator Temperature coef- X

, ficient 3.1.1.4 Minimum Temperature for X Criticality 3.3.1 Various Reactor Trips Overtemperature AT X X X X X X Overpower AT X X X X X X Steam Generator Level X X X X X X 3.3.2 X X X X X X

( Safety Injection - All automatic functions Auxiliary feedwater automatic start X X X X X X 3.4.4 Pressurizer 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 Onsite Power Distribu- X X tion System 3.8.2.3 DC Distribution System JC X 3.10.3 Special Test Exceptions - X Physics Tests l

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4.0 SAFETY EVALUATION

( In this section the safety tffects of those special test conditions which are outside the bound: of conditions assumed in the FSAR are evaluated. The interaction of these conditions with the transient analyses in the FSAR are discussed.

4.1 EVALUATION OF TRANSIENTS The ef fect of the unusual operating conditions on die transients analyzed in the FSAR are evaluated.

4.1.1 CONDITICN II - FAULTS OF MODERATE FREQUENCY 4.1.1.1 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from ,

a Suberitical Condition Restriction of control rod operation to manual control, and constant operator monitoring of rod position, nuclear power and temperatures

greatly reduces the likelihood of an uncontrolled RCCA withdrawal.

Operation without reactor coolant pumps, and in some cases with a posi-tive moderator temperature reactivity coefficient, tend to make die consequences of RCCA withdrawal worse compared to the operating condi-tions assumed in the FSAR. For these reasons the operating' procedures require diat following any reactor trip at least one reactor coolant pump will be restarted and die reactor boron concentration will be such diat it will not go critical with less dsan 114 steps withdrawal on D I

Bank. An analysis of diis event is presented in Section 4.2.1. For Test 6, d2is transient is bounded by the FSAR analysis, since all reac-tor: coolant pumps are operating.

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4.1.1.2 Uncontrolled Rod Control Cluster Assembly Bank Withdrawal at Power The same considerations discussed in Paragraph 4.1.1.1 apuly here. In t

,,s addition, the low operating power and the Power Range Neutron Flux Low

() and Intermediate Range Neutron Flux trip setpoints act to mitigate this 4-1

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incident, while lack of the Overtemperature AT trip removes some of s the protection provided in the FSAd case. An analysis is discussed in Paragraph 4.2.2.

4.1.1.3 Rod Control Cluster Assembly Misalignment The FSAR discuseion concerning scacic RCCA sisalignment applies to the test conditions. The consequences of a dropped RCCA would be a decrease in power. Thus no increase in probability or severity of this incident is introduced by the test conditions.

4.1.1.4 Uncontrolled Boron Dilution The couraquences of, and operator action time requirements for, an uncontrolled boron dflucion undrer the test conditions are bounded by i those discussed in the FSAR. The fact that the control rods will never be inserted to the insertion limits, as well as the Power Range Neutron Flux Low Setpoint and the constant operator monitoring of reactor power, temperature and charging systen operation, provides added protection.

4.1.1.5 Partial Loss of Forced Reactor Coolant Flow Secause of :he lo.i power limits the consequences of loss of reactor coolant pump power are trivial; indeed they are bounded by normal opera-ting conditions f or these tests.

4.1.1.6 Startup of an Inactive Reactor Coolant Loop When at least one reac:or coolant pump is operating, the power limit for chese tests results in such maall temperature differences in the reactor coolant system that startup of another loop cannot introduce a signifi-cant reactivity disturbance. In natural circulation operation, inadver-tent startup of a pump would reduce the core water temperature and thus provide a change in reactivity and power. 3ecause of the small nodera-tor reactivity coefficient at beginning of life the power increase in

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ratio in the core would be increasing. The Power Range Neutron Flux Low Setpoint reactor trip provides an upper bound on power. Because of the I increase in flow-to power ratio and because of the low setpcint 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 low power level, the disturbance caused by au Loss of

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load is small. The FSAR case is bounding.

4.1.1.8 Loss of Normal Feedwater Because of the low power level, the consequences of 'a 16 : n 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 Level. Ample time is available to rein-stitute auxiliary feedwater sources.

4.1.1.9 Loss of Offsite Power to the Station's Auxiliaries (Station Blackouel Because of the low power level, the consequences of a loss of off-site power are bounded by the FSAR case.

