ML20100Q805

Forwards Draft Version of Comanche Peak Steam Electric Station,Unit 2 FSAR Update for Optimized Fuel Assembly & Accident Analyses Methodologies for Review.Draft Will Be Incorporated in FSAR Amend 85

ML20100Q805
Person / Time
Site:	Comanche Peak
Issue date:	03/17/1992
From:	William Cahill TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TXX-92119, NUDOCS 9203230175
Download: ML20100Q805 (120)

Text

k s

M O" E

"""***"f",",." Log # TXX-92119

. .: File # 10010

~~ ~

915 Ref. # 10CFR50.34(b)

TUELECTRIC William J. Cahl!!, Jr.

March 17 1992 t

U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)

DOCKET NO. Ed 446 ADVANCE FSAR SUBMITTAL - UNIT 2 UPDATE FOR OPTIMIZEP FUEL ASSEMBLY AND UNIT 2 ACCIDEt4T ANALYSES METHODOLOGIES Gentlemen:

Amendment 84 to the CPSES FSAR was transmitted in TV Electric letter TXX-92082, dated February 28, 1992. As stated in TYX-92082, a portion of the Unit 2 update was not available in time for incorporation intG Amendment 84. The remaining portions of the Unit 2 update are attached. The material in Amendment 84 and I this letter should provide the balance of information required for the review of Unit 2.

For large and Small Break LOCA, separate Figures which show the Unit 2 plant parameters have been added. The current FSAR Figures for Large and Small Break LOCA are being relabeled from " Units 1 and 2' to " Uni. 1,* but have not been provided as part of the attachment. Only the figures which cre relevant to the RUnit 2 submittal are attached.

As the NRC Staff is aware, a license amendment has been requested concerning the

, operation of the Flux Doubling Actuation System durin2 a Joron Oilution Event in Operating Modes 3, 4, and 5. Se',eral Unit 2 parameters and results for this event remain labeled as "[TBD)* in the attachment pending resolution of the license amendment request.

To facilitate NRC Staff review of the Unit 2 update, t e attachment is organized

i. as follows:

A marked-up copy of the revised FSAR pages (changes are indicated in 1.

L the margin by the word " draft").

2. A description / justification of each change.

n b

9203230175 92O317 PDR ADOCK 05000446 doo R oiive stree La.si Dalias. Tex.s 752oi

% PDR i

p.

-TXX-92119

~Page 2 of 2; The material attached to this letter will be incorporated in FSAR Amendment 85,

which is-currently scheduled for'May,-1992. If you have any questions regarding-this_ submittal, please contact David Bize at (214) 812-8979.

Sincerely, William J. Cahill, Jr.

By:

J. 5. Marshall Generic Licensing Maneger Attachment DNB/dnb c - Mr. R.~D. Martin, Region-IV Resident Inspectors, CPSES (2) -

Mr. H. B. Fields (NRR) p i

+

y a

N e

F

- . - - - , ,,e -%,, - ,, ,r , ;..-., , -.  %% -

'N CPSES/FSAR A-iecent-measurement in the second cycle of a 121 assembly, 12 foot,

core is compared with a simplified one dimensional core average axial

. calculation in figure 4.3-25. This calculation does not give explicit representation to the fuel grids.

The accumulated data on power distributions in actual operation is basirally of three types:

1. Much of the data is obtained in steady state operation at constant power in the normal operating configuration; ,

2. Data with unusual values of axial offset are obtained as part of 84 the excore detector calibration exercise which is performed every 92 EFPD:

i

3. Special tests have been performed in load follow and other transient xenon conditions which have yielded useful information on power distributions.

These data are presented in detail in Reference [8], Figure 4,3-26 contains a summary of measured values of FO as a function of axial offset for five plants from that report.

4.3.<. 2,8 Testing and Operations Support DRAFT

A very extensive series of physics tests is performed on the first cores. These tests ind ti criteria for satisf actory results are 1 l-described in Chapter 14. Since not all limiting situations can be

. created at BOL, the main purpose of the tests is to provide a check on

-the-calculational methods used in the predictions for the-conditions of the test. Limited tests are performed at the beginning of each DRAFT reload cycle to verify that the core is loaded as designed and can be safely operated.- The methodology is described in References 35 and 36 are-employed to predict core characteristics -required for physics testing and reactor operations.

4.3-33 Amendment 84 February 28 1992

\

. CPSES/FSAR

28. _Hoore, J._S.,_

" Nuclear Design of Westinghouse Pressurized Water- t Reactors with Burnable Poison Rods," WCAP-'7806, December 1971.

29. Nodvik, R. J., "Saxton Core 11 Fuel Performance Evaluation,"

WCAP-3385-56, Part II, " Evaluation of Mass Spectrometric and ,

Radio-chemical Analyses of Irradiated Saxton Plutonium fuel,"

July 1970.

30. Leamer, R. D., gi al., "PLO-2-UO2 Fueled Critical Experiments,"

WCAP-3726-1, July 1967.

31. Vassallo, D. B. "Interium Safety E,: wition Report on Westinghouse fuel Rod Bowing,* U5NRL, April 1976. l

32. Camden T. M.,Het, c'., "PALADON - Westinghouse Nodal Computer- 84 i Code," WCAP-9485-PA, December 1978.

33. Ankey, R. D., "PALDON - Westinghouse Nodal Computer Code " WCAP- 84  :

9485-PA, Supplement 1 September 1981. ,

. 34. Skaritka, " Westinghouse Wet Annular Burnable Absorber Evaluation Report." WCAP-10021, Rev. 1, October 1982. 84

35. Edwards, D. J., " Control Rod Worth Analysis. 'XE-90-005, DRAFT December, 1990.

36. Edwardt, D..J., et al . " Steady State Reacto'r Physics DRAFT Methodology," RXE-89-003-A, July, 1989.

9 k

Amendment 84 _4.3-74 3 February 28, 1992

CPSES/FSAR-TABLE 4.3-2B- :l.84 el (Sheet 1 of 3), j

' NUCLEAR DESIGN' PARAMETERS: j (First Cycle - Unit 2) i e Core Averace Linear Power. kW/ft. includino ,

.densification effects '5,45 Total Heat Flux Hot Channel Factor. Fg 2. 32 j Nuclear Enthalohv Rise Hot Channel Factor. F pg 1,55 Reactivity Coefficients + Deslan Limits .Best Estimale Doppler-only Power. Lower Curve -19.4'to -12.6 -14 to -10 Coeffic*ents, pcm/% Power ++ ,

(See Figure 15.0-2). Upper Curve -10.2 to -6.0 -12.0 to -8.0

• Mppler Temperature. Coef ficient. -2.9 to -0.91 -1.9 to -1.3 pcm/0F++.

Moderator Temperature Coefficient. +5 to -40 0 to -36.4 pcm/op++

Rodded Moderator Density, pcm/gm/cc++ 10.43 x 105 10.28 x 105 Bovon Coefficient. pcm/ ppm ++ -16 to -7 -16 to -10 'I Boron Coefficient for Boron Dilution.

pcm/ ppm ++

.[

Modes.1 and 2 -13.3 -12.7 ,;

l DRAFT Hode 5 -14.0. -13.7

+ Uncertainties are given in Section 4.3.3.3

+& Note: 1 pcm' .(percent mille) 5.as where no is calculated from two statepoint values of keff by In (K 2/K1 )- .

Amendment 84-February 28, 1992 s

'k

^

CPSES/FSAR-TABLE 4.3-20 #

-l-84 (Sheet 2).

NUCLEAR DESIGN PARAMETERS (First Cycle - Unit 2)

Delayed Neutron Fraction and Lifetime '

peff BOL, (EOL) -0.0075, (0.0044)

/*. BCL, (EOL) . p sec 20.7 (21.3) .l DRAFT ,

Control Rods Rod Requirements See Table 4.3-3 Maximum Bank Worth, pcm < 2000 Maximum Ejected Rod Worth See Section 8.2.2 l Bank Worth +++, pcm++ BOL. HZP. Xe frqg EOL HZP. Equilibrium.xE i .

Bank D 770 690 ,

Bank C 1150 1170 Bank B 1180 1190 Bank A 360 420 Radial Factor (BOL to E0L). ,

Unrodded 1.a2 to 1.32 0 Bank 1 56 to 1.52 D+C 1.58 to 1 54

+++These are typical values for Ag-In-Cd or Hf and will vary somewhat (about 7% higher) with hybrid B 4C r absorber design.

Draft Version

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

A'

, CPSES/FSAR For. transients which may be ONB limited the radial peaking factor is of:importance. The radial peaking factor increases with decreasing power level due to-rod insertion. This increase in Fag is included in the core limits illustrated in Figure 15.0-1. All transients that-may be DNB limited are assumed to begin with a 3F g consistent with  !

the initial power level defined in the Technical Specifications.

