Text

Special Low Power Tests Safety Evaluation.
	ML19331D948
Person / Time
Site:	Farley
Issue date:	08/31/1980
From:	; ALABAMA POWER CO.
To:	;
Shared Package
ML19331D947	List: ML19331D946; ML19331D948; ML19331D949; ML19331D950; ML19331D951; ML19331D952; ML19331D953;
References
	NUDOCS 8009040387
	Download: ML19331D948 (47)
	v • d • e

.

O,. .

'

, .

O FARLET UNIT 2 SPECIAL LOW-POWER TESTS

~"

SAFETY EVALUATION S-

.

AUGUST 1980

%

O k.

I

_

!

80 00040 3 f J

-. ___

\

. .- .

'. -

1.0 INTRODUCTION

AND

SUMMARY

g

,

i In an effort to meet the NRC regulatory requirements of NUREG-0694 i "TMI-Related Requirements for New Operating Licenses", special tests ll similar to those performed at Sequoyah for reactor power lavels at or below 5% of Rated Thermal Power are proposed. These tests would demonstrate the plant's capability iu several simulated degraded modes of operation and would provide opportunities for operator training. The l basic mode of operation to be demonstrated is natural circulation with b' various portions of the plant equipment not operating, e.g. , pressurizer heaters, loss of offsite power (simulated), loss of onsite AC power (simulated), and RCFs for plant cooldown.

.

Westinghouse has reviewed the proposed tests and has determined, with the exception of TVA proposed tests 8 and 9 (startup from stagnant con-dicious and boron shing and cooldown), that with close operator sur-veillance of parameters and suitable operator action points in the event of significant deviation from cast conditions, that the tests as out-

-

lined in the Farley Special Test procedures are acceptable and can be

"

performed with minimal risk. It is recognized that ih order to perfors these tests som 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 that several technical specification requirements

_

will not be met while either preparing for or performing these tests.

Again Westinghouse has determined that the low 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 a

.

. _ _ . ._

1-1

-

-- --

. - - - _ - _ _ _ _

'

avcilablo heva nst shown, with rolituca on automatic protseciou systcm

'

.

,

action alone, that the FSAR analyses are bounding. In those ececo reliance is placed on expeditious operator action. The operator action c

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 that use of core exit thermo-couples and wide range loop RTDs are acceptable for determination of margin to saturation temperature under natural circulation flow condi- 3 b i

tions. This determination was based on comparison of the average core exit thermocouple temperature to average of the wide range loop RTD's It was found in bo;n cases that the comparison resulted in agree- i Th.

ment to within 1 7. Ifurthercomparisonwasmadebetweenfullcore, incore flux map asseably Fg values and the core e:ic thermocouple 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, Wer.tinghouse has concluded that core exit thermocouples and wide

--- range RTDs are reliable means of determining margin to saturation tem- l perature, the thermocouples for transient and equilibrium conditions, ,

and the RTDs for equilibrium and slow transient condit. ions in plants with and without Upper Head Injection.

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

-

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.

-

1-2

. - - , .- - _ . . -. --. ___ . - _ _ _. . -

_ _ -

i

-

2.0 DESCRIPTION

OF ESTS

.

2.1 COOLDOWN CAPABILITY OF THE CHARGING AND LETDOWN SYSTEM (TEST 1)

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

_

Method ,- With the reactor shutdown, trip two of the RCP's and isolate '2

_

all steam generators. Vary the charging and letdown flows and monitor

.

the primary system temperatures to determine the heat removal 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 approvimmtely 3% power and all Reactor Cool-anc Pumps (RCP 's) are operating. All RCP's are tripped simultaneously with the establishment of natural circulation indicated by the core exit

-

s.

,

thermocouples and the wide range RTD's.

2.3 NATURAL CIRCULLTION WITE LOSS OF PRESSURIZER HEATERS (TEST 2b)

_

Obiective ,- To demonstrate the ability to maintain natural circulation and saturation ' margin with the loss of pressuriser 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 fecernine; the effect en saturation margin and the depressur-izatico rate. Demonstrate the effects of charging / letdown flow and I steam generator pressure on the saturation margin. l

,

o 4

l

_

2-1

., _ _ . _ . - - _ _ _ _ - -- ---- __ _ __ __ - - - - - - - - - - - - - - - - - _ _ __ _ - - - - - - _ _

.

'

- '2. 4 NATORAL CIRCULATION AT REDUCED PRESSURE (TEST 2c)

-

.

!

,

Objective - To demonstrate the ability to maintain natural circulation at reduced pressure and saturation margin. The accuracy of the satura-tion meter will also be verified. 1 Method - The test method is the same as for Test 2b, with the exception that the pressure decrease can be accelerated with the use of auxiliary pressurizer sprays. The saturation margin will be decrsased to approxi ;- g 1 mately 20 F.

7. 5 NATURAL CIRCUIATION WITH SIMUIATED LOSS OF OFFSITE AC POWER (TEST 3a)

-

I l

Objective _- To date istrate that following a loss of offsite AC power, natural circulation can be establishe.d and maintained while being powered from the emergency diesel generators.

(

'

Method - The reactor is at approximately 17. power and all RCP's are operat,ing. 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 SIMILATED LOSS OF ALL CNSITE AND OFFSITE AC POWER (TEST 3b) l

_

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

Methed - Ihe reactor is shut down and all RCP's are running. A total station blackout is simulated. Ins trument and lighting power is provided by the backup batteries since the diesels are shutdown.

m l

,

2-2

-

- - - - - .- _ _ _ __

.

,

'2. 7 EFFEN OF STwAM GENERATOR SECONDARY SIDE ISOLATON

,

'

ON NATTJRAL CIRCULATION (TEST 4) e l Obiective - To determine 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 and establish equilibrium. Repeat this for one more steam generator so , _

that two are isolated 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 FQLCED CIRCULATIGi C00LDOWN

~ . s

,

This. test is performed as preparation for the Boren liizing and Cooldown Test. Since Westinghouse does not believe it is advisable to perform

,the Boron Mixing Test as defined using core heat, it is not necessary to perform the Forced Circulation Cooldown Test.