4.1.1.10 Excessive Heat Removal Due to Feedwater System Malfunctions The main feedwater control valves will not be used while the reactor is at power or near criticality on these tests. Thus, the potential water flow is restricted to the auxiliary feedwater flow, about 6% of normal flow. The transient is further mitigated by the low operating power level, small moderator temperature reactivity coefficient, the low set-points on the Intermediate and Power Range Neutron Flun Low setpoint

trips, and close operator surveillance of feed (1,v, RCS temperatures.

RCS pressure, and nuclear power. The case of excess heat removal due to feedwater system malfunctions with very low reactor coolant flow is among the cooldown transients discussed in more detail in Section 4.2.3.

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4.1.1.11 Excessive Laad Increase Incident

/ The turbine will not be in use during the perforscnce of these tests, and loau control will be limited to operation of a single stemn dump or steam relief valve. The small moderator temperature reactivity coeffi-cient also reduces the consequences of dtis transient. Close operator surveillance of steam pressure, cold leg temperature, pressurizer pres-sure, and reactor power, with specific initiation criteria for manual reactor trip, protect against an unwanted 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

Close operator surveillance of pressurizer pressure and of hot leg sub-

, cooling, with specific initiation points for manual reactor trip, pro-vides protection against DNB in the event of an accidental depressur-

) ization of the RCS. In addition, automatic reactor trip caused by the Low Pressurizer Pressure Safety Injection signal would occur when core outlet subcooling reached approximately 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, the pressurizer PORV block valves will be closed to remove the major credible source of rapid

< inaevertent depressurization. (The Low Pressure trip is automatically reinstated when pressure goes above 1955 psig and die PORV block valves will be reopened at that time.)

4 4.1.1.13 Accidental Depressurization of the Main Steam System i

The FSAR analysis for accidental steam system depressurization indicates th at if 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 temperature reactivity coefficient which will exist during die

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test period, the reactor would remain suberitical even if it were cooled to room temperature without Saf ety Inj ection. Thus the SAR analysis is bounding.

4.1.1.14 Spurious Operation of the Safety Injection System at Power In order to reduce the possibility of unnecessary thermal fatigue cycling of the reactor coolant system components, the actuation of- high head charging in the safety injection mode, and of the safety injection pumps, by any source except manual action will be disabled. Thus, the most 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 Ruptured Pipes or from Cracks in Large Pipes Which Actuates ?;nergency Core Cooling O A review of the plant loss of coolant accident behavior during the low power testing sequence indicates that without automatic Safety Injection there is sufficient cooling water readily available to prevent the fuel rod cladding fra overheating on a short term basis. The system inven-tory and normal charging flow provide the short term cooling for the small break transient. A sample calculation for a 2 inch break shows that the core remains covered for at least 6000 seconds. This is suf-ficient time for the operator to manually initiate SI and align the system for long term cooling.

It must be noted that the magnitude 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.

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1 A.I.2.2 Minor Secondard S stem Pipe 3reaks The consequences of minor 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 l The FSAR analysis shows that assuming limiting parameters for normal operation a maximum of 5 percent of the fuel rods could experience a DN3R cf less than 1.3 following a single RCCA withdrawal. As the FSAR poi nts out, no single electrical or mechanical failure in the control i system could 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 t.2e restriction to manual control, and by the .

close operator surveillance of res' :or power, rod operation, and hot les temperature.

4.1.2.4 Other Infrequent Faults i

The consequences of an inadvertent loading of a fuel assembly into an improper position, complace loss of forced reactor coolant flow, and waste gas decay tank rupture, as described in the FSAR, have been reviewed and found to bound the consequences of such events occurring during test operation. l i

l 4.1.3 CONDITION IV - LIMITING FAULTS 4.1.3.1 Major Reactor Coolant Pipe Ruptures (Loss of Coolant Accident)

A review of the plant loss of coolant accident behavior during the low power testing sequence indicates that without automatic safety injection there is sufficient cooling water readily available to prevent the fuel rod cladding from over heating on a short term basis. Daring the large break event the system inventory and cold leg accumulators will have removed enough energy to have filled the reactor vessel to the bottom of the nozzles. Following the system depressurization :here is enough 4-6 7M7A

4 wnter in the racetor vassel'below the nozzles to ksep the core covarad for over one hour using conserva, a assumptions. This is sufficient time for the operator to manually initiate SI and align the system for A long tem cooling. At no time during this transient will the core be uncovered.

