ML19347D013

to Special Low Power Tests:Safety Evaluation.

ML19347D013
Person / Time
Site:	Farley
Issue date:	01/31/1981
From:	ALABAMA POWER CO.
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NUDOCS 8103100531
Download: ML19347D013 (51)

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i I FARLEY UNIT 2 i SPECIAL LOW-POWER TESTS i

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SAFETY EVALUATION ..

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- JANUARY, 1981 I

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1.0 INTRODUCTION

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SUMMARY

In an eff ort to meet the NRC regulatory requirements of NUREG-0694 "TMI-Related Requirements for New Operating Licenses", special tests l

similar to those perf ormed at Sequoyah f or reactor power levels at or

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below 5% of Rated Thermal Power are proposed. These tasts would

demonstrate the plant's capability in several simulated degraded modes

,. of operation and would provide opportunities for operator training. The basic mode of operation to be demonstrated is natural circulation with i various portions of the plant equipment not operating, e.g., pressurizer heaters, loss of offsite power (simulated), loss of onsite AC power l'

(simulated), and RCPs for plant cooldown.

Westinghouse has reviewed the proposed tests and has determined, with the exception of TVA proposed test 8 (startup from stagnant conditions),

that with close operator surveillance of parameters and suitable opera-tor action points in the event of significant deviation from test condi-i tions, that the tests as outlined in the Farley Special Test procedures are acceptable and can be performed with minimal risk. It is recognized

,~ that in order to perf orm these tests some autematic 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 that several technical specification requirements will not be met while either preparing f or or ,

j_ performing these tests. Again Westinghouse has determined that the low 1

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

Westinghouse has reviewed the effect of the proposed test conditions on the incidents and faults which were discussed in the Accident Analysis section of the Farley Final Safety Analysis Report. In most cases, the FSAR discussion was found to bound the consequences of such events occurring under. testing conditions. Consequences of an ejected RCCA have not been analyzed-because of the low probabilities. For some inci-dents, because of the far-off-normal conditions, the analysis methods S

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available have not.shown, with reliance on auto =atic protection system

, cetion alona, that the FSAR analyses are bounding. In those cases reliance is placed on expeditious operator action. The operator action points as defined will provide protection for such events.

After perfor=ance of Special Low Power Test Programs at North Anna and Sequoyah, Westinghouse has determined that use of core exit ther=o-

_ couples and wide range loop RTDs are acceptable for determination of margin to saturation temperature under natural circulation flow condi-ti ns. This determination was based on comparison of the average core exit thermocouple temperature to average of the wide range loop RTD's T.H It was f und in both cases that the co=parison resulted in agree-ment to within 1 F. A further comparison was made between full core, incore flux map assembly FAH ilues and the core exit ther=ocouple readings. This comparison resulted in the conclusion that the tempera-ture distribution indicated by the thermocouples agreed reasonably well with the power distribution indicated by the flux map. Based on the above, Westinghouse has concluded that core exit thermorouples and wide range RIDS are reliable means of determining margin to saturation tes-perature, the thermocouples for transient and equilibrium conditions, and the RTDs for equilibrium and slow transient conditions in plants with and without Upper Head Injection.

During performance of cooddown with the reactor critical, data was taken

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to determine the excore detector response as a function of vessel down-comer temperature. In both plant tests the error in indicated power, introduced by the decreasing temperature, was less than 0.5%/1 F.

This is less than half the error assumed in the Special Test accident analyses.

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2.0 DESCRIPTION OF TESTS 2.1 COOLDOWN CAPABILITY OF THE CHARGING AND LETDOW 3YSTEM (TEST 1)

Objective - To determine the capability of the charging and letdown system to cooldown the RCS with the steam generators isolated and one RCP operating.

Method - With the reactor shutdown, trip two of the RCP's and isolate all steam generators. Vary the charging and letdown flows and monitor the primary system temperatures to determine the heat remoial capability.

2.2 NATURAL CIRCULATION TEST (TEST 2a)

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

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

2.3 NATURAL CIRCULATION WITH LOSS OF PRESSURIZER HEAT RS (TEST 2b)

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 off the pressurizer heaters at the main control board. Monitor .the system pres-sures to determine; the effect on saturation margin and the depressur-ization rate. Demonstrate 'the effects of charging / letdown flow and steam generator' pressure on the saturation margin.