_ DRAFT The axial power shapes used in the DNB calculations are discussed in Section 4.4.

The radial and axial power distributions described above are input to the THINC Code as described in Section 4.4.

For transients which nay be overpower limited the total peaking factor ,

(Fq ) is cf importance, All transients that may be overpot' imited are assumed to begin with plant conditions including power distributions which are consistent with reactor operation as defined in the Technical Specifications.

For overpower transients which are slow with respect to the fuel-rod

- thermal time. constant, for example tne Chemical and Volume Control System malfunction that results in a decrease in the boron concentration in the reactor coolant incident which lasts many minutes, and the excessive increase in secondary steam flow incident which may reach equilibrium without causing a reactor trip, the fuel rod thermal evaluations are performed as discussed in Section 4.4.

For overpower transients which are fast with respect to the fuel rod thermal time constant, for example the uncontrolled rod cluster

l. control assembly bank withdrawal from_subcritical or low power startup I

l and rod cluster control assembly ejection incidents which result in a L large power rise over a few seconds, a detailed fuel heat transfer calculation must be performed. Although the fuel rod thermal time i constant is a function-of system conditions, fuel burnup and rod l power, a typical value at beginning-of-life for high power rods is approximately 5 seconds.

Draft Version 15.0-12 i

l lL

.- ._. . . _ ~ _

1. - CPSE3/FSAR TAB 12 15.0-2 4 (theet 3) ,

Sthecut.Y OF INITIAL CGCIT10N3 AND C6 700ES U$ED f i-Reactivity Coefficients issumed Initial NESS ,

Thezzel Power Output bkaderator Modesator As summedh- Irpy- 34-computer, Teranre censity . Wait 2 j'se

- Fanit s Codes Utilize <j (pem/*T1 (A kla/cel P vpier 94*t) only!_ l 84 cm plete loss of forced MFTRM, FACTDAN, 4 5 (Unit 2) 0.0 (Unit 1) upper

• 3425 Yee l . 84 }!

reactor coolant flow THINC .[

l84 t

i Rosetor coolant pep shaf t MFTPAN, FACTEAN +$ - upper

• 342$ '

16 l 84

, seirere (locked rotor) 1

[

15.4 Reactivity and Power .l V

' Distribution Ancunaltes 4 i i

Uncontrolled rod cluster TWDrKLE, FACTPAN, Refer to See- Defect 14 Wait 1) G *

. f 84 control essembly bank TRIWJ ' tion 15.4 1.2 Defect 0.424 (Unit 2) l 84  ;

wit.birawal fra a i t

i ,

6

, suocratical or low power a startup condition ,

Uncontrolled rod cinater LCFTFAN +5 +.41 lower and 7425 Yes lDRAr control ass ==bly back wr* 2082

, withdrawal at pcwer 403 i

TtntTI.E, MFORAN, Sectien 13.4.3 305 Yee Foi cluster control -

Sectica 15. 4.3

]84- -

assembly misalignment [

1 Chenacal and Volume Control NA M1 WA ha N4 No 64 systeen malfunction tha*

results in a decrease in the- (

boron concentration in the '

' reactor coolant -

[

t t

, _m=- c - y , w , y y ._.-_ - e.- t _+, - _ _ _ _ _ , _ _

e f.:

CPSES/FSAR TABLE 15.6-2 tsh*et $)

StMGLRT CF INITIAL CONCITIMf3 AND CCN71TR CTES USED Remotivity Coef ficiente Assumed Initial NSSS Thermal Power Output Moderator M rator Assumed @- ITCP' f 84 C e ter Temperature Density Mnit 2 f84 Foalt s Codes tMilized ' f oes/CE dkige/ed C e ler Owe l only) l 84 Lose of coolu t accidents 1Asct BREAK 'See sectima -

see section 34118 No lDFAIT ,

resulting frte the spectrim ' (UNITS 1 a 2) 15.6.5, 1546.S, j DPAfT of postulated piping breaks 3ATAN-VI, references references .

within the reactor samlant YJ.EFIOCD , COCO, f76

- pressure boundary LCCTA-IV 'I 1

.. i 2MLLL BREAK (UtfIT 1) f DFAFT ' Ii MTLASE, b6 .IV fDSAFT t

SPALL EREAK (UNIT 2) lDFAFT

>m , IoCTA.I. } D,m a

f 76 CThis power outpet'does not include the thenmal power generated by the remotor onolent pumps fee. Table 15.6-5), '

3 i

I i

i i

NOTE 1 " UPPER CURVE" MOST NEGATIVE DOPPLER ONLY POWER DEFECT = -1.6 Ap (0 TO 100% POWER)

NOTE 2 " LOWER CURVE' LEAST NEGATIVE DOPPLER ONLY POWER DEFECT = -0.78' Ap (0 TO 100% POWER)

-20 -

-18

, -16 -

-14 -

5 12 - >

a NOTE 2

-10 -

g llE c 8 -

6 -6 -

e

-4 -

-2 -

~

0 '

0 20 40 60 80 100

% POWER COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT .

., UNIT 2 Doppler Power Coefficient Used

, in the Accident Analysis flG W E 15.0-2B

CPSES/FSAR

?* TABLE 15.1-3 )

(Sheet 3 of 4) -

d. Initial steam. release 164,000(Unit 1) _0 RAFT from affected steam 176,000(Unit 2) ORAFT ,

generator (1b)(0to30 min) 66 l

e. Long term steam release 0 (Units 1 and 2) ORAFT from affected steam 66 generator (1b) (0 to 8 hr) 66

f. Steam release from three 66 unaffected steam generators 66 (1b) (0 to 2 hr) 438,000 (Unit 1) ORAFT 413,000 (Unit 2) DRAFT (2 to 8 hr) 860,000 (Unit 1) DRAFT 873.000 (Unit 2) ORAFT

3. Dispersion data

a. EAB and LP7 distances 1544m and 4 miles 66

b. x/Q 0 EAB 2.6 x 10-4 sec/m3 66 (0 - 2 hr) 66 0 LPZ (0 - 8 hr) 66 2.3 x 10-5sec/m3 66 l

4. Dose data L

_a. Method of dose calculations See Appendix 158 66 l b. Dose conversion assumptions See Appendix 15B 66-L l

l l

l __

Draft Version

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

CPSES/FSAR

^

The following conditions were assumed for an inadvertent boron 52 dilution while in these modes:

1) The boron concentration required to meet a SDM of 1.6% AK/K 84 (Unit 1) and [TBD]AK/K (Unit 2)-is~very conservatively estimated to be 1465 (Unit 1) and (TBD] (Unit 2) ppm. This corresponds to >

a critical CB of-1350 ppm (Unit'1) and (TBD] ppe (Unit 2),

assuming a very conservative, constant boron worth of 13.9

pcm/ ppm (Unit 1) and [TBD] pcm/ ppm (Unit 2).

L

2) Dilution flow rate is limited by design to a maximum of 167 gpm. 74 i

3) A minimum RCS water volume of 4169 (Unit 1) and {TBD] (Unit 2) 84 ft 3. This is a conservative estimate of the active volunie of the RCS while on one train of RHR, and is a very conservative 14 estimate of the active RCS volume with one reactor coolant pump operating.

Q212.78 'F Q212.136 Dilution Durina Startup 14

- Startup is a transitory mode of operation. in this mode the plant is 14 being taken from one long term mode of operation, Hot Standby, to

! another, Power. The plant is maintained in the Startup mode only for l the purpose of startup testing at the beginning of each cycle.

l During this mode of operation the-plant is in manual control, i.e.,

Tavg/ rod control is in manual. All normal actions required to change power level, either up or down, require operator initiation.

The Technical Specifications require a SDN of 1.6% A K/K (Unit 1) and DRAFT 1.3% A K/K (Unit 2) and four reactor coolant pumps operating. Other conditions assumed are: 14 Q212.78

-0212.136 l- 1) Dilution flow rate is limited by the design of the CVCS and 74 l RMWS. The makeup flow rate is limited to a maximum of 167 gpm for startup.