It must be noted that the magnitude of the resulting clad heatup 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.

4.1.3.2 Major Secondary System Pipe Rupture The small moderator temperature reactivity coefficient, close operator surveillance of pressurizer pressure, cold leg temperature, and reactor j oower, with specific initiation criteria for, reactor trip; low trip l setpoints on the Intermediate-Range and Power-Range Neutron Flux trips; MSIV closure on Low Steam Pressure; and Low Pressurizer Pressure trip l (S.I. initiation) assure a Reactor Trip without excessive reactor power following a cooldown transient caused by the secondary system.

% Following reactor trip, assuming the worst RCCA stuck out of the core, the reactor would remain. suberitical even if it were cooled to rcom j temperature. Transient analyses for a steam pipe cureure are provided

, in Section 4.2.3. The consequences of a main feedline rupture are bounded in the cooldown direction by the steam pipe rupture discussion.

Because of the low operating power, the heatup aspects of a feedline rupture are bounded by the FSAR discussion.

4.1.3.3 Steam Generator Tube Rupture The steam generator tube rupture event may be categorized by two dis-tinct ph as es . The initial phase of the event is analogous to a small LOCA event. Prior to operator-controlled system depressurization, the steam generator tube rupture is a special class of small break LOCA I

w 4-7 7052A

trcusisnes, cnd the operator actions required to daal with diis situ-ation during this phase are identical to those required for mitigation of a small LOCA. Hence, evaluation of the steam generator tube rupture t

\~ during this phase is wholly covered by the safety evaluation of ths small LOCA.

Af ter the appropriate operator actions have taken place to deal with die initial LOCA phase of the event, the remainder of the steam generator tube rupture accident mitigation would consist of those operator actions required to isolate the faulted steam generator, cooldown die RCS, and depressurize the RCS to equilibrace primary RCS pressure with the faulted steam generator secondary pressure. These actions require util-ization of the following systems:

1. Auxiliary feedwater control to the faulted steam generator.

2. Steam line isolation of the f aulted steam generator.

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

4. RCS depressurizatior capability.

Evaluation of the McGuire special test procedures has verified that all of the above systems are immediately available for operator control from the control room. Therefore, it is concluded that the ability to miti-gate the steam generator tube rupture event is not compromised by the modifications required for operation at 5% power during the proposed tests, and that the analyses performed for die SAR regarding this event remai:.1 bounding.

1 4.1.3., Stagle Reactor Coolant Pump Locked Rotor Because of the low power level, the locking of a single reactor coolant pump rotor is inconsequential.

nv 4-8 7052A

4.1.3.5 Fusi Htndling Accidsnes The FSAR analysis of fuel handling accidents is bounding.

\_ 1 4.1.3.6 Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)

The 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 ejec- j ted rod will be substantially less dsan the delayed neutron fraction. l Th us , the power rise following a control rod ejection would be rela-  !

tively gradual and terminated by die Power Range and Intermediate Range l Neutron 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 d2e FSAR, the result of combining these ameliorating effects with the effect of the natural circulation flow rate on clad-to-water heat 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.

CD .2 ANitxS1S e, 1RANS1sN1S 4.2.1 ANALYSIS OF RCCA 3ANK WITHDRAWAL FROM SUBCRITICAL CONDITION 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:

I

1. Reactor coolant flow was 0.1% of nominal.

2. Control rod incremental worth and total worth were upper bound values for the D bank initially 114 steps withdrawn.

3. Moderator temperature reactivity coefficient was an upper bound (positive) for any core average temperature at or above 4850F.

\_

4-9 7052A

l l

4. The lower bound for total delayed neutron fraction for the beginning of life for Cycle 1 was used.

5. Reactor trip was initiated at 10% of full power.

6. DNB was assumed to occur spontaneously at the hot spot, at the beginning 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 130007, as is shown in Figure 4.2.2. No clad failure is expected as a result of this transient.