', 2.4 NATURAL CIRCULATION AT REDUCED PRESSURE (TEST 2c)

Objective - To demonstrate the ability to =aintain natural circulation at reduced pressure and saturation margin. The accuracy of d2e satura-tion meter will also be verified.

Method - The test method is d2e same as for Test 2b, with the exception that the bressure decrease can be accelerated with the use of auxiliary pressuriz[r sprays. The saturation margin will be decreased to approxi-mately 20 F.

l 2.5 NATURAL CIRCULATION WITH SIMULATED LOSS OF OFFSITE AC POWER (TEST 3)

Objective - To demonstrate that following a loss of offsite AC power, natural circulation can be established and maintained while being powered from the emergency diesel generators.

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

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

2.6 SIMULATED LOSS OF ALL ONSITE AND OFFSITE AC POWER (TEST 6)

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

Method - The reactor is shut down and all RCP's are running. A total station blackout is simulated. Instrument and ligs ting power is provided by the backup batteries since the diesels are shutdown.

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. 2.7 EFFECT OF STEAM GENERATOR SECONDARY SIDE ISOLATCN ON NATURAL CIRCULATION (TEST 4)

Objective - To deter =ine the effects of steam generator secondary side isolation on natural circulation.

Method - Establish natural circulation conditions as in Test 2a but at 1% power. Isolate the feedwater and steam line for one steam generator

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and establish equilibrium. Return the steam generators to service in reverse order.

2.8 ESTABLISHMENT OF NATURAL CIRCULATION FROM STAGNANT CONDITIONS Westinghouse does not believe that it is advisable to perform this test as noted in a letter from T. M. Anderson, Westinghouse, to H. Denton, NRC, NS-TMA-2242, April 29, 1980. ,

2.9 FORCED CIRCULATION COOLDOWN (TEST 5)

.,_ This test establishes the initial conditions necessary for hest 4, Effect of Steam Generator Secondary Side Isolation On Natural Circulation. Any evaluations, technical specification exceptions or analysis necessary to bound the consequences of this test are the same as Tes t 4. (Note: This test (5) is performed as part of test 4 above) ,

2.10 BORON MIXING AND C00LDOW (TEST ,Q Objectve - To demonstrate that the RCS can be uniformly borated while in natural circulation. Also demonstrate the capability to cooldown the RCS in the natural circulation mode.

Method - Establish natural circulation based upon core decay heat gener-ation following the plant 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> NSSS acceptance test. Borate the RCS by approximately 100 ppm through the normal boratien path.

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3.0 IMPACT ON PLANT TECHNICAL SPECIFICATIONS In the evaluation of the proposed tests '4estinghouse has determined that twelve technical specifications 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 which they will not be met. The following notes the reasons these

' specifications 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 as a function of power, RCS pressure (pressurizer pressure) and loops operable. These limits provide protec-tion by insuring that the plant is not operated at higher temperatures or lower pressures than those previously analyzed. The core linits in v- the Farley tech specs are for three loop operation. Obviously when 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 wir.h

' limits on core exit temperature (< 610 F), T, (< 578 F) and Loop AT (< 65 F) such that no boiling will be : experienced in the core and the limits of specification 2.1.1' for temperature will be met.

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The limits will not be met simply because less than three 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 conditions by tripping the plant when various process pa.ameters

. exceed their analyzed values. When in natural circulation two trip functions will be rendered inoperable, Overtemperature AT and Over-power AT. There is a temperature input to these functions which originates from the RTD bypass loops. Due to the low flow conditions, 5% or less, the temperature indications from these loops will be highly e

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suspact. To prevent the inadvertent tripping of the plant when in the

, natural circulation mode these functions will be bypassed. Their pro-tection functions will be perfor=ed by the operator verifying that Pres-suriser Pressure and Level, Stea= Generator Level, and subcooling margin (Tsat) 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 unco = mon for this function to 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. The amount of water is based on the decay heat present in the core and to prevent dryout of the st.eam generators. With the plant limited to 5%

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 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

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lower the trip setpoint from 17% span to 5% span for all of the special tests. In addition, the Steam Generator Low-Level se point which is part of the steam /feedwater mismatch trip may be lowered from 25% span to 5% an.

3.1.3 T.S. 3.1.1.3 MODERATOR TEMPERATURE C0t.crICIENI The Moderator Temperature Coefficient is limited to O pcm/0F or nore negative. When performing tests with the plant critical below 5410F this coefficient may be slightly positive. However, it is expected that the Isothermal Temperature Coefficient will remain negative or approxi-mately.zero. The tests will be performed such that this is the case and thus minimizing any impact from rapid heatups or cooldowns. 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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', 3.1.4 T.S. 3.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY The Minimum Temperature for Criticality is limited to 541 F by spec.