15.4-33 Draft Version l

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

CPSES/fSAR

2) A minimum RCS water volume of 9000 (Unit.1) and 9100 (Unit 2) DRAFT ft3 . This is a very conservative estimate of the active RCS volume, minus the pressurizer volumet 14

3) Initial CB for criticality is assumed to be-1600_ ppm (Unit 1) - DRAFT and 1500 ppm (Unit 2) with a very conservative, constant boron  ;

worth of 12.5 pcm/ ppm (Unit 1) and 13.3 pcm/ ppm (Unit 2). l l

Qjljation Durino full Power Operation 14 i

Q212.65 Q212.78 Q212.136 The plant may be operated at power two ways, automatic Tavg/ rod 15 control or under manual (operator) rod control. The Technical 84 >

Specifications require an available shutdown margin of 1.6% AK/K (Unit 1) and 1.3%a K/K (Unit 2) and-four reactor coolant pumps

, operating. With the-plant at power and the RCS at pressure, the ,

, dilution rate is limited by the capacity of the centrifugal charging .14 -

pumps (analysis is performed assuming two charging pumps are in operation even though normal operation is with one pump). Conditions assumed for this mode are:

1) Dilution flow rate is limited by the design of the CVCS und 7a RMWS. The makeup flow rate is limited to a maximum of 167

-gpm. When the pressurizer level control is in manual, the l- maximumdilutionflowrateis167gpmandwhenii) automatic l

pressurizer level control, the dilution is limited to the maximum j _ letdown flow rate.(approximately 125 gpm).

Q212.78 .

Q212.136 2)_ A minimum RCS water volume of 9000 (Unit 1) and 9100 (Unit 2) _ DRAFT ft . This is very conservative estimate of the active RCS 3

volume,-minus the -pressurizer volume. 14

3) Initial CB for criticality is assumed to be-1600 ppm (Unit, 1) DRAFT and.1500 ppm (Unit 2) with a very conservative, constant boron worth of 12.5 pcm/ ppm (Unit 1) and 13.3 pcm/ ppm (Unit 2).

~ Draft Version 15.4-34 e W 1 w m 7 g-b-+- g r*- ?T+rwt'-t- it 4 T- T- T-71---

CPSES/FSAR the Intermediate Range. Too fast a power escalation (due to an 15

unknown dilutic.1) would result in reaching P-6 unexpectedly, leaving 14 insufficient time to manually block the Source Range reactor trip.

Failure to perform this manual action results in a reactor trip and immediate shutdown of the reactor.

! Q212.78 Q212.137 However, in the event of an unplanned approach to criticality or 14 dilution during power escalation while in the Startup mode, the plant

~

status is such that minimal imptct will result. 'ihe plant will slowly escalate in power to a reactor trip on the Power Range Neutron Flux - High, low setpoint (nominally 25% RTP). After reactor trip DRAFT there is et least 21.5 (Unit 1) and 17.9 (Unit 2) minutes for operator action prior to return to criticality. The required operator action is the opening of valves 1,2-LCV-1120 and E to initirle boration and 14 the closing of valves 1,2-LCV-112B and C to terminate dilution.

Q212.78 Q212.136 '

Dilution Durino full Power Opfration 14 With the reactor under manual rod :ontrol and no operator action taken 14 to terminate the transient, the power and temperature rise will cause the reactor to reach the Overtemperature N-16 trip setpoint resulting in a reactor trip, After reactor trip there is at least 21.5 (Unit DRAFT

1) and 17,9 (Unit 2) minutes for operator action prior to return to criticality. The required operator action is the opening of valves 1,2-LCV-1120 and E and the closing of valves 1,2-LCV-112B and C. The 14 boron dilution transient in this case is essentially equivalent to an uncontrolled rod withdrawal at power. A reactor trip occurs when 74 either the Hi Neutron Flux or the Overtemperature N-16 setpoint is reached. The maximum reactivity insertion rate for a boron dilution DRAFT transient is conservatively estimated to be 1.2 pcm/sec (Unit 1) and 1.2 pcm/sec (Unit 2) and is within the range of insertion rates analyzed for uncontrolled rod withdrawal at power.

15.4-37 Draft Version

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

CPSES/FSAR It should be noted that prior to reaching the Overtemperature N-16 14 reactor trip the o'perator will have received an alarm on Overtemperature N-16 and an Overtemperature N-16 turbine runback.

With the-reactor in automatic' rod control the pressurizer level controller will limit the dilution flow rate to the maximum letdown rate, approximately 125 gpm. If a dilution rate in excess of the letdown rate is present,_ the pressurizer level controller will throttle charging flow down to match the letdown rate.

Q212.78 Q212.136 -

Thus with the resctor in automatic rod control, a boron dilution will 14 result in a power and temperature increase such that the rod controller will attempt to compensate by slow insertion of the control  ;

rods. This action._by the controller will result in at least three l alarms to the operator: l

1) rod insertion limit - low level alarm, 14

2) rod insertion limit - low-low level alarm if insertion continued 14 after (1) above, ano

3) axial flux difference alarm ( AI outside of the target band). 14 ,

s Given the many alarms, indications, and the inherent slow process of 14 dilution at power, the operator has sufficient time _for action. For DRAFT example, the operator has at least 28 (Unit 1) and 23 (Unit 2) minutes y from the rod insertion limit low-low alarm until 1.6% AK/K (Unit 1) and 1.3% AK/K (Unit 2) is inserted at beginning-of-life. The time -

would be significantly longer at end-of-life, due to the low initial 14 boron concentration,_when shutdown margin is a concern.

Q212.66 The above results demonstrate that in all modes of operation an 14 inadvertent boron-dilution is precluded, or responded to by automatic functions, _or sufficient time is available for operator action to--

terminate the transient. Following termination of the dilution-flow and initiation of-boration, the reactor is in a stable condition with the operator regaining the required shutdown margin.

Draft Version 15.4-38

CPSES/FSAR.-

~

' YABLE :15.4-1 i - (',heet .4 )

l TIME SE00ENCE OF EVENTS FOR INCIDENTS WHICH CAUSE' REACTIVITY

! AND POWER DISTRIBUTION AN0H/l.IES Accident- -Event Time (seconds)  ;

1 i Unit 1 Unit 2 84 4

2. Dilution during hot shutdown-~anC hot 74~

standbv Dilution begins 0 74 .

Flux doubli.ng occurs 416 _84 ,

l and source of dilution 74 is automatically isolated 74 l Minimum margin to loss of 74 l shutdown occurs ~ 683 84 l

3. Dilution-during full power operation i

a. Automatic reactor control Operator receives low-low 0 0 ' DRAFT  :

rod insertion alarm due 74 i to dilution 74 Shutdown margin lost 1728 1434 l DRAFT

b. Manual Dilution begins 0 0 l DRAFT-control ,

t Draft Version 1

}

CPSES/FSAR TABLE 15.4-1 _,

(Sheet 5) l TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH CAUSE REACTIVITY AND POWER DISTRIBUTION AN0MALIES Eyc_n1 Time (seconds)

Accident I) nit 1 Unit 2 l84 Reactor trip setpoint reached for Overtemperature N-16 60 68 l DRAFT Rods begin to fall into 62 68.5 DRAFT 5

core Shutdown margin is lost (if dilution continues 1342 afte" trip) 1350 l DRAFT Rod cluster control l

assembly ejection l

accident 1

1. Beginning-of-life. Initiation of rod 0.0 0.0 l84 f full power ejection Power range high neutron 0.05 0.05 l84 flux setpoint reached Peak nuclear power occurs 0.13 0.14 l84 Draft Version

. CPSES/FSAR 80 break transient is much less severe than an equivalent sized break in the cold leg. Thus, the transient due to the inadvertent opening of a pressurizer safety valve would be much less severe than the small break cold leg transients discussed in Section 15.6.5. Initially the event results in a rapidly decreasing RCS pressure until this pressure reaches a value corresponding to the hot leg saturation pressure. At this time, the pressure decrease is slowed considerably. The pressure continues to decrease throughout the 84 transient. The effect of the pressure decrease would be to decrease or increase the neutron flux depending on the moderator density feedback, but the Reactor Control System (if in the automatic mode) f functions to maintain the power and averaga coolant temperature essentially constant until the reactor trip occurs, initially, the pressurizer level increases due to expansion caused bv the depressurization, and then decreases following the reactor trip.

The reactor may be tripped by the following Reactor Protection System signals:

5 1. Overtemperature N-16, 5 2. Pressu-izer low pressure.

An inadvertent opening of a pressurizer safety valve is classified as an American Nuclear Society (ANS) Condition II event, a fault of -

moderate frequency. See Section 15.0.1 for a discussion of Condition 11 events.