4.2.2 ANALYSIS OF RCCA 3ANK WITHDREAL AT POW 1!1 Analyses of RCCA bank withdrawal transients were performed for natural.

circulation conditions. The transients were assumed to start from steady-state operating conditions at either 1% or 5% of full power, and 1 with either all stemaline isolation valves open or two of those valvits f closed. A range of reactivity insertion rates up to the mumum for two banks moving was assumed for cases with all stesalines open, and up to the maximum for one bank moving for the cases with two steamlines iso-laced. Both =14 mum and minimum bounds on reactivity feedback coeffi-cients for beginning of life, Cycle 1, were investigated. In all cases, reactor trip was initiated at 10% nuclear power.

Reactor condici se the time of maximum core heat flux are shown in Figures 4.2.3 and . .4 as functions of the reactivity insertion race for three four-loop active cases. For high reactivity insertion races, the minimum reactivity coefficient cases give the greate. - heat flux after the trip setpoint is reached, and have the lowest coolant flow rate at the time of peak heat flux. For these cases even the slowest insertion rates studied did not result in any increase in core inlet temperature at the time of peak heat flux. For maximum feedback cases, however, the transients for very low insertion rates go on f or so long C

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4-10 7052A

that the core inlet temperature finally increases before trip, i.e.,

A af ter approximately one and one-half minutes of continuous vf thdrawal.

Thus, the cases shown bound the worst cases.

1 l

l 4.2.3 ANALYSIS OF C00LDOWN TRANSIZNTS l

Cooldown transients include feedwater system malfunctions, excessive i steam load increase, accidental depressurization of the main steam sys-tem, and minor and major secondary system pipe ruptures. Attention has been focused on the possibility and magnitude of core power transients resulting from such cooldev.as before reactor trip would occur. (Follow-ing reactor trip, no coo' down event would return the reactor to a cri-tical condition.)

During natural circulation operation, approximately one ce two minutes would elapse following a secondary side event before cold water from the stema generator reached the core; thus, considering the close and con-stant surveillance during these tests, time would be available for the operator to respond to such an event. Analyses were also performed to determine the extent of protection provided by automatic protection systems under trip conditions.

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 cause an increase of less than four percent in reac-tor power, with a corresponding increase in core flow with natural cir-culation, assuming the bounding negative moderator temperature coeffi-cient for the beginning of life, Cycle 1. Thus no automatic protection is required, and ample time is available to the operator to trip the reactor, isolate feedwater to the faulted steam generator, and isolate the break to the extent possible. Calculated results for the sudden opening of a single steam valve, assuming the most negative 30L Cycle one moderator reactivity coefficient and 57. initial power are shown in Figures 4.2.5 and 4.2.6.

4-11 7052A

4.2.3.2 High Flux Protection i

Reactor trip on high nuclear flux provides backup protection for larger  ;

pipe breaks or load increases. Analyses were performed to determine the worst core conditions that could prevail at the time of high-flux trip, independent of the cause. The following assumptions were used:

1

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

l

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

3. Initial operation with core inlet temperature 5550F.

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

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

This results in some boiling in the reactor. The negative reactiv-icy introduced by core boiling would effectively limit power; this l 1

negative reactivity was conservatively neglected. I i

l l

6. Uniform core inlet temperature and flow. i l

l l

7. Reactor trip equivalent to 10% of full power at the initial inlet  !

1 temperature. The power as measured by the NIS is assumed to be diminished from the true power by 1% for each 107 decrease in reactor inlet temperature, resulting in a true power of greater than  ;

iG% ac che time of trip.

8. Core flow race as a function of core power was assumed equal to the predicted flow under steady-state operating conditions.

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+

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i A-12 7052A

L l Anslysss of core conditions bassd on thass assumptions indi sta that the DNB criterion of the 1SAR 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 in any two lines. Low Pressurizer Pressure and Power Range Neutron Flux

low setpoint trips serve as further backups. An example is the 1

double-ended rupture of a main steamline downstream of the isolation valves, with all' isolation valves initially open. Figures 4.2.7 and 4.2.8 show the response to such an event, with an initial power of 5%

and natural circulation. The Low Steamline Pressure trip occurs almost immediately. Ir the example shown, the main steamline isolation valve on loop one was assumed to fail to close. No power excursion resulted, and the reactor remained suberitical after the trip.