2.10 3ORON MIIING AND COOLDOWN Wes::inghouse does not believe that it is advisable to perform this test utilizing core heat as noted in NS-TMA-2242, T. M. Anderson, Westinghouse, to H. Denton, NRC.

_

O

%

1 l

2-3

'

.

. .

'. .

3.0 IMPACT ON PLANT TECHNICAL SPECIFICATIONS

-

In the evaluation of the proposed tests Westinghouse has determined that twelve technical specifications will be violated, and thus require exceptions, during the performanes 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 ,_

1

. *~ . . . 1 during the ;:ests. l 3.1 IMPACT SUMMART 3.1.1 T.S. 2.1.1 REACTOR CORE SAFETY LIMITS -

The core limits restrict RCS T, as a function of power, RCS pressure (pressurizer pressure) and loops operable. These limitJ provide protec-tion by insuring enat the plant is not operated at higher temperature.s l

or lower pressures than those previously analyzed. The core limits in

~

th's 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 perfor:ned with limits on core exit temperature (< 610 7), T (< 578 F) and Loop AT (< 65 7) such that 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 three RCP's would be running.

3.1.2 T.S. 2.2.1 REACTOR TRIP SYSTEM INSTRmfENTATION SETPOINTS The Reactor Trip System provides protection from various transients and faulted conditions by tripping the plant when various process parameters exceed their analyzed values. When in natural circulation two trip functions will be rendered inoperable, overtemperature AT and over-pover 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 i

-.. _ '.-__ _ Sh-R ,

.

.,

,'sutpect. To prevent tho inadvIrtsat tripping of tha plcnt whsn in tha natural circulation mode these functions will be bypessed. Thair pre-tection functions will be performed by the operator verifying that Pres- c surizer Pressure and Level, Steam Generator Level, and subcooling margin l

(T,,g) are above the operator action points for Reactor Trip and Safety Injection.

t

.

l Steam Generator Level-Iow-Low is the third trip function that can be j aff ec ted. When at low power levels it is roc uncommon for this function j g, I

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 en 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 i there will be little or no decay heat present. The beat source will be the core operating. t the limited power level. Tripping the reactor on any of the different operable trip functions or the operator actica points will assure that this require-

.,_

ment will be met. Thus, Westinghouse finds that it is acceptable to lower the trip setpoint from 17% span to 5% span for all of the special ,

tests. In addition, the Steam Generator Low-Level setpoint which is part of the steam /feedwater mismatch trip may be lowered from 25% span

>

to 5% span.

3.1.3 T.S. 3.1.1.3 MODmTOR TEMPERATURE COEFFICIENI The Moderator Temperature Coefficient is limited to O pcm/0F or more 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 vill remain negative or at proxi-mately zero. The tests will be performed such that this is the case and thus minimizing any impact from rapid heacups or cooldowns. In addi-tion, the effect of a small positive Moderator Temperature Coefficient has b'een considered in the accident analyses performed for the test conditions. )

l 3 i l

9 3-2

'

.

. .

.\

'

' 3.1.4 T.S. 3.1.1.4 MINIMUM IDfPERATURE FOR CRITICALITY

,

. : C l

The Minbmma Temperature for Criticality is limited to 541oF by spec.

, .

>

l 3.1.1.4 and 5310F by spec. 3 40.3. To perform test 4 it is expseted that ene RCS average temperature will\. drop below 531oF. Westinghouse has determined that operation with T ay as low as 5300F is accept-able assuming that:

Control Bank D is inserted to no deeper than 100 steps withdrawn, and E' 1.

.

2. Power P.ange Neutron Flux Low Setpoint and Intermediate Range Neutron Fluz 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 (3ank D) on unplanned with-drawal, 2) reducing bank worth (Bank D) on unplanned withdrawal, 3)

! ==v4=izing reactivity insertion capability consistent with operational requirements, 4) limiting mav4== power to a very low value on an

uhplanned power excursion, and 5) allowing the use of the "at power" <

teactg trips as back-up trips rather than as primary trips. -

3.1.5 T.S. 3.3.1 REACTOR TRIP SYSTEM INSTRUMENIATION 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 evaluas ten.

3.1.6 T.S. 3.3.2 ENGINEERED SAFETT FEATURE ACTUATION SYSTEM INSTRUMENTATION To prevent inadvertent Safety Injection and to allow performance of the special tests, all automatic Safety Injection functions will be blocked. Indication of partial Safety Injection logic trips and manual initiation will be operable, however, the autematic Safety Inj ection

,

actuation functions will be made inoperable by forcing the logic to see that the reactor trip breakers are open. Westinghouse believes that this mode of operation is acceptable for the short period of time these

.

tests will be carried out based on the following:

- Jb51

.

. .

.

,

, 1. Closo cbs;rvation of tha ptrtial trip indication by tha opsratcr,

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

!

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 b i

low power, pressure, or temperature and close operator surveillance will assure initiation of Safety Injection, if required, within a short time

< '

peniod.

The actuation setpoint for the auxiliary feedvater pumps is also affected. The actuation setpoint is lowered from 17% span to 5% span for all the special tests. With the plant Itaited to 5% RTP or less and being a BOL on Cycle 1, there will be little or ao decay heat present.

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

--

Westinghouse finds that initiating the auxiliary feedwater pumps at the

lower setpoint meets all the applicable requirements. -

3.1.7 T.S. 3.4.4 PRESSURIZE 1L 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 pressuriser heaters will be either turned orf 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 AUXILIART FEEDWATER SYSTEM The auxil* Ary feedwater system will be rendered partially inoperable for two tests. The two tests simulate some form of loss of AC power, i.e.,

motor driven auxiliary feedwater pumps inoperable. Westinghouse has e

l

_ _ _.

3-4 __

_.

.

,

l

. .

'

-

d7:crain d that thic is acespeablo for thasa two costs bcccusa of th2 l little or no decay heat present allowing sufficient time (# 30 J

!

minutes) for operating personnel te rack in the pump power supplies and regain steam generator level.

.