15.6.1.2 Analysis af Effects and Consecuences Method of Analysis The acc^idental depressurization transient is analyzed by employing the detailed digital computer code LOFTRAN {1]. The code simulates the neutron kinetics, RCS, pressurizer, pressurizer relief and safety ORAFT VERSION 15.6-2 _______ _ ___ _______ _____ _ - ___ . _ _ _ -

l 1

Y CPSES/FSAR I

valves,; pressurizer spray, stean generator, and steam generator safety _,

valves. The code computes pertinent plant variables including temperatures. pressures, and power level. 3 Plant characteristics and initial conditions are discussed in Section 15.0.3. -For Unit 2, this transient is analyzed with the Improved 84 Thermal Design Procedure as described in WCAP-8567[27). In order to j  ;

give conservative results in calculating the departure from nucleate boiling ratio (DNBR) during the transient, the following assumptions -)

are made:

t

1. Initial conditions of maximum core power, maximum reactor coolant 84 average temperature and minimum reactor coolant pressure, plus ,

uncertainties are assumed, resulting in the minimum initial margin to DNB (see Section 15.0.3). For Unit 2, uncertainties in initial conditions are included in the limit DNBR as described in WCAP-8567[27).

2. The most positive moderator temperature coefficient of reactivity 84 is assumed in order to provide a conservatively high amount of positive reactivity feedback due to changes in the moderator density. The spatial effect of void due to local or subcooled boiling is not considered in the analysis with respect to >

reactivity feedbeck or core power shape. These voids would tend 84 to flatten the core power distribution.

3._The least negative Doppler coefficient of reactivity is assumed 84 stch that the resultant amount of negative feedback is conservatively ~ low in order-to maximize any power increase due to moderator reactivity feedback, i Plant systems and equipment which are necessary to mitigate the effects of RCS depressurization caused-by an inadvertent safety valve opening are discussed in Section 15.0.5 and listed in Table 15.0-6.

i 15.6-3 DRAFT

. - - .: = - -.-..:=

CPSES/FSAR

' 70 6.' Single Active Failure 78 The failurt # the main feedwater regulating valve, in its as-is position, is the limiting single active failure with respect to the filling cf the ruptured steam generator. For this single '

failure scenario, 100% main feedwater flow to the ruptured steam  ;

generator is conservatively assumed to continue until the closure of the main feedwater isolation valve, and the elapsed  !

, time from reactor trip to automatic feedwater isolation is conservatively maxin.ized. For additional conservatism, no credit is taken for the early identification and isolation of the ruptured steam generator by the reactor operators.

t Resu'its Q212.66 81 Evaluation of the steam generator tube rupture event indicates that .

DNBR is not reached and, thus, no clad damage would be expected in this transient. This is consistent with the fact that when the '

reactor is at power, the reactor coolant pumps are operating and, for this event, only a small fraction of the total primary system fluid j .. inventory.has leaked to the seconaary side. Thus, it is very _;

unlikely that DNB would ',ccur as a result of the reduced RCS flow.

The RCS depressurization that results due to flow out of the tube ,

rupture presents another possibility for obtaining a low DNBR.

l However, the depressurization that occurs in a steam generator tube

( rupture is much less than considered in the depressurization transient analyzed in Section 15.6.1 for the Inadvertent Opening or a '

ORAFT Pressurizer Safety or itelief Valve. In the analysis of that event, it was determined that the DNBR remains above the limit yhlue throughout the transient and, thus, no clad damage is expected, c rom _,

81 this, it is concluded that no clad damage is expected in the steam generator tube rupture event.

78 Representative transient responses for the design-basis steam generator tube rupture event are shown in Fis"res 15.6-3 and 15.6-3A.

1 I

Draft Version 15.6-14

_ . _ . _ . . _ . . _ ~ . _ _ _ _ . _ _ . _ . __ _ _ _

. . CPSES/FSAR 76 described. Equalization of the primary and secondary system pressurss terminates the leakage flow into the ruptured steam generator in sufficient time to prevent filling the ruptured steam 78; _ generator. The volume available prior to filling the ruptured steam generator-is greater than for the scenario described in Section 15.6.3.2. ,

76 The time dependent mass releases used to assess the radiological cor.:equences of the postulated steam generator tube rupture are calculatea from the RETRAN02 thermal-hydraulic analysis described above. Tirredependent values of the leakage rate into the ruptured i lr, team generator and the flashing fraction were also used to essess the

.adiological _ consequences- for the 0-2 hour une period following the event. Following the closure of the atmospheric relief valve block .

valve, the additional radiological dose is due to the leakage from the o

primary systern into the intact steam generators and the initial concentration of radioactivity contained in the intact steam g

4 4 _ generators.

70 Two separate iodine spikes are considered:  ;

70 Case I A reactor transient has occurred prior to the tube rupture and raised the primary coolant iodine concentration to

72. 60 uCi/gm Dose Equivalent Iodina-131 (DEQ I-131). The resulting preaccident isotopic iodine concentrations are shown in Table 15.6-3.

.- 70 Case II The reactor trip or primary system depressurization

~ rissociated with the postulated accident. creates an iodine spike in the primary system. The spike is assumed to increase the iodine 4 appearance rate (inleakage from the defective fuel rods-to the.

primary coolant) to 500 times the equilibrium appearance rate.

The concurrent iodine spike appearance rates are presented in Table l

15.6-4.-

The assumptions below are used to determine the initial pr' mary and secondary activities and to calculate the activity released and the Draft Version 15.6-16

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

l*. ~)

i 1 CPSES/FSAR offsite' doses forL the postulated steam generator tube rupture - 70 accident.

,1 1 lhe initial primary coolcot iodine activity (i.e. prior to any 70 l iodine spike considerativns) is assumed to be at 1.0  :

uC1/gm DEQ'I-131, 70 g

4 2. The primary coolant activity has been leaking into-the secondary 70 side at one gpm for a period of time long enough to establish equilibrium activity concentrations is the steam generators. t

3. All noble gas activity transported from the primary syster to 76 l the secondar) system and all noble gas and iodine activity <

, initially in the steam region of the steam ge' arators is immediately released to the environment. The initial iodine activity in the water region of the ruptured steam generator increases over time due to the unflashed portion of the leakage.

4. Dee to the pressure differential between the primary and 76

-secondary sides, a fraction of the primary coolant that leaked

~

! to the defective steam generator flashes to steam. This flashed fraction does not mix with the-steam generatar water and, therefore, is not subjected to any iodine removal process in the steam generator. However, the flashed! fraction

l. experiences iodine removal in the condenser when that path is available.

L 1 .;

5. Radioactive decay of parent iodines to noble gas products is 76 considered during the_ iodine spiking processes and.as unfleshed iodine accumulates in the steam generators, the radioactive decay of the parent iodines is conservatively assumed to not decrease the activity of the parent iodines.

L p

15.6-17 Draft Version

m . :_. __-.__~_;

. CPSES/FSAR feedwater flow by closits the main feedwater isolation valves and also initiates emergency feedwater flow by starting the auxiliary feedwater pumps. The secondary flow aids in the reduction of RCS pressure.

When the RCS depressurizes to 600 psia, the accumulators begin to inject borateo water into the reactor coolant loops. Since the loss of offsite power is assumed, the reactor coolant pumps are assumed to trip at the inception of the accident. The effects of pump coastdown are included in the blowdown analysis.

DRAFT The blowdown phase of the transient ends when the RCS pressure (initially assumed at 2250 psia for Unit 1 and 2280 psia for Unit 2) falls to a value approaching that of the Containment atmosphere.

Pi ;or to or at the end of the blowdown, the mechanisms that are responsible for the bypassing of emergency core cooling water injected into the RCS are calculated not to be effective. At this time (called end-of-bypass) refill of the reactor vessel has filled the lower plenum of the reactor vessel which is bounded by the bottom of the fuel rods (called bottom of core recovery time).

The e 'lood phase of the transient is defined as the time period

' ; ting from the end-of-refill until the reactor vessel has been filled with water to the extent that the core temperature rise has been terminated. From the later stage of blowdown and toen the ~

beginning-of-reflood, the safety injection accumulator tanks rapidly discharge borated cooling water into the RCS, contributing to the filling of the reactor vessel downcomer. The downcomer sater elevation head provides the driving force required for the reflooding of the reactor core. The low head and high head safety injection pumps aid in the filling of the downcomer and subsequently supply water to maintain a full downcomer and complete the reflooding process.

Q212.40 Draft Version 15.6-22 '

__ - - - - _ . - . _ _- _______m__ _ . _ . _ _ _ _ _ . - _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . . - _ _ _ _ _ _ ________m_.______________.__.i:.__________.__ --___ ._a

CPSES/FSAR Continued nperation of the ECCS pumps supplies water during long term cooling. Core temperatures have been reduced to long term steady state. levels associated with-dissipation of residual heat generation.