3.1.1.4 and 531 F by spec. 3.10.3. To perform test 4 it is expected that the RCS average temperature will drop below 531 F. Westinghouse has determined that operation with T,y as low as 485 F is accept-able assuming that:

l. Control Bank D is inserted to no deeper than 100 steps withdrawn, and

2. Power Range Neutron Flux Low Setpoint and Intermediate Range NeEtron Flux reactor trip setpoints are reduced from 25% RT? 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 prinary trips.

3.1.5 T.S. 3.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION 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 l

l To prevent inadvertent Safety Injection and to allow performance of the

( special tests, all autenatic Safety Injection functions will be blocked. Indication of partial Safety Injection logic trips and manual t

i initiation will be operable, however, the automatic Safety Injection actuation functions will be made inoperable by forcing the logic to see that the reactor trip breakers are open. Westinghouse believes that L this mode of cperation is acceptable for the short per,iod of time these tests will be carried out based on the following:

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

2. Rigid adherence to the operator action points as defined by Westinghouse, see Section 3.2.

3. Little or no decay heat is present in the syste=, thus Safety Injec-tion serves primarily as a pressuri:ation function.

Blocking these functions will allow the perfor=ance of these tests at low power, pressure, or temperature and close operator surveillance will assure initiation of Safety Injection, if required, within a short ti=e period.

• The actuation setpoint for the auxiliary feedwater pumps is also a ff ec t'ed. The actuation setpoint is lowered frem 17% span to 5% span for all the special tests. With the plant limited to 5% RT? or less and being a BOL on Cycle 1, there will be little or no decay heat present.

The heat source will be the core operating at the limited pcwer level.

Westinghouse finds that initiating the auxiliary feedwater penps at the lower setpoint meets all the applicable requirements.

3.1.7 T.S. 3.4.4 PRESSURIZER The Pressurizer provides the means of maintaining pressure control for l- 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 sciutained through the use of pressurizer level and charging / letdown flew.

i 3.1.8 T.S. 3.7.1.2 AUXILIARY FEEDWATER SYSTEM 1

l 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.,

I motor driven auxiliary feedwater pumps inoperable. Westinghouse has

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, determined that this is acceptable for these two tests because of the little or no decay heat present allcwing sufficient time (# 30 minutes) for operating personnel to rack in the pu=p power supplies and regain steam generator level.

3.1.9 T.S. 3. 8.1.1, 3. 8. 2.1, 3. 8. 2.3 POWER SOURCES i

These specifications are outside Westinghouse control, however it is 1

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acceptable to alter power source availability as long as manual Safeg(y Injection is operable and safety related equipment will function when .

required.

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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 ar 5310F. Since it is expected. that RCS Tayg will be taken as low as 5300F this specification will be excepted. See Section 3.1.4 for basis of acceotability.

3.1.11 TECHNICAL SPECIFICATIONS NOT EXCEPTED While 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 - q F (g)

( At low temperatures and flows F q(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 - (F6H)

At low temperatures and flow F6H can be expected to be higher i than if pumps are running. However, no significant consequences for full power operation are expected.

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. 3. 3.2.4 QUADRANT POWER TILT RATIO

, With no, one, two, or three pu=ps running and critical, core po er ,

distributions resulting in quadrant power tilt nay for=. At low power levels and for short perieds of tices these tilts will not significantly influence core burn-up.

4. 3.2.5 DNB PARA.ETERS In the performance of several tests d2e plant will be depressurized below 2230 psia. At low operating power levels this depressuriza-tion is not significant as long as subcooling targin is =aintained.

3.1.12 SPECIAL TEST EXCEPTIONS

1. Special Test Exception Specification 3.10.3 allows li=ited excep-tions for the following:

3.1.1.3 Moderator Temperature Coefficient 3.1.1.4 Minimum Temperature for Criticality 3.1.3.1 Movable Control Assemblies r._.