3.1. 9 f. S. 3. 8.1.1, 3. 8. 2.1, 3. 8. 2.3 POWER SOURCES These specifications are outside Westinghouse control, however it is l acceptable to alter power source availability as long as manual Safety

-' E-Injection is operable and safety related equipment will function when required.

3.1.10 T.S. 3.10.. SPECIAL TEST EXCEPTIONS - PHYSICS TESTS This specification allows the minimum temperature for criticality to be as low as 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 acceptability.

s.-

--

,

"

3.1.11 TICHNICAL SPECIFICATIONS NOT EICEPTED While not applicable at power levels below 5: RTP the following tech- l nical specification limits can be expected to be exceeded:

1. 3.2.2 MEAT FLUX HOT CHANNEL FACTOR - q F (z)

At low temperatures and flows Fq(Z) can be expected to be above

~

l normal for 5% RTP with RCPs running. However at such a low power level no significant deviations in burnup or Ie peaks are expected.

I

2. 3.2.3 NUCLEAR ENIHALPY HOT CHANNEL FACTCR - (FAa)

-

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

1

'

s 35

[ 05tw\

- -

. ,

I

'

,

3. 3.2.4 QUADRANT POWER TILT RATIO I

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. .

,

4. 3.2.5 DNB PARAMETERS In the performance of eeveral tests the plant will be depressurized

', '-

below 2230 psia. At icw operating power levels this depressuriza- - - -

tion is not significant as long as subcooling margin is maintained.

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

-

u 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.

.

es

'

1 I

3-6

._ __

_ _ _ _ _ _ _

_ _ _ _ _ _ _ _ _ _ _ _ , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

. .

l

. l

~3. 2 OPERATIONAL SAFETY CRITERIA During the performance of these tests the operator must meet the follow--

ing set of criteria for operation:

.

1. Maintain For All Tests a) Primary System Sub-cooling (T,,g Margin) > 20 F j b) Steam Generator Water Level > 30% Narrow Range Span c) Pressurizer Water Level (1) With RCPs running > 22% pan (2) Natural Circulation > Value when RCPs tripped d) Loop AT < 65*F -

e) T avg -

< 578 7 f) Core Exit Temperature (highest) < 610 F g) Power Range Neutron Flux Low Setpoint and Intermediate Range Neutron Flux Reactor Trip Setpoints < 7% RTP

h) Control Bank D 100 steps withdrawn or higher

'

i) T eold > 530 F

2. Raactor Trip and Test Termination must occur if any of the following condi-

~

tions are met: l l

a) Primary System Sub-cooling (T,,e Margin) < 15 7 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 :

concurrent with a T,yg

-

change s

l 3-7

.

676&G

'

l 1

.'

' '

'

o) Any Loop AT 8

> 65 7 l f) T avg > 578 7 I

g) Core Exit Temperature (highest) > 610 F W h) Uncontrolled rod motion i) Control Bank D less chan 100 steps withdrawn j j) T < 530 F eold

3. Saf ety Injection must be manually initiated if any of the following condi _ ,

tions are met:

a) Primary System Sub-cooling (T,,e Margin) < 10 F b) Steam Generator Water Level < 0% Narrow Range Span or Equivalent Wide Range Le- el 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

-

,

change.

'e) Pressu;'.zer Pressure Decreases by 200 psi or mory in an unplanned or unexplained manner.

_

l

,

9 3-8 6761A

__

.

S3fcty Injcetion must n:t ba tarminctsd until the Wsseinghousa 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.,

1. Sufficient margin to saturation temperature at system pressure to assure adequate core cooling (no boiling in the hot channel), G ..

. - -

2. suf ficient steam generator level to assure an adequace secondary side heat sink;

3. sisfficient level in the pressurizer to assure coverage of the .

hearers 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.

.

%

<w l

.

J 2

. .

.

-

TABLE 3-1 TECHNICAL SPECIFICATION Ud. PACT Test Technical Specification 1 22 2b 2c 3a 3b 4

-

2.1.1 Core safety Limits 1 I I I X c4

' .~~~.'

2.2.1 Various Reactor Trips Overtemperature AT I I I I I I Overpower AT I I I I I I Steam Generator Level I I I I I I 3.1.1.4 Moderator Temperature Coef- I ficient 3.1.1.5 Minimum Temperature for I Criticality 3.3.1 Various Reactor Trips Overtemperature AT I I I I I I

_. .

Overpower AT I I I I I it

' '

Stema Generator Level I I I I I I ,

3.3.2 Safety Injectiot - All I I I I I I automatic functions l Auxiliary Feedwater Initiation I' I I I I I

- 3.4.4 Pressurizer I I I 3.7.1.2 Auxiliary Feedwater I I 3.8.1.1 AC Power Sources I I 3.8.2.1 AC Onsite Power Distribu- I I tion Systen 3.8.2.3 DC Distribution System I I 3.10.1 Special Test Exceptions - I Physics Tests

-

&

.

3-10

_

. .

.

\

, \ 4.0 SAFETY EVALUATION

. i. ,

+

In this section the saf'aty effects of those special test conditions

c

<

which are outside the bounds of conditions assumed in the FSAR are evaluated. The interaction of thes'e conditions with the transient analyses in the FSAR are discussed.

-

I, 4.1 EVALUATION OF TRANSIENTS .; _

g.

The effect of the unusual operating conditions en the transients analyzed in the FSAR are evaluated.

4.1.1 CONDITION II - FAULTS OF MODERATE FREQUENCY 4.1.1.1 Uncontrolled Rd Cluster Control Assembiv Bank Withdrawal from a Suberitical Condition Restriction of control rod operation to manual control, and constant 6- saperator monitoring of rod position, nuclear power and temperatures greatly reduces the likelihood of an urcentrolled RCCA withdrawal. ,.

,

Operation without reactor coolant pumps, and in some 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 vsactor trip at least one reactor coolant pump will be restarted and the reactor baron concentratien will 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 Section 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 Assembiv Bank Withdrawal at Power

-

The same considerations discussed in Paragraph 4.1.1.1 apply here. In s

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

and Intermediate Range Neutron Fluz trip setpoints act to nitigate this l

4-1 i

' e

.

. . .