4.3 ADDITIONAL CONSIDERATIONS In the great majority of cases it was concluded, either by reanalysis or -

by comparison with previously analyzed FSAR conditions, that fuel clad integrity would be maintained without need for operator mitigating action. For the LOCA or steambreak events, it was concluded that the 4

operator 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 manual action, e.g., manually initiate saf ety injection, to preclude fuel damage.

Finally, in certain other cases, primarily associated with certain inadvertent RCCA withdrawal events, the postulated accident conditions were neither amenable to direct analysis nor credit for operator inter-vention. In particular, the postulated accident conditions were outside the bounds of accepted analysis techniques so that fuel damage was not precluded either by analysis or identified operator action. For these cases, the basis for acceptability was primarily associated with the low probability of an inadvertent rod withdrawal event during the limited t duration of the special tests.

N 4 '

7052A

This ssction provid2s cn additionni esssssm:nt relctiva to the potential for and consequences of fuel f ailure f or these "unanalyzed" accident f~~3: onditions associated with certain rod withdrawal events. This assess-e

\_ / 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:

1. Thermal margin associated with normal test conditions.

2. The potential f or DNB during accident conditions.

3. The clad temperature response assuming that DNB occurs.

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.

p)

(v

2. Even 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 co servatisms, the resulting off site dose would be small.

4.3.1 DE3IGN CONSIDERATIONS Margin to hot channel boiling has been incorporated with all normal test conditions by establishing a lower bound requirement on the degree of reactor coolant subcooling. This test requirement assures that postula-ted accidents are initiated from a condition of excess thermal margin.

O i-14 7052A

l 4.3.2 DNB CONSIDERATIONS O For certain cooldown transients, the conclusion that DN3 is precluded was drawn based on use of the W-3 critical heat flux correlation.

Although the analyses for the cooldown events discussed in section t

4.2.3.2 result in mass velocity below the range of direct applicability of the correlation, the reactor heat flux was so low relative to the predicted critical heat flux that even a factor of 2 would not result in serious concern for DNB for this event.

For the non-cooldown transients 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.

Comparison of the W-3 DNB correlation to low flow DNB test data and correlations (references 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 (refer- l l

ence 3) at these pressures are higher than those predicted by the low ]

flow correlations. Further review of the data in ref erence 1 indicates that the critic ~al heat flux at higher pressure is significantly lower than the abova data at 1000 psi. The minimum critical heat flux of the l data set is .16 x 100BTU /hr-ft2for a data point at 2200 psia at a  ;

mass velocity of .2 x 100 lbm/hr-ft2, Since the exit quality for this data point was 64%, it is unlikely that the reactor would be able to maintain a heat flux of that level due to the nuclear feedback from voiding. The power distribution would tend to peak towards the bottom thus further reducing the local quality at the peak flux locstions.

O v

4-15 7052A

Also ths pool boiling correlations in

• erence 3 show soma dscrease in critical heat flux above 1000 psir to the maximum pressure of applica-bility of 2000 psia. However cstrapolation of the correlations to a

{

• value of zero critical heat f.'ux at the critical pressure (3206.2 psia) would not result in lower crit.. cal heat fluxes than shown in the data set from reference 1. Since tht core average heat flux at IC% of nom-iaal power (highest expected powtr for heatup events) is only on the order of'.02 x 10 BTU 6 /hr-ft a2 Isrge peaking factor would be required to put the reactor heat flux as high as the critical heat flux.

)

For the transients considered, the only ones that lead to significant  :

off normal peaking factors are rod motion transients. The rod with-drawal from suberitical is a power burst concern. As such, it is expec-ted that even if DNB occurred, the rod surface would rewet. For the rod bank withdrawal, the combination of maximum power and peaking factor would result in a peak power lower than the data referenced above. .

Given tre lack of data, it is difficult to completely preclude DNB, although a prudent judgement indicates that it is indeed remote.

I 4.3.3

'}

/

CLAD TEMPERATURE CONSIDERATIONS Should DNB occur, the peak clad temperature reached would depend pri-marily on the local nuclear transient following DNB and on the behavior of the post-DNB heat transf er coef ficient.

For a rapid power transient, as is illustrated by the SER analysis for RCCA bank withdrawal from a suberitical 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-perature calculated was under 1300 F even assuming an extremely low 2

heat transfer coefficient (# 2 BTU /hr-ft - F).