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 1Lsited 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 =ust =eet the folicw-ing set of criteria for operation:

1. Maintain For All Tests a) Primary System Sub-cooling (T,,e Margin) > 20 F b) Steam Generator Water Level > 30% Narrow Range Span c) Pressurizer Water Level (1) With RCPs running > 22% Span (2) Natural Circulation > Value when RCPs tripped d) Loop AT ;c 65 F e) T avg < 578 F f) Core Exit Temperature (highest) ;c 610 F g) Power Range Neutron Flux Low Setpoint and Intermediate Range Neutron Flux Reactor Trip Setpoints j:,7% RTP

___ h) Control Bank D 100 steps withdrawn or higher i) T > 485 F eoid

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

a) Primary System Sub-cooling (T,,e Margin) ;i 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 Wcter Level < 17% Span or an unexplained decrease of more than 5% not concurrent with a T,yg change 4

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. e) Any Loop AT > 65 ?

. "f) T avg > 578 7 g) Core Exit Temperature (highest) > 610 ?

h) Uncontrolled rod motion i) Control Bank D less than 100 steps withdrawn j) T eold < 485 7

3. Safety Injection must be manually initiated if any of the following condi-tions are met:

a) Primary System Sub-cooling (T sat Margin) -

< 10 ?

b) Steam Generator Water Level < 0% Narrow Range Span or. Equivalent Wide Range Level c) Containment Pressure > 4.7 psig d) Pressurizer Water Level < 10% Span or an unexplained ,

decrease of more than 10% not concurrent with a T,y

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~ Decreases by 200 psi or more

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e) Pressurizer Pressure in an unplanned or unexplained manner.

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• Safety Injection =ust not be terminated until the *4estinghouse criteria as defined in E01:E-2, Loss of Secondary Coolant are =et.

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

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 cdequate secondary side heat sink,

3. sufficient level in the pressuri:er 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 TECE'iICAL SPECIFICATION IMPACT Test Technical Specification 1 2a 2', 2c 3 4 6 7 2.1.1 Core Safety Limits X X X X X X 2.2.1 Various Reactor Trips Overtemperature AT X X X X X Overpower AT I X X X X Steam Generator Level X X X X X 3.1.1.4 Moderator Temperature coef- X X ficient 3.1.1.5 Mi.nimum Temperature for X X Criticality ,

3.3.1 Various Reactor Trips overtemperature AT X X X X X Overpower AT X X X X X Steam Generator Level X X X X X

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3.3.2 Safety Injection - All X X X X X X X automatic functions Auxiliary Feedwater Initiation X X X X X X X

, 3.4.4 Pressurizer X X X l

3.7.1.2 Auxiliary Feedwater X 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 X X i 3.10.3 Special Test Exceptions - X X Physics Tests p

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. 4.0 SAFETT EVALUATION In this section the safety effects of those special :es: conditions which are outside the bounds of cenditicas assu=ed in :he FSAR are evaluated. The interaction of these conditions with :he ::ansien:

analyses in the FSAR are discussed.

4.1 EVALUATION OF TRANSIENTS The effect of the unusual operating conditions en the transients analyzed in the FSAR are evaluated.

a 4.1.1 CONDITION II - FAULTS OF MODERATE FREQUENCY 4.1.1.1 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal frem a Suberitical Condition Restriction of control rod operatica to manual control, and constant operator monitoring of rod position, nuclear power and temperatures t greatly reduces the likelihood of an uncontrolled RCCA withdrawal.

Operation without reactor coolant pumps, and in seme cases with a posi-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 concentratica vill be such that it will not go critical with less than 100 steps withdrawal on D Bank. An analysis of this event is presented in Secticn 4.2.1. For Test 3b, this transient is bounded by the FSAR analysis, since all reactor coolant pumps are operating.

4.1.1.2 Uncontrolled Rod Control Cluster Asse=b1v 3enk Withdrawal at Power The same considerations discussed in Paragraph 4.1.1.1 apply here. In addition, the low' operating power and the Power Range Neutron Flux Lev and Intermediate Range Neutron Flux trip setpoints a,ct to =itigate this 4-1

incident, while lack of the Overte=perature aT trip re= oves so=e of

. the protection provided in the FSAR case. An analysis is discussed in Paragraph 4.2.2.

4.1.1.3 Rod Control Cluster Assembly Misaliennent The FSAR discussion concerning static RCCA misalignment applies to the j _. test conditions. he consequences of a dropped RCCA would be a decrease in power. Thus n increase in probability or severity of this incident is introduced by 9e test conditions. '

I Uncontrolled Boron Dilution 4.1.1.4

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The consequences of, and operator action time requirements for, an uncontrolled boron dilution under the test conditions are bounded by those discussed in the FSAR. The fact that the control rods will never be inserted to the insertion limits, as well as the Pcwer Range Neutron Flux Low Setpoint and the constant operator monitoring of reactor power, temperature and charging system operation, provides added protection.