.

incident, while lack of the Overtemperatura AT crip removss s:me 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 Misalignmect The FSAR discussion concerning static RCCA misalignment applies to the test conditions. The consequences of a dropped RCCA would be a decrease b'

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 consequences of, and operator action time requirements for, an uncontrolled boren dilution under the test conditions are bounded by ,

1 j

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 Seepoint and the constant operator acnitoring of reactor power,

-- (esperature and charging system operation, provides added protection. ,

.

4.1.1.5 Partial Loss of Yorced Reactor Coolant Flow Because of the low power limits the consequences of loss of reactor coolant pump power are trivial; indeed they are bounded by normal opera-

- ting conditions for these tests.

4.1.1.6 Startup of an Inactive Reactor Coolant Loop When at least one reactor coolant pump is operating, the power limit for these tests results in such small 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 a :d thus provide a change in reactivity and pcwer. Because of the small modera-tor reactivity coefficient at beginning of life the power increase in

,

the worse condition would be small and gradual and the flow-to pover 4-2

.

w

. . .

.

ratio in the core would be increasing. The Power Range Neutron Flux Low 9

Setpoint reactor trip provides an upper bound on power. Because of the increase in flow-to power ratio and because of the low setpoint on the l reactor trip', DN3 is precluded in this transient. j

- '

, ,

4.1.1.7 Loss of External Load and/or Turbine Trip

_

- l Because of the low power level, the disturbance caused by any losa of 3 b 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 PSAR 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 Generrtor Water Level. Ample time is available to re-institute auxiliary feedwater sources. ,

-

e 4.1.1.9 Loss of Offsite Fower 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 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 main feedwater bypass valve flow or auxiliary feedvater flow, about 15% of normal flow. Tae transient is further mitigated by the low operating powcr level, small moderator temp trature reactivity coefficient, the low setpoints on the Intermediate and Power Range Neutron Flux Law setpoint trips, and close operator surveillance of fe<.d 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.

4-3

, _ _

-

. . .

'

-

4.1.1.11 Excessive Load Increase Incident o

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 temperature 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 ,_

2.. .

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. In addi.cion, 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 ad of hot leg sub-cooling, with specific initiation points for manual reactor trip, pro-

- erides protection against DNB in the event of an accidental depressuriza-tion of the RCS. In addition, automatic reactor trip caused b* the Low ,

Pressurizer Pressure Safety Injection signal would occur when core out-let subcooling reached approv4== rely 100 F'as an automatic backup for manual trip. During testa 2b and 2c, when this trip is bypassed to I

allow deliberata operation at low pressure, the pressurizer PORV block

- valves will be closed to remove the major credible source of rapid f iredvertent depressurization. (The Low Pressure trip is automatically reinstated when pressure goes above 2000 psig and the FORV block valves will be reopened at that time.) l I

,

4.1.1.13 Accidental Depressurization of the Main Steam System

.

i

)

The FSAR analysis for accidental steam system depressurization indicates 1 that if the transient starts at hot shutdown conditions with the worst l l

,

RCCA stuck out of the core, the negative reactivity introduced by Safety i 1

[ Injection prevents the core from going critical. Because of the small l k moderator temperature reactivity coefficient which will exist during the

.

4-4

- - _

-- .

-. .

.

. .

.

test period, the reactor would remain subcritical even if it were cooled

-

, . to room temperat:re without Safety Injection. Thus the SAR analysis is bounding.

4.1.1.14 Sourious Ooeration of the Safety Injection System at Power

._

In order to reduce the possibility of unnecessary thermal fatigue ,

s- ,"

'

"

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 ahnual action will be disabled. Thus, the most likely sources of spurious Safety Injection, i.e., spurious or '

" spike" pressure or pi' essure-difference signals from the primary or ,

secondary systems, have been eliminated.

4.1.2 CONDITION III - INFREQUENI FAULTS 4.1.2.l_ . Loss of Reactor Coolant from Small Ruptured Pipes er from

- y Cracks in Large Pipes Which Actuates ?.mergency 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 readi*y 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 ecoling 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 fi-cient time for the operstor 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- ,

1 sient du ".ng 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 :nd short operating history.

o s

- --

. ._ 4-9._

. _ _ - _

. . . ,

, .

.

4.1.2.2 Minor Secondarv System Pipe Breaks q

,

The consequences of minor secondary system pipe breaks are within the bounds discussed in Faragraph 4.2.3.

!

4.1.2.3, Single Rod Cluster Control Assembiv Withdrawal at Power

) ',

D The FSAE 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 ia 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 ope. ration, and hot leg temperature.

4

..__ 4,.l.2.4 other Infrequent Faults

'.. .

The consequences of an inadvertent loading of a fuel assembly into a

improper position, complete loss of forced reactor coolant flow, and waste gas decay tank rupture, as described in the FSA1, have been reviewed and found to bound the consequences of such events occurring

_

during test operation.

1 4.1.3 CONDITICN 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. During the large break event the system inventory and cold leg accumulators will have removed enough energy to bave filled the reactor vessel to the bottom of the nosales. Following the system depressurization there is enough

_ _ _ 4_6

. - .

.

,

.

water in the reactor vessel below the nozzles 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 teria 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 .* c..

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 power, with specific initiation criteria for raaetor trip; low trip setpoints on the Intermediate-Range and Power-R.mp Neutron Flux trips;

'

Low Flow Mismatch setpoint fcr Reactor Trip and 4tSIV 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 system. Following reactor trip, assiuming 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 discuseion.

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

%

.

4-7

- .

. . .

'

o

'

transients, and the operator actions required to deal with this situa- "

tion during this passe 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..

tube rupture accident mitigation would consist of those operator actions required to isolate the faulted steam generator, cooldown the RCS, and depressurize the RCS to equilibrate primary RCS pressure with the 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.

s

-

'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 abovs systems are i==diately available for operator control from the control room. Therefore, it is concluded that the ability to miti-

~ '

gate the steam generator tube .upture event is not compromised by the sodifications requirad 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

_ _ _ _

\

- .

.

l

.

l

. 4.1.3.5 Fuel Handling Accid uts :

d The FSAR analysis of fuel handling accidents is bounding.