A possibly more limiting condition for RCCA withdrawal would be the case in which a power increase causes DNB but would either not result in reactor trip on high nuclear flux or the trip is delayed. In the former

('~)

s_ -

4-16 7052A

case, a stssdy state condition with hSt spot DNB could ba 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 Flux low setpoint. The trip setpoire is at 7% for these tests. To allow for calorimetric errors and #2rmal system errors, trip is assumed to occur at 10% of rated thernal power (RTP), unless a large decrease-in downcomer coolant temperature occurs during the test. In tests 2 and 3, depressurization 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 power (fraction of RTP). For the event of an uncon-trolled RCCA bank or single RCCA the upper bound of this heat flux product is approximately 0.34. Using this value, the heat transfer coefficient required to keep the peak clad temperature below 1800 F, the threshold of significant heat flux increases due to zirconium-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 extrspolated to obtain representative heat transfer coefficients. The Westinghouse UHI film boiling correlation (reference 4), was developed at low flow condi-tions similar to those postulated for incidents occurring during the 5 McGuire tests. This correlation was extrapolated to the higher pressure conditions of the tests to obtain representative film boiling coeffi-

, cients. This resulted in a heat transfer coefficient in excess of (100 STU/hr-ft2 _oF )"'c at 2200 psia and 5% flow with quality between 10-50%, Other film boiling heat transfer correlations, devel-f-

oped at higher pressures, were also examined. These correlations were

?

4-17 7 052A

extrapolcced down to tha low r flow conditions of ths McGuire tests as another approach to obtain representative film boiling coefficients.

Using both the Mattson'et c1 (reference 5) and the Tong.(reference 6) film boiling correlaticus resulted in post-DNB heat transfer coeffi-cients in excess of 150 BTU /hr-f 2t - F at the condition: given above.

These results indicate that a' 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 were performed for a hypothetical accideat senario using conservative assumptions so as to determine an extreme upper bounc 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 assumptions made in the analysis include:

170 Mwt (5% power) ss, 1.0 dose-equivalent I-131 RCS activity (tech spec limit) 500 gpd steam generator leak in each SG (tech spec limit) 100% clad damage and gap activity release 10% iodine / noble gas in gap space 100 DF in steam generators j 500 iodine spike factor 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/m X/Q percentile value

• 3 The results of the analysis show that the two hour site boundary doses would be 5 rem thyroid, 0.9 rem total body and 0.4 rem to the skin.

The analysis of the accidents has incorporated some very conservative assumptions which goes beyond the normal degree of conservatism used in FSAR analyses. The most prominent of these assumptions and a brief description of the extreme conservatism includes:

O,-

s l

4-18 7052A I

1) Equilibrium rsdionuclida invantoriss establishsd at 5% powar. For iodines, this requires # 1 month.7f steady state operation at 5%

uninterrupted.

1 (

w

2) Fuel clad gap inventories at 10% of core inventory, this is a time dependent, temperature dependenc phenomona. At 5% power, very e

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

3) 100% fuel rod clad damage.

4) Primary to secondary leakage to tech spec values. Since McGuire is a new plant, no primary to secondary leakage is expected. If leakage were present, it would most likely slowly increase in steps up to tech spec levels.

5) Percentile meteorology, there is 95% probability of atter diffusion characteristics and thus lower offsite doses. Additionally, the fifth percentile I/Q for McGuire is significantly less than the generic value used in this analysis.

(

\

For these reasons, in the unlikely event of a potential accident during the tests, the resulting dose is suall, even assuming 100% clad damage and other extreme conservatisms.

• This is a generic conservative value representing the worst meteor-ological dispersion characteristics of any Westinghouse nuclear plant site in the United States.

i 4-19 7052A

4.3.5 OTHER CONCERNS The LOCA analyses presented indicate that there are over 6,000 seconds s_ ,/ for the operator to take action. This is more than sufficient time for the operator to take corrective action. Some transients were not analyzed or discussed in this supplement due to the combination 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. The combination of the low probability of occurring and the bounding dose evaluation f or 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 results in somewhat higher doses than the condition II analysis. These dose results indicate that the fact that the NIS channels are not completely qualified does not alter the conclusion that the results are bounded.