After the' water level of the refueling water storage tank reaches a minimum allowable value, coolant for long term cooling of the core is obtained by switching to the cold leg recirculation p"ese of operation in which spilled borated water is drawn from the en Jr ?d safety [

features sumps by the low M9ad safety injection (resioual heat removal) pumps and returned to the RCS cold legs. The Containment 6 Spray System continues to operate to further reduce Containment pressure. Within approximately 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after initiation of the -78 LOCA, the ECCS is realigred to supply water to the RCS hot legs in order to control the boric acid concentration in the reactor vessel.

Descriotion-of small Break,LOCA Transient As contrasted with the large break, the blowdown phase of the small break occurs over a longer time period. Thus, for the small break LOCA there are only three characteristic stages, i.e., gradual blowdown in which the decrease in water level is checked, core recovery. and long term recirculation.

15.6.5.3 Care ani Sistem Performance 15.6.5.3.1 Mathematical Model The requirements of an acceptable ECCS evaluation model are presented in Appendix K of 10CFR50 [3).

i 115.6 DRAFT

, ~ . - .

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

.- CPSES/FSAR Larae Break LOCA Evaluation Model lhe analysis of a large break LOCA transient is divided into three phases: 1) blowdown, 2) refill, and 3) reflood. There are three distinct transients analyzed in each phase, including the thermal-hydraulic transient in the RCS, the pressure and temperature transient within the Containment, and the fuel and clad temperature transient of the hottest fuel rod in the core. Based on_these j considerations, a system of inter-related computer codes has been developed.for the analysis of the LOCA.

I The_ description of the_various aspects of the LOCA analysis I methodology is given in Reference [5]. This document describes the major phenomena modeled, the interfaces among the computer codes, and 1 the features of the codes which ensure compliance with the Acceptance Criteria. The SATAN-VI, WREFLOOD, COCO, and LOCTA-lY codes, which are used-in the LOCA analysis, are described in detail in References [6]

DRAFT through [9]. Modifications to the model are described in References.

[10), Ill), [12), [29] and (30]. These codes are.used to assess the 6 -core heat transfer geometry and to determine if the core remains amenable to cooling throughout and subsequent to the blowdown, refill, and reflood phases of the LOCA. The SATAN-VI computer code analyzes the thermal-hydraulic transient in the RCS during blowdown and the WREFLOOD computer code is used to calculate this transient during the refill and reflood phases of the accident. The C0C0 computer code is used to ';alculate the Containment pressure transient during all three

. phases of the LOCA analysis. Sinillarly, the LOCTA-IV computer code is used to compute =the-thermal transient of the hottest fuel rod during the three. phases.

l l'

l I

.Oraft Version 15.6-24

~

. CPSES/FSAR DRAFT The analysis presented here was performed for Unit 1 with the February 1978 version of the evaluation model which includes the modifications delineated in References [15], [16], [17] and (17a]. The analysis was performed for Unit 2 with the approved 1981 version of the evaluation model which includes modifications delineated in References (15], [16a), (17] and [17a).

6 The analysis in this section was performed with the upper head fluid temperature equal to the reactor coolant system cold leg fluid l temperature.

6 The upper head fluid tenperature has been made equal to the cold leg temperature by increasing the upper head cooling flow (20].

Small Break LOCA Evaluation Model DRAFT The WFLASH program used in the analysis of the small break LOCA for Unit 1 is an extension of the FLASH-4 code [13] developed at the Westinghouse Bettis Atomic Power Laboratory. The NOTRUMP cnmputer program was used ir. the analysis of the small break LOCA for Unit 2, which incorporates a number of advanced features. Among these new features are the utilization of nonequilibrium thermal ca'culation in all fluid volumes, flow regime-dependent drift flux calculations with counter-current flow limitations, mixture level tracking logic in -

multiple-stack fluid nodes and regime-dependent heat transfer ORAFT correlations, The NOTRUMP {31, 32] small break LOCA ECCS evaluation model was developed to determine the RCS response to design bas's small break LOCAs and to address the NRC concerns expressed in NUREG-0611, " Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plants."

The WFLASH and NOTRUMP programs permit a detailed spatial representation of the RCS.

The PCS is nodalized into volumes interconnected by flowpaths. The brcken loop is modeled explicitly with the intact loops lumped into a second loop. The transient behavior of the system is determined from l

l.

DRAFT VFRSION 15.6-26 _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ -

G CPSES/FSAR-r the governing conservation equations of mass, energy and momentum-applied throughout the system. Detailed descriptions of WFLASH and DRAFT

- NOTRUMP are given in References [14) and [31, 32, 33], respectively.

The use of WFLASH and NOTRUMP in the analysis involves, among other DRAF T things, the representation of the reactor core as a heated control volume with the . associated bubble rise model to permit a transient mixture height calculation. The multi-node capability of the program 4

enables an' explicit and detailed spatial representation of various system components. In particular it enables a proper calculation of the b'ehavior of the loop seal during a loss of coolant transient.

Clad thermal analyses arr performed with the LOCTA-IV code [9] which DRAFT uses the RCS pressure, fuel rod power. history, steam flow past the uncovered part of the core and mixture-height history from the WFLASH and NOTRUMP hydraulic calculations as input.

Q212.80 The small break analysis for Unit I was performed with the October, DRAFT-1975 version of the Westinghouse ECCS Evaluatton Model (refer to References [9), [14], [14a} and [14b]) and for Unit 2 with the

l. May,'1985 NOTRUMP ECCS Evaluation model (refer to References [33]). 6

• r Schematic representations of the computer code interfaces are given in DRAFT .

Figures 15.6-5 and 15.6-6 for Large Break LOCA and Small Break LOCA, l .. respectively.

l-15.6.5.3.2 Input Parameters and Initial Conditions 4

Table 16.6-$ lists important input parameters and initial conditions used in the analysis.

The analysis presented in this section was performed with a reactor 6

/ vessel upper head temperature equal to the RCS cold leg temperature.

The effect of using the cold leg temperature in the reactor vessel upper head is described in Reference [20]. In addition, the large 36 break analysis in this section utilized the upflow barrel-baffle l - methodology described in Reference [25].

, -15.6-27 DRAFT V . . .. -. . . . ._ - - .L. - . . - - - - . - - - - . --. - --

_.y 4

.. CPSES/FSAR i I

Q212.134 I

~ Q212.651 DRAFT The. bases used to select the numerical values that are input parameters to the analysis have been conservatively determined from sensitivity studies (refer to References [18), [19) and [20] for Unit 1-and References [18), (19), [20) and [20a] for Unit 2). In addition, the requirements of Appendix K regarding specific model

-features-were met by selecting models which provide a significant DRAFT. overall conservatism in the analysis. For the Unit 1 large and small t break LOCA base cases, the worst single failure assumed in the analyses is one RHR pump and one safety injection train, respectively.

For Unit 2 the worst large break LOCA (CD =0.6) is analyzed in j accordance with the methodology in Reference [28) and found to be most limiting when no single failure (maximum safegcards) is us'ed. The

- 72 assumptions made pertain to the conditions of the reactor and associated safety system equipment at the time that the LOCA occurs and include such items as the core peaking factors, the Containment pressure, and the performance of the ECCS. Decay heat generated through the transient is also conservatively calculated.

15.6.5.3.3 Results 78 The_results of-the base case LOCA analyses and the peak cladding temperature (PCT) penalties associated with subsequent safety Eevaluations are-described in this section.

Larae Break R f sults -

ORAFT Based on the results of the LOCA. sensitivity studies (References [18],

[19] and [20) for Unit 1 and References [18], [19), [20] and (20a]

for-Unit 2), the limiting large break'was found to be-the double. ended cold leg ~gaillotine-(DECLG). Therefore, only the DECLG break is

considered in the large brea6 Ev'; < erformance analysis. Calculations were performed for a range of Mocar m eak discharge coefficients. The result

of these calculations are summarized in Tables 15.6-1 and p i5.6-6.

P

, Draft Version 15.6-28

. CPSES/FSAR The mass and energy release data for the break resulting in the highest calculated peak clad temperature are presented in Section 6.2.1.5.

For Unit 1 figures 15.6-7 through 15.6-23 and 15.6-47A present the DRAFT transients for the principle parameters. The following items are nrted: 6 Figures 15.6-7A - 15.6-90: Quality, nass velocity, and clad 6 heat transfer coefficient for the hotspot and burst locations _

Figures 15.6-10A - 15.6-12D: Cnre pressure, break flow, and core 6 pressure orop. The break flow is the sum of the flowrates from both ends of the guillotine break. The 72 core pressure drop is taken as the pressure just before the core inlet minus the pressure just beyond the core outlet Figures 15.6-13A - 15.6-15D: Clad temperature, fluid 6 temperature, and core flow. "e -

clad and fluid temperatures are for the hot spot and buist locations Figures 15.6-16A - 15.6-17D: Downcomer and core water level 6 during reflood, and flooding rate figures 15.6-1BA - 15.6-190: Emergency core cooling system l6 flowrates, for both accumulator snd pur w e .afety injection Figures 15.6-20' -

15.6-210: Conto ament pressure, and core: 6

-power transients 15.6-29 Draft Version

CPSES/FSAR 6 Figures 15.6 15.6-23: Break energy release during blowdown, and the containment wall condensing heat transfer coefficient for the worst break DRAFT Figures 15.6-47A large Break pumped safety injection flow rate 6 Section 6.2.1.5: Presents a discussion of the containment pressure transient resulting from a LOCA.