4.1.3.6 Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembiv Eiection)

,-

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

-

.ut -

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 gradur.i. and terminated 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 res e of combining these ane-liorating effects with the effect of the nat il circulation flow race on clad-to-water heat transf er and RCS press. i have not been analyzed.

The extremely low probability of an RCCA ejec. on during ths brief

__. peqiod in the test sequence does not warrant such an analysis.

.

4. 2 ANALYSIS OF TRANSIINES 4.2.1 ANALYSIS OF RCCA 3ANK WITHDRAWAL FROM SU3 CRITICAL CONDITION

_ An acalysis 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 neminal.

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 530 7.

a 4-9 6761A 1

-

l i

. . . !

i

-

.

j l

The lower bound for total delayed neutron fraction for the beginning

-

4.

of life for cycle I 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. .- g. ,

,

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 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 with either all scesaline isolation valves open or two of those valves -

closed. A range of reactivity insertion races up to the maximum for two banks moving was assumed for cases with all steanlines 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

-

coeffi:ients for beginning of life, Cycle 1, were investigated. In all cases, reactor trip was initiated at IC: 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 races, the minimum reactivity coefficient cases give 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 esses evsn the slowest insartion 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 races go on for so long that the core

.

4-10 l 1

6761A -

l

. .

.

that the core inlet temperature finally increases before trip, i.e. , c after approximately one and one-half minutes of continuous withdrawal.

Thus, the cases shown bound the worst cases.

4.2.3 ANALYSIS OF C00LDOWN TRANSIENTS Cooldown transients include feedvater system malfunctions, excessive

' ~

stema 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 cooldowns before reactor trip would occur. (Follow-ing reactor trip, no cooldown event would return the reactor to t criti-cal condition.)

During natural circulation operation, approximately one to two minutes would elapse following a secondary side event before cold water from the steam generator reached the core; thus, considering the close and con-stant sur veillance 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 singla power-operated steam pressure relief valve, a dump valve, or a safety valve, would cause an iceresse of less than four percent in reac-tar 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 I reactor, isolate feedvater 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 BOL Cycie one moderator reactivity coefficient and 5% initial power are shown in 1

'

Figures 4.2.5 and 4.2.6. i

!

l l

4-11 i

. _ _ _ _ ._ _

.

. ' . . l

- i

.

High Fluz Protection !

4.2.3.2 W

Reactor trip ors 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. Upper-bound negative moderator isott.ermal temperature coefficient, -

~

1-vs. core average temperature, for beginning of life, Cycle 1.

2. Lower-bound fuel temperature power reactivity coefficient.

3. Initial operation with core inlet temperature 555cF.

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

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

This results in some boiling in the reactor. The negative reactiv-

-

v icy introduced by core boiling would effectively

limit power; this -

negative reactivity was conservatively neglected.

6. Uniform core inlet. temperature and flow.

_

7. Reactor trip equivalent to 10% of full power at the initial inlet temperature. The power as measured by the NIS is assumed to be diminished 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 assumed equal to the predicted flow under steady-state operating conditions.

.

9

-

1 l

4-12

- - .

. .

' .

.

Analyses of core conditions based on these assumptions indicate that the

-

DN3 criterion of the FSAR is met.

4.2.3.3 Secondary Pressure Trio 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 .- g, 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 isusediately. In the example shown, the main steamline isolation valve on loop cie was assumed to fail to close. No power excursion resulted, and the reactor remained suberitical af ter the trip.

4.3 ADDITIONAL CONSIDERATIONS

-

S. .

'

In the great ma ority of cases it was concluded, either by reanalysis or

by comparison with previously analyzed FSAR conditions, that fuel clad integrity would be maintained without ceed for operator mitigating ac tion. For the LOCA or steambreak events, it was concluded that the operator would have more than ample time (> 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ) 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 I 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 cs l

l 4-13

i

. . .

.

\i probability of an inadvertent rod withdrawal a

event during the limited

"

duration of the special tests.

n 1

Thissectionprovidesanadditionaljassessment \

relative to the potential for and consequences cf fuel failure for ch'ese "unanalyzed" accident ;

\ \

.

conditions associated with certain rod withdrawal events. This assess- '

sent is partially based upon an attempt to bound certain effects which ,

l may exist for conditions removed from the range of direct model appli- c g. j I

cability. Additional information (attached) is provided for four areas:

1. Thermal rargin associated with normal test conditions.

l

2. The potential for DNB during accident conditions. ;

3. The clad temperature response assuming that DNB occurs.

l

4. Radiological consequences associated with presumed gross fuel failure.

..

y

.

The catclusions of this assessment are as follows:

1 i

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

l

!

2. Even assuming DNB, there should be adequate heat ciansfer to prevent clad overheating. 1

!

3. Fuel clad failure is not expected.

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

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

conditions by establishing a lower bound requirement on the degree of s

4-14

___

_- . _____ _. _ - _ - _ _ _

. . .

, .

.

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

4.3.2 DNB CONSIDERATION 3 .

,.

For certain cooldown transients, the conclusion that DNB is precluded was drava based on t W the W-3 critical heat flux correlation.

Although the analyses for the cooldown events discussed in section # 1-4.2.3.2 result in masa velocity below the rangi dir-c.: applicability of the correlation, the reactor heat flux was eo 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 traasients the li:niting conditicus, 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 r correlations (references 1 and . ) iodicate that it will conservatively -

predict critical heat flux at low pressure (r 1000 psi) conditions with low coolant flow. kool boiling critica'. 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 nuclear feedback from voiding. The power distribution would tend to peak towards the bottom thus further reducing the local quality at the .

peak flux locations.

,

i l

l

'

4-15

'

i

. . . I

.

l

'

-

Also the pool boiling correlations in reference 3 show some decrease in e critical heat flux above 1000 psia to the maximum pressure of appli-cability of 2000 psia. However extrapolatica of the correlations to a value of zero critical heat flux at the critical pressure (3206.2 psia) vculd 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 en the order of .02 x 106 3TU/br-f t2 a large peaking f actor would be

E-required to put the reactor heat fit._ ... high as the critical heat flux.