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O 4-20 7052A

TABLE 1

SUMMARY

OF SAFETY EVALUATION. SECTION 4.0*

v Section Transient Test: 1 2 3 4 5 6 4.1 1.1 RCCA Bank With. , Suberit 2,4 2,4 2,4 2,4 2,4 1 1.2 RCCA Bank With . , at Power 4 4 4 4 4 1 1.3 RCCA Misalignment 1 1 1 1 1 1 1.4 Boron Dilution 1 1 1 1 1 1 1.5 Partial Loss of Flow 1 1 1 1 1 1 1.6 Start Inactive Loor. I 1 1 1 1 1 1.7 Loss of Load 1 1 1 1 1 1 1.8 Loss of Feedwater i 1 1 1 1 3 1.9 Loss Offsite 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 1.12 RCS Depressurisation 1 4 4 1 1 1 1.13 Steam Depressurization 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 Misloaded Fuel Assembly 1 1 1 1 1 1

() Complete Loss of Flov Waste Gas Decay Tank Brk.

1 I

1 1

1 1

1 1

1 1

l 1

3.1 Major LOCA 3 3 3 3 3 1 3.2 Major Seconuary Break 2,3 2,3 2,3 2,3 2,3 1 3.3 S/G Tube Rupture 1 1 1 1 1 1 3.4 RCP Locked Rotor 1 1 1 1 1 1 3.5 Fuel Handling i 1 1 1 1 1 3,5 3.6 Ruptured CRDM 3,5 3,5 3,5 3,5 1

• Bases of Evaluation

1. Bounded by FSAR analysis results

2. Reanalysis shows fuel clad integrity is maintained

3. Operator action is required for protection

4. Probability of occurrance reduced by restrictions on operation conditions

5. Pro.ubility of occurrance reduced by short testing period solely p

k 4-21 7052A

REFERENCES s

1. J. S. Gellerstedt, R. A. Lee, W. J. Oberjohn, R. H. Wilson, L. J.

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

2. Hao, 3. E., Zielke, L. A., Parker, M. B., " 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 Socisty, 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. ,

" Regression Analysis of Post-CHF Flow Boiling Data," paper 33.8, Vol. 4, Proc. of Sch Inc. Heat Transfer Conference, Tokyo, Septembe:

(1974).

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

Engng. and Design 6, 301 (1967).

i 4-22 7052A

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< 0. aw w -

i

= .00010 - -

i 8

1 1

600.00  : .

1 C

-- Core Avg w -

e O C 500.00 --

C e v w C C, z

• w w

u M 400.00 -- *All Loops --

< w

< a.

1 MW C

300.00 - - --

v 200.00  : . .

.15000 2 ,

C -

a a " --

v 5

All Loops g 10000 - - --

w z --

x o C e v .05000 - - --

g C -*

g -- -.

o, 0.0 C C C C C C C C C C C C C C C C C C C C C C C C C

- O C C C, a:3 C C C o - cu m o w ~ C

\

4 i V TIME (SEC)

FIGURE 4.2.5 T2ANSII'ITS I'I THE FIAC CR CCPI AND CCOLA'r: LOOPS FOLLCWING THE OPE'i!'iG OF A STIAM DDT VAL 7E FRCM 5: POWF.R, ALL LOOPS AC !VE

e a

2400 0 m ..

W 2200.0 - -

a. ..

= < ..

40 2000.0 - r -

= .. ..

3 e

• v$

w 1800.0 - - --

2 C

2 . . , ,

o 1500 0 - -

= ..

m - ..

5

• 1000 00 - -

, u ... ..

w =

2m 500.00 -- -

= ht n .. ..

W O.0 .

2

a. '

, 1200 0 l f

1000 00 - -

311 tcap, s

v'.

a 800 00 - - --

w E

3, 600 CC - - --

v w

= A ,

I Q.

1 5

400 00 -- --

l vi 200 00 - -

1 1

0. 0 e e o o o e e e C o o o - - -

. o O

=

b C

C C C b C b C

b C

c C

b C

o - <u m C, o e ~ m l

l l TIME (SEC) s_

FIGURE 4.2.6 T?aNSIENTS IN '"HE PRESSL'RIZER .Al!D STE.Di GENE?aTOR FOLLOWING THE OPENING OF A STE.ui Di;MP VAL 7E 72CM 5: POWER, ALL LOOPS AC*IVE

.100C0

, .. Y 5 .08000 --

_ _5, ..

o .06000 - -

< z

] Yo .04000 1 ..

w m .