4.1.1.5 Partial Loss of Forced Reactor Coolant Flow

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Because of the low- power limits the consequences of loss of reactor

, coolant pump power are trivial; indeed they are bounded by normal cpera-ting conditions for these tests.

l 4.1.1.6 Startup of an Inactive Reactor Coolant Loop l

When at least one reactor coolaat pump is operating, the power limit for

.these tests results in such small temperature differences in the reactor t

coolant system that startup of another loop cannot introduce a signifi-i cant reactivity disturbance. In natural circulation operation, inadver-tent .startup of a pump .aould reduce the core water tesperature and thus

- provids a change in reactivity and power. Because of the small modera-tor : reactivity coefficient at beginning of life the power increase in i

the worst condition would be small and gradual and the flow-to power l-4-2

ratio in the core would be increasing. The ?:ver Range Neutr:n Flux '_cv Setpoint reactor trip provides an upper bound on power. 3e:ause of the 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 Trio

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Because of the low power level, the disturbance caused by any loss of 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 loss 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 th=e is available to re-institute auxiliary feedwater sources.

.- 4.1.1.9 Loss of Offsite Power to the Station's Auxiliaries (Station Blackout)

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 Systen Malfunctions The main feedwater control valves will not be used while the reactor is at power or near criticality on these tests. Thus, the pote=tial water flow is restricted to the main feedwater bypass valve flew or auxiliary feedwater flow, about 15% of normal flow. The transient is further mitigated by the low operating power level, msall moderator te=perature reactivity coefficient, the low setpoints on the Inter =ediate and Power Range Neutron Flux Low setpoint trips, and close operator surveillance of feed flow, 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 Load Increase Incident The turbine will not be in use during the performance of these tests, and load control will be limited to operation of a single steam dump or steam relief valve. The small moderator te=perature reactivity coeffi-cient also reduces the consequences of this 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 Inter =ediate-Range Neu-tron Flux reactor trips limit any power transient. In addition, modifi-cation of the High Steamline Flow setpoint allows a reactor trip on Low Steam Pressure only. Analyses are discusssed 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 depressuriza-tion of the RCS. In addition, automatic reactor trip caused by the Low Pressurizer Pressure Safety Injection signal would occur when core out-let subcooling reached approximately 10 F as an automatic backup for manual trip. During tests 2b and 2c, when ta p trip is bypassed to allow deliberate operation at low pressure, the pressuri'zer PORV block l_ valves will be closed to remove the major credible source of rapid l inadvertent depressurization. (The Low Pressure trip is automatically I reinstated when pressure goes above 2000 psig and the PORV block valves i .

l will be reopened at that time.)

l l 4.1.1.13 Accidental Depressurization of the Main Steam Svstes

! The FSAR analysis for accidental steam system depressurization indicates that -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 er.ist during the-O t

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test pariod, the reactor would remain suberitical even if it were cooled to room temperature without Safety Injection. Thus the SAR analysis is bounding.

4.1.1.14 Spurious Operation of the Safety Injection System at 'over In order to reduce the possibility of unnecessary thermal fatigue cycling of the reactor coolent system components, the actuation of high head charging in the safety injection mode, and of the safety injection pumps, by any source except minual 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 primsry 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 Emergency Core Cooling 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 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 suffi-cient 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 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.

e A -_

i 4.1.2.2 Minor Secondary System Pine 3reaks The consequences of minor secondary syste= pipe breaks are within the bounds discussed in Paragraph 4.2.3.

4.1.2.3 Single Rod Cluster Control Assembly Withdrawal at Power The FSAR analysis shows 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 FSAR ,

points out, no single electrical or mechanical failure in the control 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 the restriction to manual control, and by the close operator surveillance of reactor power, rod operation, and hot leg

temperature.

4.1.2.4 Other Infrequent Faults The consequences of an inadvertent 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 FSAR, have been reviewed and found to bound the consequences of such events occurring during test operation.

4.1.3 CONDITION IV - LIMITING FAULTS 4.1.3.1 Major R'accor 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. During the large break event the system inventory and cold leg accumulators will have removed enough energy to hav,e, filled the. reactor vessel to the bottom of the nozzles. Following the system depressurization there is enough 4-6

water in the reactor vessel below the no::les to keep the core covered .._ .

for over one hour using conservative assumptions. This is sufficient time for the operator to manually initiate SI and align the system for long teds cooling. At no time during this transient will the core be uncovered.