DRAFT for Unit 2, figures 15.6-49 through 15.6-75 present the parameters of principal interest from the large break ECCS analyses: For all cases analyzed, transients of the following parameters are presented:

ORAFT a. Hot spot clad temperature. (Figures 15.6-49, 15.6-49A, 15.6-64, 15,6-70)

DRAFT b. Coolant pressure in the reactor core. (Figures 15.5-50, 15.6-65,15.6-71)

ORAF1 c. Water level in the core and downtomer during reflood. (figures .

15.6-51,15.6-51A,15.6-66,15.6-72)

DRAFT d. Core reflooding rate. (Figures 15.6-52, 15.6-52A, 15.6-67, 15.6-73)

DRAFT e. Thermal power during blowdown. (Figures 15.6-53, 15.6-68, 15.6-74)

DRAFT I f. Containment Pressure (Figures 15.6-54, 15.6-54A, 15.6-69, 15.6-75)

The Containment pressure transient resulting from a LOCA is presented in Section 6.2.1.5. i 15.6-30

CPSE5/FSAR for the limiting break analyzed for Unit 2, the following additional DRAf1 g transient pirameters are presented:

a. Core flow during blowdown (inlet and outlet). (figure 15.6-55) DRAFT

b. Core heat transfer coefficients. (figure 15.6-56) DRAFT

c. Hot spot fluid temperature. (F igure 15.6-57) DRAfi k d. Mass released to Containment during blowdown. (figure 15.6-58) DRAFT L

e. Energy released to Containment during blowdown. (Figure 15.6- DRAFT 59)

f. fluid quality in the hot assembly during blowdown. (figure DRATT 15.6-60)

g. Mass velocity during biowdown. (Figure 15.6-62) DRAFT

h. Accumulator water flow rate during blowdown. (Figure 15.6-61) i DRAFT

i. Pumped safety injection water flow rate during reflood. DRAff (figure 15.6-63) _

j. Large Break pumped safety injection flow rate for the worst DRAf]

single failure of one RHR pump (figure 15.6-47A)

The maximum clad temperature calculated for the large break base case DRAfl is 20110f for Unit 1 and 18080f for Unit 2 which are less than the 10CFR50.46 Acceptance Criteria limit of 22000F, The maximum local metal-weter reaction is 3.92 percent for Unit 1 and 2.04 percent for Unit 2, which are well below the embrittlement limit of 17 percent as ,

required by 10CFR50.46. The total core metal-water reaction is less than 0.3 percent for both units for all breaks, as compared with the 1 percent criterion of 10CFR50.46, and the clad temperature transient is terminated at a time when the core geometry 15.6-31

. CPSES/FSAR is still amenable to cooling. As a result, the core temperature will continue to drop and the ability to remove decay heat generated in the fuel for an extended period of time will be provided.

DRAFT A number of safety evaluations for the large break LOCA analysis for both units have been performed subsequent to the calculations reported above. For Unit 1, these safety evaluations were for reduced accumuletor water volume, thimble tube nodeling, reduced RHR flow due to miniflow, steam generator feedwater flow split modification,1%

uniform steam generator tube plugging and a 2.1% correction in steam generator flow area, steam generator tube seismic-LOCA assumption and resulted in a total peak cladding temperature (PCT) penalty of l 55.00F for Unit 1. The final limiting PCT for the Unit I large braak, including the penalties from these safety evaluations, is 2a65.70F.

DRAFT for Unit 2 these safety evaluations were due to rey'sion of thimble tube modeling, fuel ron modeling, steam generator tube seismic-LOCA assumption, modification in feedwater flow split, c reduction in the minimum SI flow and steam generator tube plugging (1%). Due to the

_ reduction in the minimum 51 flow, the minimum $1 flow case becomes more limiting compared to the base case. This is because the reduction in the 51 flow affects the PCT (18040F) associated with the minimum 51 flow case Lnly, whereas the maximum 51 case (18080F) remains unaffected. The above noted changes result in o peak clad temperature (PCT) penalty of 1110F for Unit 2. The limiting peak clad temperature (PCT) for the Unit 2 large break LOCA, including penalties, is 19150F.

DRAFT The PCTs for both Units are below the 10CFR50.46 Acceptance Criterion for peak clad temperature of 22000F.

1 Draft Version 15.6-32

l -

CPSES/fSAR 1m311 Break Resulti As noted previously, the calculated peak clad temperature resulting 6 from a small break LOCA is less than that calculated for a large break. Based on the results of the LOCA sensitivity studies (Reference [19]) the limiting small break was found to be less than a 10 inch diameter rupture of the RCS cold leg. A small break specturm DRAFT analysis showcd that the limiting small breaks were 4 inch and 3 inch for Unit 1 and Unit 2. respectively. The results of the analyses are summarized in Tables 15.6-1 and 15.6-7

, figures 15.6-34 through 15.6-46, 15.6-47B, and 15.6-48 present the DorrT -

principle parameters of interest for the Unit I small break ECCS analyses. For all cases analyzed the following transient parameters are presented: .

1. RCS pressure

2. Core mixture height

3. Hot spot clad temperature

4. Core power after rea'ctor trip for the limiting break analyzed for Unit 1, the following additional DRAFT transient parameters are presented:

1. Core steam ficw rate 2, Core heat transfer coefficient

3. Hot spot fluid temperature lhe maximum calculated peak clad temperature for the Unit I small DRAFT break base case is 17880F, These results are well below all Acceptance Criteria limits of 10CFR50,46 and in all cases are not limiting when compared to the results presented for Unit I laege breaks.

15.6-33 Draft Version

CPSES/fSAR l

78  ;

s I '* w Aar ed ' tty evaluations for the small break LOCA analysis have

)Rffi !e .ri wyfrA cad subsequent to the calculations reported above, for j 5 l 's 1, these safety evalestions were for reduced safety injection

> f '_., rates, an incrt'se in the auxiliary f eedwater line purge volume, a lower low pressurizer pressure safety injection signal setpoint, increased auxiliary feedwater flow and an increase in the signal processing delay time, correction of Zr-H 2O reaction error and resulted in a total PCT penalty of 2470f. The final limiting PCT for the small break, including penalties from these safety evaluations, is 20350F. This is less than the final limiting large 78 break PCT and remains well below the 22000F Acceptance Criterion value in 10CFR50.46.

DRAFT The results of the analyses for Unit 2 are summarized in Tables 15.6-1 and 15.6-7. The principle parameters of interest for Unit 2 small break ECCS analyses are presented in figures 15.6-47B and 15.6-76 through 15.6-89. The following is a list of the above parameters:

DRAFT 1, RCS pressure DRAFT 2. Core mixture height DRAFT 3. liot spot clad temperature DRAFT 4. Core power after reactor trip DRAFT for the limiting break analyzed for Unit 2, the following additional transient parameters are presented:

DRAFT 1. Core steam flow rate DRAFT 2. Core heat transfer coefficient DRAFT 3. hat spot fluid temperature Draft Version 15.6-34

. CPSES/FSAR l The maximum calculated peak cladding temperature for the Unit 2 small DRAFT break case is 14340F. These results are well below the Acceptance Criteria limits of 10CFR50.46 and in all cases are not limiting when compared to the results presented for Unit 2 large breaks.

15.6.5.4 Environmealal Consequences 66 70 demonstrate in a conservative manner that the operation of a nuclear power station does nct present any undue radiological hazard to the general public, a hypothetical accident involving a gross release of fission products is evaluated. No mechar. ism for such a release has been postulated because it would require a number of simaitaneous failures to occur in the engineered safety features. The 66 core fission product inventory is assumed to be released into the containment as described in TID-14844 (21]. Numerical values for the totat core fission product inventory of the isotopes considered in calculating the radiation doses are listed in Table 15.6-8.

The radiological evaluation of this accident is divided into two parts;: internal (thyroid) dose from inhalation of iodines in the leakage plume, and external (whole. body). exposure as a result of immertion in the leakage plume.