Es U

.02000 - -

0.0 1

> .01000 - ..

.02000 -

u ..

' ~

5. .03000 -- ..

.04000 600.00

" Core Avg

" Looo 3&4 o, ,C o o 500.00 --

\ s v w a o .. X.. /

x L.oop 2

< e wW 400.00 - -

Loop 1 M [a x < ..

w =

> m

< o.

1 -

Mw o

300.00 - -

V 200.00 6

.15000 o

a ..e ..

w < --

x v - "

gg .:0'00 - -

Loop 2 Loop 1 2 e "

s v .05000 --

> < = -- Loop 3&4 m

o -

o 00 e o o o a o o o o o o o o = c

o. a. o. o. . .

o o o o e o g

o.-. N m e o W w TiuE :SEC)

FIGURE d.2.7 *R.CISIECS I'i T'?.E REACTCR CORE .MiD CCOLMIT LCOPS FOLLOWING A DCU3LE E'IDED RU?TURE OF A "_C'i STE.O!-

LI'IE DOWNSTRE.Ci 0F 'F.E STE!_LI"E ISCLAT*CN l 4.28 VALVES AT 5'; PCWER , ALL LOOPS ACTI'7E l

l l

. 1 g 2400.0

.C ..

M --

W 2200.0 - -

a. .- ..

(

= < ..

f 2000.0 - -

= .. -

C e

e 1800.0 - -

)

l w I M

)

4 l

w .

I l C <.

.. 1

.a C

j 1500.0 -- .

x -

w ..

- w --

$ " 1000.00 -- ..

C e ..

w m --

2 ow 500.00 --- 1 x ..

I C ..

m

@ pn w.w w

E L

1200.0

(

l l

1000,00 --

e e 500.00 --

w e --

5 e

600.00 - -

Loop 2 Loop 3&4 \

w e N A  %

I --

3 400.00 - -

M Loop i 200.00 --

0.0 C C C 2 C

C C

C C C C C C C C C c C. . .

C C.

- C C. C C m e- 0 4.0 C == tu s

• T 31.y f. f6 C. * .* 9..I

Tt"'"V

• *v N97 4 . 7. . OQ *a . .t .e

• 5='. . .C

.5%4JA '.u.

=. .*;; *7 ..?..".r*.%

3. ;. ..;..C .c.J'6

.s*-.

.ss .) s .r.s.

a= . gg r=

o v.;;..s. * ..

a r.s

0L,v~ . G A ,sC.t.:r. _s..nen :J. . o. . .e. L,;. a , s_

-L*.,.. .: . e n.<-. - .._

'I 29 DC'a7iSTR"AM 0 T:-II S I.U".LINE . , . .

ISOLA""!ON g r.3.

v. :. ..i .c,. ,. r, C.

w:.R . .% -v 0., .. e aC .

l

r- . . .

n

> - i 2500 O 2400-

'J ( > .

2300-

[(1,4) 22 L'() & 3%)

210(I -

200(i -

Test No. 5 P wer Level (1%)

1105 1900 .

Pressure, 5 (2,3)

Psig 180(1 (3%)

170(1 16ml. ,

1500

,e 1400 4y

' Limited by llot Leg

/ Subcooling = 20*F at 3%

f 130(l ,

./

1200 . . .

540 560 580 480 500 520 460 Active Cold Leg Temperature. *F FIGtlitE 4.3.1 NAltlHAL ClitClit.ATION TEST CONDITIONS

120 h

• i 100 Peak Clad Tenperature = 1800"F 00 llcal l'ransfer Coefficient Pressure si 2600 psia Blu/Hr"F Ft.2

-1600 psia 40 l

20

/

a A a a a a 8 a a a a 0.5 0.6 0.7 0 ,1 0.2 0.3 0.4 0

Fq x Power fraction l

FIGURE 4.3.2 IIEAT TRANSFEll COEFFICIENT VS. IIEAT FLUX FOR CLAD TlMPERATilRE OF Ifl00 F ,

_ _ _ _ _ _ - _ _ - . _ _ _ _ _ - - _ _ _ _ _ - _ _-