It must be noted enat the magnitude of the resutting 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 pr'essurizer pressure, cold leg temperature, and reactor power, with specific initiation criteria for reactor trip; low trip setpoints on the Intermediate-Range and Power-Rarge Neutron Flux trips; Low Flow Mismatch setpoint for Reactor Trip and MSIV closure'on Low Steam Pressure; and Low Pressurizer Pressure trip (S.I. initiation) assure a Reactor Trip without excessive reactor power following a cool-down transient caused by the secondary systc;i. Following reactor trip, assuming the worst RCCA stuck out of the core, the reactor would remain suberitical even if it were cooled to room temperature. Transient analyses for a steam pipe rupture 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.

l 4.1.3.3 Steam Generator Tube Rupture The steam generator tube rupture event may be categorised by two dis-tinct phases. 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 maall break LOCA L __

_u

~

trensients, and the operator actions required to deal with this situs-

. tion during this phase are identical to those required for mitigation of a maall LOCA. Hence, evaluation of the steam generator tube rupture during this phase is wholly covered by the safety evaluation of the small LOCA.

Af ter the appropriate operator actions have taken place to deal with the initial LOCA phase of the event, the remainder of the steam generator [

1 cube rupture accident mitigation would consist of those operator actions t required to isolate the faulted steam generator, cooldown the RCS, and depressurize the RCS to equilibrate primary RCS pressure with the l f aulted 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 faulted steam generator.

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

4. RCS depressurization capability.

Evaluation of the Farley special test procedures has verified that all of the above systema 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 the SAR regarding this event remain bounding.

4.1.3.4 Single Reactor Coolant Pumo Locked Rotor Because of the low power level, the locking of a single reactor coolant pump rotor is inconsequential.

4-8

4.1.3.5 Fuel Handling Accidents The FSAR analysis of fuel handling accidents is bounding.

4.1.3.6 Rue.ture'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 100 steps withdrawn) that the worth of an ejected rod will be substantially less than the delayed neutron frac-tion. Thus, the power rise following a control rod ejection would be

~

relatively gradual and ter=inated by the Power Range and Intermediate Range 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 the FSAR, the result of combining these ame-liorating 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 ths brief period in the test sequence does not warrant such an analysis.

g._

4.2 ANALYSIS OF TRANSIEhTS 4.2.1 ANALYSIS OF RCCA BANK WITHDRAWAL FRO}i 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:

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 100 steps withdrawn.

3. A typical moderator temperature reactivity coefficient was used

-(positive) for any core average temperature at or above 485 F.

4-9

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 1300 F, 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 BANK WITHDRAWAL AT POWER Analyses of RCCA bank withdrawal transients were perforned for natural circulation 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 maxican for two banks moving was assumed for cases with all steamlines open, and up to the maximum for one bank moving for the cases with two steamlines iso- .

laced. Both upper and lower bounds on typical reactivity feedback coeffiaients for beginning of life, Cycle 1, were investigated. In all cases,-reactor trip was initiated at 10% nuclear power.

Reactor conditions at the time of maximum core heat flux are shown in Figures 4.2.3 and 4.2.4 as functions of the reactivity insertion rate for three loop active cases. For high reactivity insertion rates, the

. minimum reactivity coefficient cases giva the greatest 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 for so long that the core 4-10 6761Q _ _

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

after approximately one and one-half minutes of continuous withdrawal.

Thus, the cases shown bound the worst cases.

4.2.3 ANALYSIS OF COOLDOWN TRANSIINTS Cooldown transients include feedwater system malfuncticas, excessive steam load increase, accidental depressurization of the main steam sys-tem, and minor and major secondary system pipe uptures. Attention has been focused on 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 to a criti-cal condition.)