The rtdiological consequences due to the release of core fission 66 products during a postulated loss-of-coolant accident are evaluated in the following sections:

1. Radiological consequences of containment leakage 66 The integrated thyroid doses and the integrated whole body doses 66 art! calculated using methods and assumptions in conformance with Regulatory Guide 1.4. The assumptions used in the analysis are li sted below.  ;..

15.6-35 Draft Version

• i CPSES/fSAR

a. Twenty-five percent of the equilibrium radioactive iodine  ;

inventory developed from maximum full-power operation of the  !

core is immediately available for leakage from the reactor  ;

containment. Of this 25 percent, 91 percent is in the form of i elemental iodine, 5 percent is in the form of particulate  !

iodine, and 4 percent is in the form of organic iodides, f

b. All (i.e., 100 percent) of the equilibrium radioactive noble gas inventory developed from maximum full-power operation of the l

core is immediately available for leakage from the reactor containment.

t

c. The effects of radiological deca) during holdup in the -

containment are taken into account, 66 d. The containment volume is divided into separate regions by concrete floors at different elevations (see Section 6.5.2). A radial gap between the concrete floors and the inner wall of the Containment Building permits a limited amount of convective  :

-72 mixing-between these regions. The region not covered by f containment spray is treated as a separate unsprayed volume which is assumed to mix with the volumes in the sprayed areas at -

t mixing rate of two turnovers per hour.

The Containment Spray System is actuated by a high containment-pressure signal. For a discussion of the sequence of events of "

66 spray system operation, see Section 6.5.2. A sodium hydroxide  :

spray is used to reduce the amount of fii.sion product iodine available for release during the LOCA. ffhe containment spray solution is assumed to interact with the elemental iodine and particulate iodine. The mathematical model which calculates the iodine spray removal coefficient is presented in Section 65 6.5.2. For each region the calculated elemental iodine renoval  !

coefficients are above 10 hr-1 The removal coefficient for I elemental iodine used in the offsite dose calculation is limited ,

w e maximum value of 10 hr-1 [22). A conservative value of 1.07 hr-1 is used for the particulate iodine removal

'I Draft Version 15.6-36 -

i CPSES/fSAR

-coefficient, although higher removal coefficients have been 66  ;

calculated. The elemental iodine removal effectiveness may be expected to diminid after the concentration in the containment atmosphere has been reduced by several orders of magnitude. The 66  ;

elemental iodine removal effectiveness of the spray system is conservatively assumed to cease after a decontamination of 100 '

in the containment atmosphere has been achieved. -

i

e. The lodine and noble gases available for release to the 66 eny!ronment are assumed to leak from the Containment at a maximum leak rate of 0.10 percent of the containment volume per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 0.05 percent of the containment volume per day for the duration of the accident.

f. The duration of the accident is considered to be 30 days.  ;

g. A ground-level release is assumed. Atmospheric dilution 66 factors are discussed in Section 2.3 [23] and listed in Table I

i 15.6-9.

r

h. No credit is taken for depletion of fission products in the 72 plume due to ground deposition

• radioactive decay in transit. *

1. For the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident, the breathing rate of persons offsite is assumed to be 3.47 x 10-4 cubic meters per i second (m3 / sec). Eight to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the breathing rate is assumed to be 1.75 x 10-4 m3 /sec, l

from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> through 30 days after the accident, the rate is 66 ,

assumed to be 2.32 x_10-4 m3/sec.

66

j. The mathematical model and dose conversion factors presented in 66 Appendix 158 are used for evaluating the radiological consequences of the L.0CA 1

k. Other assumptions are detailed in Table 15.6-9. 66 l

15.6-37. Draft Version

CPSES/FSAR 66 Using the assumptions presented above and the mathematical models presented in Appendix 15B, the doses at the EAB were conservatively calculated to be 117.5 rem to the thyroid and 1.48 rem to the whole body; the doses at the LPZ were conservatively calculated to be 22.2 rem to the thyroid and 0.29 rem to the whole body.

66 2. Radiological consequences of engineered safety features equipment Q022.19g leakage outside containment.

66 Following a postulated LOCA, a potentis.1 source of fission product release is the leakage of water from engineered safety features (ESF) equipment located outside the containment. Such leakage could occur during the recirculation phase through components such as pump flanges, valves, and heat exchangers. The fission products could then be released from the water into the atmosphere resulting in of fsite radiological consequences that contribute to the total oose from the LOCA.

06 66 An analysis of the offsite effects attributable to ESF equipment ,

leakage is performed based on the following conservative assumptions:

66 a. 50 percent of the halogens originally present in the core are .

intimately mixed with the coolant water and are assumed to be available for release through ESF equipment outside containment (see Table 15.6-10).

66 b. All of the noble gases produced from the decay of halogens which remain in the leakage water are released to rooms housing the leaking components.

78 c. The leakage from all ESF components is conservatively assumed to start 10 minutes after the LOCA and continue for the duration of the accident at a rate of 2 gallons per minute.

Draft Version 15.6-38

__. _ . . _ _ _ . . . _ _ _ _ _ _ _ _ . - . _ _ _ . - _ __..___._____m 1

I CPSES/fSAR

d. An iodine partition factor of 0.1 is assumed. This factor is 66 taken as the fraction of iodine in the leakage that becomes airborne.

66

c. Gaseous radioactivity released to rooms housing the leaking 66 components is considered to be immediately swept away by the

. ventilation system and released to the atmosphere. (See Section 9.( 5 and Figures 1.2-16, 1.2-35 and 9.4-9). 53 i Q312.12 l

f. An iodine adsorber efficiency of 95 percent is applied since the 66 l ventilation exhaust passes through HEPA filters and iodine l adsorbers prior to release to the atmosphere. The iodine adsorbers are designed to the requirements of HRC Regulatory 59 Guide 1.52 (See Appendix-1A(B)) as discussed in Section 9.4.3.

g. No credit is taken for an elevated release; all meteorological 66 parameters are considered to be identical to those previously  ;

defined In this section.

(312412 ,

Based upon the foregoing model, the thyroid and whole body dose 66 contributions due to ESF equipment leakage are conservatively calculated to be 26.9 rem and 0.742 rem, respectively, for the EAB.

I The LPZ doses are conservatively calculated to be 14.1 rem to the thyroid and 0.817 rem to the whole body.

3. Total dose due to a postulated LOCA 66 i

The total dose attributed to a postulated LOCA is the combined 66

( doses due to containment leakage and ESF equipment leakage. The combined EAB doses are 145 rem to the thyroid and 2_2 rem to the whole body. The combined LPZ doses are 36.3 rem to the thyroid 72 and 1.1 rem to the whole body. As expected, the doses are below the values set forth in 10CfR100. l r

1 15.6-39 Draft Version

CPSES/ TSAR Q312.12 72 The dose to personnel engaging in mineral extraction operations within the exclusion area, in the event of a postulated LOCA, would be less than the dose values of 300 rem to the thyroid and 25 rem l to the whole body set forth in 10CFR100. )

66 ,

i

4. Dose to the control room occupants .

66 IntheeventofaDesignBasisAccident(DBA),thesafetyinjection actuation signal or a high radiation signal from the control room air intake monitors will initiate energency recirculation and pressurization of the Control Room air conditioning system.

46 Later,- the emergency ventilation air makeup system can be brought l J

into operation as described in Section 9.4.1.

66- The control room doses were analyzed for various design basis accidents. It was determined that the LOCA doses represent the ,

limiting case. Therefore the methodology and the doses calculated for the LOCA are reported here.

The following assumptions are applied in the calculations of the i

' dose to the control room occupants following the LOCA: "

66 a. The basic' assumptions presented in Items 1 and 2, above, are applied,-except a constant breatning rate of 3.47 x-10-4 m3/ sec is assumed throughout the accident.

l l

L -

L p

Draft Version 15.6-40

CPSES/FSAR

b. The control room pressurization (air intake) and recirculation 52 iodine adsorbers are assigned a 99 percent decontamination efficiency for both elemental and organic iodines in accorfance with Table 2 of Regulatory Guide 1.52 (See Appendix 1A(B)).

The pressurization adsorbers are arranged in series with the 59 recirculation adsorbers (see Figure 9.4-1) during the emergency pressurization mode, thus providing an equivalent decontaminating efficiency in excess of 94 percent for both elemental and organic iodines from the pressurization makeup air.

c. The control room air-conditioning system runs either in the emergency recirculation mode or the emergency ventilation mode during a LOCA. 66

d. During the emergency recirculation mode of operation, a constant 3

air intake flow rate of 800 ft / min is assumed. This makes up 72 for losses caused by leaks and maintains the control room atmosphere at a positive pressure of 0.125 inch water gauge relative to adjacent areas.