For the transients considered, the only ones that lead to significant off normal peaking factors are rod motico transients. The rod with-drawal from suberitical is a power burst concern. As cuch, it is

'

expected that even if DdB occurred, the rod surf ace would revet. For the rod bank withdrawal, the combination of :uximum 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 preclude

_

DNB, although a prudent judgement indicates that, it is indeed remote.

4.3.3 CLAD M ERATURE CONSIDERATIONS Should DNB occur, the peak clad temperature reached would depend prima-rily on the local nuclear transient following DN3 and on the behavior of

_

the post-DNB heat transfer coefficient. l For a rapid power transient, as is illustrated by :he SER analysis for RCCA bank withdrawal from a subcritical condition, the fuel temperature reactivity feedback and reactor trip on a nuclear flux signal would shut down the reacter before sufficient energy could be gener :ed to cause a damaging rise in clad temper.ature. In that case, the maximum clad tem-parature calculated was under 13000F even assuming an extremely low heat transfer coefficient (# 2 STU/hr-f t2 _oy), j I

1 A possibly more limiting condition for RCCA withdrawal would be the esse 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 4-16

. . .

)

.

. l

.

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

In this state the clad temperature could be calculated given only the total core power, local heat flux channel factor, heat transfer coeffi-1 cient and saturation temperature.

The core power is postulated to be essentially at the power which would i cause a reactor trip on high Power Range Neutron Fluz low setpoint. The trip setpoint is at 7% for these tests. To allow for calorimetric .- g.

errors and normal 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 approwimmtely 1450 psis 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 I

transfer coefficient, and the product of heat flux bot channel factor (Fq) times core power (fraction of RTP). For the event of an uncon-

~

Irelled RCCA bank or single RCCA the upper bound of this heat fl.ux

'

product is approximately 0.34. Using this value, the heat transfer coefficient required to keep the peak clad temperature below 18000y, 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 tre.usfer coefficient for post-DNB conditions. i 1

Although no correlations were found which cover the complete range of conditions being tested, s,me data exist which can be extrapolated to obtain representative heat transfer coefficients. The Westinghouse UHI film boiling correlation (reference 4), was developed at low flow ecadi-tions similar 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 BTU /hr-fg2.oy)a,e at 2200 psia and 5% flow with quality

.

,

4-17

,

. -

-, -

. .

,

. -

,

between 10-50%. Other film boiling heat transfer correlations, devel- J oped at higher pressures, were also examined. These correlations were extrapolated down to the lower flow conditions of the PSE&G tests as another approach to obtaic representative film boiling coefficients.

Using both the Mattson et al (reference 5) and the Tong (reference 6) fila boiling correlations resulted in post-DN3 heat transfer coeffi-cients in excess of 150 BTU /hr-ft2 - 7 4e the conditions given above.

.- 1 These results indicate that a clad temperature excursion resulting in fuel damage is not likely to occur even if UNB is assumed.

4.3.4 DOSE ANALYSIS CONSIDE3ATIONS The dose analyses were performed for a hypothetical accident senario using conservative assumptiens so as to determine an extreme upper bound on pos tulated accident consequences. The analysis assumed a reactor accident involving no pipe-break with a coincident loss ot condenser vacuum. This accident scenario is representative of the Condition II

-- ,

type events analyzed in the FSAR. The bounding were assumpticus made in -

the analysis which include:

133 Mwt (5% power) 1.0 dose-equivalent I-131 RCS activity (tech spec 1 Lait) 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 500 iodine spike factor over steady state 509,000 lb. at spheric steam denp over 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1

~3 3 1.7 x 10 sec/m z/Q percentile value The results of the analysis show that the two hour site boundary doses would be 5 ren thyroid, 0.9 rea total body and 0.4 ren to the skin. l l

4 4-18 i 6761A t

L ~

.

j I

i l

. .

.

. .

i The analysis of the accidents has incorporated some very conservative 1 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:

l

1) Equilibrium radionuclide inventories established at 5% power. For

'~

lodines, this requires # 1 month of steady state operation at 5% ;

_unin t erruo t ed. .- i... '

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. ,

l

'

l l

3) 100% fuel rod clad damage.

4) Primat, .o 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. .

O e

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

I For these reasons, in the unlikely event of a potential accident during l

-

the tests, the resulting dose is small, even assuming 100% clad damage and other extreme conservatisms. l l

l 4.3.5 OTHIR CONCZRNS l l

l' 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 te, cake corrective action. Some transients were not analyzed or discussed in this enpplement Jue to the combination of the low probability of the transient or. curring and the very short time d

l 4-19 -

6761A a

'

. .

*

. .

period of the special tests. This is true for the rod ejection acci- e dent. The combination of the low probability of occurring and the l bounding dose evaluation for a condition II transient given here indi-cate that these events do not need to be analyzed. Si:nilar 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. y ,, .

-

e 9

-

4

.

%

~

t 4-20

I

. . . .

. .

1

.

REFERENCES

<

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

1 l

- Stanek, " Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water," Sys'posium on Two-Phase Flow in Rod Bundlee, Code

$

R27, ASME Winter Annual Mee;,ing, November, 1969.

, -

' '--* -

2. Eao, B. R. , Zielke, L. A., Parker, M. B. , " Low Flow Critical Heat Flux," ANS 22, 1975.

.

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

4. WCAP-8582-P, Vol. II, "31owdown Experiments With Upper Head Injac-tion in C2 17x17 Rod Array," McIntyre, 3. A., August, 1976. (West-inghouse Proprietary)

._

5. Mattson, E. 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 5th Int. Heat Transfer Conference, Tokyo, September (1974).

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

_

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

.

1

.

s 4-21 )

' ]

. - . i

>

y .

-

\

H

.

1.0

.

-

0.80 Te .

7 E'

b- 0.60 -

a-Ki -

um

. . . . . . .. _.

=c

- *

, . .

.

9, o 0.40 -

,

a a S .

e so 0.20 -

.

_

, , , ,; k , , . , ,

-

0.0 10.0 20.0 30.0 40.0 Time (sec.)