During natural circulation operation, approximately one to two minutes would elapse following a secondary side event before cold water frem the steam generator reached the core; thus, considering the close and con-stant surveillance during these tests, tine would be available for the operator to respond to such an event. Analyses were also performed to determine the extent of protection provided by cutomatic 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 vith natural cir-culation, assuming the bounding negative moderator temperature coeffi-cient for the beginning of life, Cycle 1. Thus no autematic protection .

is required, and ample time is available to the operator to trip the reactor, isolate feedwater to the faulted steam generator, and isolate i the_ break to the extent possible. Calculated results for the sudden i opening of a single steam valve, assuming the most negative BOL Cycle one moderator reactivity coefficient and 5% initial power are shown in i

! Figures'4.2.5 and 4.2.6. .

t E

4-11

e 4.2.3.2 High Flux Protection Reactor trip en high nuclear flux provides backup protection for larger pipe breaks or load increases. Analyses were perfor=ed to deter =ine the worst cora conditions that could prevail at the time of high-flux trip, independent of the cause. The following assu=ptions were used:

Upper-bound negative moderator isothermal temperature coefficient,

~

1.

vs. core average temperature, f or beginning of life, Cycle 1.

2. Lower-bound fuel te=perature power reactivity coefficient.

3. Initial operation with core inlet terperature 555cF.

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

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

This results in some boiling in the reactor. The negative reactiv-ity introduced by core boiling would effectively limit power; this negative reactivity was conservatively neglected.

6. Uniform core inlet temperature and flev.

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 from the true power by 1% for each loF decrease in reactor inlet temperature, resulting in a true power of greater than 10% at the time of trip.

8. Core flow rate as a function of core power was assu=ed equal to the predicted flow under steady-state operating conditions.

e e-RR

Analyses of core conditions based on these assu=ptions indicate that the DNB criterion of the FSAR is met.

4.2.'3.3 Secondary Pressure Trio Protection Large steamline ruptures which affect all loops uniformly will accuate reactor trip and steamline isolation on Low Steamline Pressure signals in any two lines. Low Pressuriser Pressure and Power Range Neutron Flux low setpoint trips serve as further backups. An example is the double ended rupture of a main steamline downstream of the flow restrictors with all steamline 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. In 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 l

l 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 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 safety 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 l precluded either by analysis or identified operator action. For these cases, the basis for acceptability was primarily associated with the low L

probability of an inadvertent rod withdrawal event during the limited

~

duration of the special tests.

This section provides an additional assessment relative to the potential for and consequences of fuel failure for these "unanalyzed" accident conditions associated 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 appli-cability. Additional information (attached) is provided for four areas:

1. Thermal margin associated with normal test conditions.

2. The potential for DNB during accident conditions.

3. The clad temperature response assuming that DN3 occurs.

4. Radiological consequences associated with presumed gross fuel failure.

The conclusions of this assessment are as follows:

l l

1. DNB is not expected for the liniting thermal condition associ-ated with any RCCA withdrawal event.

~

2. Even assuming DNB, there should be adequate heat transfer to prevent clad overheating.

l l

3. Fuel clad failure is not expected.

4. Even assuming 100% clad failure and other extreme conservatiscs, the resulting offsite dose would be anall.

4.3.1 DESIGN CONSIDERATIONS Margin to hot channel boiling has been incorporated with all normal test j

conditions by establishing a lower bound requirement on the degree of L 8-Re __

reactor coolant subcooling. This test requirement assures that postu-laced accidents are initiated from a condition of excess thernal margin.

4.3.2 DNB CONSIDERATIONS 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 b

4.2.3.2 result in mass velocity below the range of direct applicability I

of the correlation, the reactor heat flux was so low relative to the i

predicted critical heat flux that even a factor of 2 would not result in seri6us concern for DN3 for this event.

For the non-cooldown transients the ILniting 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 DN3 correlation to low flow DN3 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-

~ ence 3) at these pressures are higher than those predicted by the low flow correlations. Further review of the data in reference 1 indicates that the critical heat flux at higher pressure is significantly lower than the above data at 1000 psi. The minimum critical heat flux of the data set is .16 x 106 BTU /hr-ft2 for a data point at 2200 psia at a mass velocity of .2 x 106 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 nnelear feedback from voiding. The power distribution would tend to peak towards the bottom thus further reducing the local quality at the peak flux-locations.

6-E -

,Also the pool boiling correlations in reference 3 show some decrease in critical heat flux above 1000 psia to the maximum pressure of appli-cability of 2000 psia. However extrapolation of the correlations to a value of zero critical heat fimc at the critical pressure (3206.2 psia) would not result in lower critical heat fluxes than shown in the data set from reference 1. Since the core average heat flux at 10% of nomi-nal power (highest expected power for heatup events) is only on the order of .02 x 106 BTU /hr-ft2 a large peaking factor would be required to put the reacter heat flux as high as the critical heat flux.

For the transients considered, the only ones that lead to significant o,ff normal peaking factors are rod motion transients. The rod with-drawal from suberitical is a power burst concern. As such, it is expected that even if DNB occurred, the rod surface would revet. 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 the lack of data, it is difficult to completely ;2 sclude DNB, although a prudent judgement indicates that it is indeed remote.

w-4.3.3 CLAD TEMPERATURE CONSIDERATIONS Should DNB occur, the peak clad temperature reached would depend prima-rily on the local nuclear transient following DNB and on the behavior of the post-DNB heat transfer coefficient.

For a rapid power transient, as lliustrated by the SER analysis for l

RCCA' bank withdrawal fres a sube: al condition, the fuel temperature reactivity feedback and reactor tr. a nuclear flux signal would shut I down the reactor before sufficient e rgy could be generated to cause a damaging rise in clad temperature. In that case, the maximum clad tem-perature calculated was under 13000F even assuming an extremely low heat transfer coefficient (# 2 BTU /hr-ft2 op),

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 v , ,

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

In this state the clad temperature could he 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 setpoint is at 7% for these tests. To allow for calorimetric errors and norsal system errors, trip is assumed to occur at 10% of rated thermal power (RTP), unless a large decrease in downcomer,. coolant temperature occurs during the test. In tests 3 and 5, depressurization to less than approxi=ately 1450 psia could require te=perature 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 trans fer coefficient required to keep the peak clad te=perature below 18000F, the threshold of significant heat flux increases due to zirconium-water reaction, can be found from Figure 4.3.2.  :

. Various fibs 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 coefficients. The Westinghouse. U'dI film boiling correlatica (reference 4), was developed at low flow condi-tions sisilar to those postulated for incidents occurring during the PSE&G 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 B,TU/hr-ft2 _op)a,e at 2200 psia and 5% flow with quality e

, 4-17 -

n between 10-50%. Other film boiling heat transfer correlations, devel-oped at higher pressures, were also examined. These correlations were extrapolated down to the lower flow conditions of the ?SESG tests as another approach to obtain representative fil= boiling coef ficients.

Using both the Mattson et al (reference 5) and the Tong (reference 6) film boiling correlations resulted in post-DN3 heat transfer coeffi-cients in excess of 150 BTU /hr-ft2 - F at the conditions 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 accident senario using conservative assumptions so as to deter =ine an extreme upper bound on postulated accident consequences. The analysis assumed a reactor accident involving no pip,-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 were assumptions made in the analysis which include:

133 Mwt (5% power) 1.0 dose-equivalent I-131 RCS activity (tech spec limit)

,_ 500 gpd steam generator leak in each SG (tech spec limit) l 100% clad damage and gap activity release 10% iodine / noble gas in gap space 100 DF in steam generators 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/m3 x/Q percentile value The results of the analysis show that the two hour site boundary doses would be 5 ren thyroid, 0.9 rem total body and 0.4 rem to the skin.

O e

4-18

4

.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:

1) Equilibrium radionuclide inventories established at 5% power. For iodines, this requires # 1 month of steady state operation at 5%

uninterruoted.

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

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

3) 100% fuel rod clad desage.

4) Primary to secondary leakage to tech spec values. Since Farley 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 better diffusion characteristics and thus lower offsite doses.

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

l 4.3.5 OTHER CONCERNS The LOCA analyses presented indicate that there are over 6,000 seconds for the operator to take action. This is more than sufficient time for the operator to take corrective action. Some transients were not l

l analyzed or discussed in this supplement due to the ce=bination of the g

low probability of the transient occurring and the very short time 4 -

tu . PMARn

e 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 for a condition II transient given here indi-cate that these events do not need to be analyzed. Sinilar 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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4-20

e REFERENCES t

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

Stanek, " Correlation of Critical Heat Flux in a Sundle Cooled by Pressurized Water," Sydposiu= on Two-Phase Flcw in Rod Bundles, Code 1,127, ASME Winter Annual Meeting, November, 1969.

2. Hao, B. R., 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 Society,1977. ,

4. WCAP-8582-P, Vol. II, " Blowdown Experiments With Upper Head Injec-tion in G2 17x17 Rod Array," McIntyre, B. 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 Boilles Data," paper B3.8,

~~

Vol. 4,. Proc. of 5th Int. Heat Transfer Conference, Tokyo, Fepte=ber (1974). -

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

Engng. and Design 6,, 301 (1967).

G 4-21

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