Figure 4.2.1 Uncontrolled Red Bank Withdrawal from a Suberitical Condition, Neutron Flux vs. Time

.

l

.

-- _ __

4.22

-.

1

. .. . 1 I

.. , ,

\

"I t,

;

'

:

1300.0 ' :

-.

1200.0 - -

.

Q ..

--

1100.0 - -

w

--

e g 1000.00 - -

a.

-

- -

d 900.00 - -

~

c --

W S00.00 - -

..

.= oo

--

E 700.00 - - -

a .

V .

--

.'~ ' S00.00 - -

--

l 500.00 - ..

'

~ ' ' '

' '

400.00

'

o e o e o e o e o e o o e o o o o e o o e o o e . . . .

a . .

o

.

o

.

o e o o

~ - o e m c tn <,0 n c:a O ~ N TIME (SEC)

.

Figure 4.2.2 Uncontrolled Rod Bank 'dithdrawal from a Subcritical Condition, Hot Spot Clad Temperature vs Time, Assuming CNB at Time = 0.

6 4.23 1

.

%

-- , _.

O -

8

-

l I l

's i I

I I a

-

,

.

x

= c., u

,

- = u _=

_ =

.

,,

__

-

,-

=_ QJ C

-

n = :.

.

- _

,.' Q $

_ -- -

'

[ _.

-- .

w ag g p.

_g,

_

-

. .

_. -

--- ,

e i

- z% 2. c.

W -,

O C

/ . -

_a.

-

= - - - m. . .

,

.

4 gm

.=

= = = 1

/ . 6 a z. z i e

.

4 w e,,

W

.

-

._

. . .

A

3=

t *1 * /

r' 1 M i . < Cb

!

1

_

o ca 4J. . 6 / e s4 e

,

e

.

- , .

=- 3

- -- , t , i,

q) .4 g- et 3: * /

,

6 >.. 1 A t I t a

6 .

-)

g #3 i l' I i I '4 1 e.

'

5: Mi 5: -

iA i f

-

- -

=

-

=s W 3#

!

-

m es 6

= o

,o

-

- .h.

- .

,

._ ._ _ = >

3 ca m

- - = z

-

_

- > e

_. _ _ _ _ - as

- + =<

. .

v =

<

..-

.

, ,

ta c=

=?

-= e C ca

-

.

=s TD

-

'

CJ 8

- _ _ . - .y

-

e _

-

- o.D we

%

p

-

- a_ e

-

=e vs c

, =-

.k

.

i .

.

. . . . . -

.

. . .

. .

, , .

. . . ,

, , . . i .

.

f I e e l 4 1

I t $ $ l a . I e I *

. . i m

, i e i . i . i i i . ,

,

. e i ! - m e e , m m - e .I.

m - e e e s

- - - a 6

_-

=

.

(W 099210 ",) xn u u sy saog gesd (ad5 00S88 Jo",)Mou CE

.(J.) aan;uadmai (ty:.pu{ SACqy GS!B SJn1MGCWB1 SS(U{

s

)

i l

- . . _ - _ _ . , _ _ m

. -_ _ _ _ _ _ _ _ - _ _ _ _ _- _ _ _ . _ _ _ _ _ - . __ . -- .. - -. . . . . . - _ _ . - - -- ._- -

,

. . . . .

.

>

< .

,

O

'

1000

.

. . . . .

. _ . ,

,,--.--:f=_==-

_-

-

_

,_e. m- - , - --. m f

-

,

-. - _. _

,

)

i .-

1

.- e_ ..

._

s M

x N

s 1

.

s ' ' ' f I t =\ I i . i

. . . 4 .

. .. , m . . . . ,

" '

, .

'

.

' ' ' ' '

' ' ' N ' ' ' '

100 ; ,

L

. - -

\, .

. . .

, e . . . . . ~ .

e .1 z

C

! O '

w m

,

t

- -

'

x

>=

w .

e -

.

u g - .,

% T N1

. .

r 1 . . , e s 6- , , ,.

-

. , . ~ .. .w, , , ,

. .

' '"

_

10

-_

' '- ' ' '

m \. ,

s s N

. . . . .

. . .

. . s. . .

.

m

- .

!

_

1 i

'

i ,

. , . .

, . .

. .. . .. ,. . , . . ... . ,. . .. .. . . ...

,

. . , i

. .. i. , , , , , , . i. 4 3 ) '

100 1 10

[ REACTIVITY INSERTION ? ATE (PCM/SEC)

Figure 4.2.4 Uncontrolled Red Sank Withdrawal at Power.

Time of Reactor Trip vs Reactivity Insertion j Rata

_ _ _

-- . _ , . _ . . _ _ _ _ . _ _ _ _ _ _

. .- .

-

.10000 --

.. .

.

= <

m z .08000 - -

4 .

cy --

,

-

o .05000 - - A

< z wwo .l.

-

.04000 - - ..

w y ..

Eg

'~

v

.02000 - - r

- .. 1 w

-

0. 0

-

I_

~

> .00050 - -

- ..

-~

5

.

.00020 -..-

~ ..

u

.00010 - - ~

-

E00 .00010 - -

,

"

'

-; ;

600.00 * -

--

,

.. Core Avg w

e

--

9 o e C 500.00 - -

u w e c

--

..

_,

z- .

us " --

All Leops gM3 400.00 - -

Cold Leg

<

=

w <

--

>.

< w a.

..

1

)

1 yW 300.00 - -

--

o u

~~

. ===

..

.

' ' '

200.00 .

'

.15000

'

m : <

o a -a --

- * --

All Loops v -

z gg 10000 -

--

l t, = .. -

m o a v .05000 - -

> < = ..

-

~

l o 5 ' l

' 00 o o e o o o o o o o o o o o o o s d d d C e 5 o

5 o

d C. o o o o ec:3 o o - eu m e o m ~ 2 TIME (SEO 7IGURE 4.2.5 TRANSIrcs ni THE REAC OR CORE AND CCOLOC LCOPS

-

F@LLOWDIG THE OPENDIG OF A STEAM DUMP 7AL7E FRO'

. ..

.

.i 2400 0 m --

e .. .

. -

e --

2 2200.0 - -

- -.

2

<

-.

..

i ?

4f

- ~

2000 0 - -

,

A i

= .. .

3 i.

e --

e w

1800.0 - - \.

.

-

a C

i2 .- ..

-'

.

,

--

r .

- -

C --

>- 1500 0 - -

-

- ..

-.

.

- - --

50 1000 00 - -

-.

. m ...

w = --

m 500.CO --

_~.

= t:t ..

..

2 e ' '

o.o .

E

a. .

,

!200 0 ;

-

s.

--

!000 00 - -

,

Al !. I. cops -

,

.u.

4

--

800 00 - -

a ws

-

= .

--

5, sco Cc - -

w A c=

%

1 --

5

-

400 00 - -

w9

.

--

200 00 - -

' '

0. 0 C e C C e C ci C C C C C C C C C

. . . .

O C C C C C C C C

- O C C C C C C s

o - nu m e o us e, C.

TIME (SEC)

FIGURE 4.2.6 TRANSIDTS IN THE PRESSURI2ZR MTD STIAM GTERAIOR FOLLOWING THE OPENUG OF A STEAM DUMP VALVI 710M ma o mwe ~

_ _

-

l

.t .

l

.

f* .

.

,

.100C0

,,

a t

w <

s - x .08000 - -

.. q

.- 2

--

"

l

-

3 06000 - -

.. )

w - ~

r

-

o . 0',000 - - -- ,

w a .

-

o v s .02000 -

..

-

..

_

-

,

0. 0

.-

2-

> '..

..

-

C .01000 -.. - --

> ..

=

w . 02C00 - ..

-

5

=

.03000 f ..

--

.04000 600.00

..

"

o Core Avg

" Loop 3 o C o e 500.00 - - \ s _

"

w w --

--

e o ..

Icop 2

-

.3 m -

-

w w =00.00 - -

Loop 1

" 3-

= < "

w> ew

..

< a.

1 -

ew o

300.00 - -

_ v ..

..

200.00 L

=

.15000 l c - -

\ g $ ..

x

v -

Loop 1 -

-

I .10000 - -

Loop 2 -

c cz

~

-- T -

5a

'

L g .0$000 - -

= -- Loop 3 l

&

e :

.

,

0.0 e e e e e e =

--

l ,

o e c e e o c e c o o a

d d d o' e

e' w

d. m a

, . ~

l l TIME (SEC)

FIGURE 4.2.7 T?ANSIICS Dr THE

-

REAC"'OR CORE

,-t, ei s ,me ANDnr CCOT.XC a va nLOOPS emyLIN E _

,,.r.

,,

..

2400.0 -- c

%. . --

e ..

g 2200.0 - -

--

a. --

, <

f 2000.0 - -

x s-

=

..

~

T

..

U 1800.0 - -

_

ea.

y . . . _z.

-.

3, o

_

e ..

> 1500.0 - -

--

..

".

.

y " 10CO. 00 - -

,,

aw w e

--

..

~ m 500.00

-

E S! --

o ..

W

0. 0 -

-

a Q.

'

120040

. _ . . . _ . . .

. . _- .

,..

.. e

"

1000.00 - -

_

< ..

G 600.00 - -

%

.

- w

a. --

y 600.00 - -

Leop 2 Loop 3 x w

=

\

Ch.

z

..

" 400.00 - -

w Leop 1

..

I 200.00 - -

-

1 0.0 e e g g 8 8 8 8 e = o. o .

8 y g i ci c, $ n

.

TIME (SEC)

IGURE 4.2.8 TRANSIDrIS IN IEE PRESSURIZER AND SFAM GEERATOR

,oacrd6Ff_A %v\556 >6: OfL?II

r

.

, -

6 se ' -

s. ,- -

o

,!

" _

%

~

5 1

'

0 _

t 0 a 5 _

' g _

eF _

) l*

0

% t2

'

2

/ . l o

' ) l -

1 1 yg

-

3 0 1 bn i

' 6 i

' D.

I

. & dl 5

' 2 1 eo S

. ( 3 t o N i c O p '_ ~

mb i u '

I T

b LS I D

'

/ N 4

F

O 0 C

'

4 .

/ 5 e T r S

'

/ u t

E T

84

/ '

a r

e N

0

' p 1 m 1 e A 0 T L

2 U 5 g C

' . e t i

o L l C

N

-

d

t

I L

- .

s "o A

' e C R

.

T 0 v

'e U T

A

- 0 i N

- 5 t

' c 1, A 3

.

' 4

- E R

U 0 G

. I 0 F

_

4

-

'-

-

0 6

4

'

s

-

- -

0 I

0 0 0 0 0 0 0 0 0 0 0 0 0 K 0 0 0 0 0 0 0 0 0 I S 4 3 2

- 0 0 0 0 9 0 7 6 5 4 3 2 1 1 1 1 1 I 1 1 1 2 2 2 2 2 2

7 ,

e r

u

-

.

s

-

s 9 1,

S e

- C r s R P P

.

e

. .

.

"

' . '

_ -

7

-

- 0 n

9 6

'

0 D

A L

' C R

'

PO F 5

' X 0 U L

F T

' A E

n l I

o .

a 4 i

t S F i V

s ' c 0 p 0 a 0 r l l

8 0 F l 1 0 E*

6 r I CF

= 1 '

e

- w I*

F0 e o F3 r P E8 u 3 x O1 t

a C r

'

0 q F

? e F lO l

p E m '

FF e SD l f T '

ii t

u T d . l A a l0 l r C Tn

' 2 Au k te i

-

a 0 l r e

P 2

f .

3 4

1 E t

. i

'

0 u G

I F

'

- 0

-

0 0 0 0 0 4 2 0 0 8 6 2 0 1 1

.

,

r e

f t t

.

s n F n e

_

a i F

_

r c "r T i l f l t f /

-

a e u e o t

.

I l 1l Il

ML19331D948

Contents

Text

1.0 INTRODUCTION

SUMMARY

2.0 DESCRIPTION

Navigation menu

ML19331D948

Text

1.0 INTRODUCTION

SUMMARY

2.0 DESCRIPTION

Navigation menu

Search