Since both recirculation trains are actuated by the safety 66 injection signal, the outside air intake flow rate during dual .

train operation is conservatively estimated to be 1600 cf m. If 78 both trains are assumed to operate for one hour, the calculated thyroid dose would decrease, due to the additional iodine filtration. The calculated whole body gamma and beta skin doses would increase slightly due to the additional intake of outside air. In both cases, the calculated doses remain below the limits specified in 10 CFR 50. Appendix A. " General Design Criteria for Nuc'iear Power Plants," Criterion 19. The reported control room doses are the higher values of the two cases.

15.6-41 Oraft Version

CPSES/FSAR 66 e. During the <.ergency ventilation mode of operation, 3800 ft 3/mia of outdoor air is used to introduce fresh air into the ,

control room.

f. The emergency ventilation mode of operation is under [

administra*'ve control so that the dose to the control room occupants is minimized. and the need for air change is  ;

satisfied.

The operating mode sequence used in this analysis is as follows:

l Air Filtered Operating Time in Intake Recirculation ,

Time Period Mode _ Hode flow Rd.g Elgg Ad.e 66 O to 96 emergency 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 800 7200 66 hours (l) recirculation ft /3 min ft /3 min 66 66 96 to 117 emergency 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> 3800 4200 t 66 hours ventilation ft /3 min c3/ min 66 117 to 720 emergency 603 hours0.00698 days <br />0.168 hours <br />9.970238e-4 weeks <br />2.294415e-4 months <br /> 800 7200  !

l66 hours recirculation ft / min-3 ft / min 3

66 (1): Since both recirculation trains are actuated by the St

[

signal, one train must be turned off manually by the operators within one (1) hour.

-g,-The distance from the Containment to the control room air intake is 94 feet, and the air intake is located 56 feet above 78- ground. The distance _from_the primary plant vent stack (i.e.

the.ESF leakage release point) to the closest air intake is- ,

138 ft.

I ,

ORAFT 15.6 42

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

i

)

,- CP5CS/ TSAR i i

h. Atmospheric dilution factors are determined from the following equation based on Reference [24): l.

X/0 - (U(x a yaz + A/(K+2)]'l 66 i

?

wheres. t t

U- wind speed at an elevation of 10 meters (m/sec) ,

i ay.az - standard deviation of the gas concentration in 49 i the horizontal and vertical crusswind directions, respectively, 46 both being evaluated at a distance of 94 feet for the containment leakage source and at a distance o+ 138 78 feet for the ESF leakage source K- 3 (S/D) 1.4 S- distance between containment surface or primary plant vent 78 ,

and closest control room lntake  ;

i f

D. diameter nf Containment for the containment leakage source -78 and a combination of portions of the Containment and Aux 111ary Building for the ESF leakage source A.- projected area of Containment Building (3265 m2) for the 78 7 containment leakage source and 2088-m2 for the ESF lenhage source -

I-Table 15.6-12 sLSmarizes the X/0 values calculated utilizing this expression.

i 15.5 DWT

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

4 CPSES/fSAR 78 l 1. The'tviel unfiltered infiltration rate in the cuntrol room is 12 cfm. including 10 cfm due to ingress / egress and 2 cfm leakage

, from the ductwork passing through the control room pressurc ,

i boundary. Leakage through the closed dampers due to the pressure differential is also included. The damper leakage air >

will be filtered by the recirculation filtration units.

66- J. Habitability of the control room is based on the following occupancy factors:  ;

i Time PerL24 Qtqvpynty Factor j O to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.0 e 1 to 4 days 0.6 I 4 to 30 days 0.4 66 k. The air volume in the control room used to determine exposures >

to operators is 423.032 ft3 [

66 1. The models for the major contributors to the control room dose f are provided in Appendix ISB.

78- Using the above assuretions and procedures, the thyroid dose is conservatively relculated to be 27.4 rem in the control room for the duration of the accident. The thyroid dose can be further 46 -reduced by the use of the fullf ace respirators which are available at all time'. in the-control room. Air packs are provided in the contrel ro',m and emergency control center for the use of operators leavtnu the control room either to go offsite, or to some control 66 point in the station, or to the control room from offsite. 'The use"of air [1cks reduces the thyroid dose considerably during such 78 movements. The total whole body gamma dose is conservatively calculated to=be 3.76 rem. This calculated dose includes whole 66 body dose contributions from containment sources (both direct and: ,

scattered i

l

[PAFT= 15.6 44

,._.u.,,- -

CPSES/FSAR

, radiation), the external passing cloud, control room atmosphere, 66 l activity buildup on filters, and streaming through doors and I penetrations. These calculated doses are less than _the limiting i values specified in 10CFR50, Appendix A. " General Design Criteria for Nuclear Power Plants," Criterion 19.

r The skin dose received in the control room during the accident 78 period is conservatively calculated to be 44.0 rem. This t

, calculated beta skin dose is less than the 75 rem limit allowed if 66 special protective clothing and eye protection are used.

Therefore, special protective clothing and eye protection are provided for use, if required, to reduce the beta skin doses to the operators to within acceptable limits in accordance with 10CFR50,  :

Appendix A. GDC 19.

5. Environmental consequences of containment purging to control 66 -

containment hydrogen concentration af ter a LOCA 4 Purging of the containment atmosphere provides a backup method for 66 controlling potential hydrogen accumulation in the Containment following a postulated LOCA. The use of the hydrogen purge system (see Section 6.2.5.2.2) is precluded by redundant electric hydrogen  !

recombiners located in the Containment Building (see Section  !

6.2.5.2.1). The electric hydrogen recombiners are the primary means of controlling post- LOCA hydrogen buildup. Thus, an I analysis of the radiological consequences of containment purging is not provided.

QO22.8

6. Environmental consequences of releases through the containn:ent 2 pressure relief line in the event of a LOCA QO22.8 An analysis of the radioactive effluents escaping the Containment 72 to the environment after a LOCA, via the line through the controlled access area exhautt system, was performed using the i following assumptions:

__ _... _ __~_ _

15.6-45 .h _

i i

CPSES/fSAR 0022.20 2 a. The maximum containment air / steam mass release to the environment assuming critical flow was calculhted as 5,427 lbm-  !

l

b. Only reactor coolant activity is assumed to be released and the largest release occurs for a 3 inch break. Hence, only the 3 i inch break is analyzed.  !

i i

66 c. A preaccident iodine spike was considered in determining the j primary reactor coolant activity. The corresponding reactor  ;

coolant iodine concentrations are listed in Table 15.6-3. The  :

noble gas activity concentrations are presented in Table 15.1-4  ;

Q312.21 5 d. The containment pressure relief line isolation valve closure l time including instrumentation delays wil: not exceed 5 seconds.

The radioactive fission products are assumed to be released -

from the Containment through the pressure relief line for a  ;

period of 38.1 seconds. This includes 33.1 seconds to initiate  ;

the low pressurizer pressure trip satpcints (see Table 15.6-1)  !

corresponding to the 3 inch line break conditions.

e. No credit was taken for radioactive decay.

f. No credit was taken for an elevated release. ,

A Based on the foregoing assumptions, the doses to the thyroid and whole body were conservatively calculated to be P.03 rem ano ,

72 1.61 x 10-3 rem, respectively, at the exclusion area boundary.  ;

The doses from this accident are well within the values set forth in 10CFR100.

L 15.6.6 A NUMBER OF BWR TRANSIENTS This section is not applicable to the CPSES.

l L Draft Version 15.6-46

CPSES/FSAR REFERENCES

1. Burnett, T. W. T. , et al. , "LOFTRAN Code Description,"

WCAP-7907-P-A(Proprietary),WCAP-7907-A(Non-Proprietary), April 78 1984

2. American National Standard Source Term Specification, N237, March

9. 1976, _ ,

3. " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10CFR$0.46 and Appendix K of 10CFR50. Federal Register, Volume 39, Number 3. January 4, 1974.

1. " Reactor Safety Ftudy - An Assessment of Accid 6nt Risks in U. S.

Commercial Nuclear Power Plants," WASH-1400, NUREG-75/014, October 1975.

5. Bordelon, F. M. , Massie, H. W. and Zordan, T. A. , " West inghouse LCCS Evaluation Model - Summary," WCAP-8339, July 1974.

6. Bordelon, F . M. , et al. , " SAT AN-!V Program: Comprehensive Space-Time Dependent Analysis of Loss of Coolant," WCAP-8302 (Prcprietary) and WRAP-8306 (Non-Proprietary), June 1974

7. Kelly, R. D. , et al., " Calculational Model for Core Reflooding After a Lost of Coolant Accident (WREFLOOD Code)," WCAP-8170 (Proprietary) and WCAP-8171 (Non-Proprietary), June 1974.

8. Bordelon, F. M. and Murphy, E. T., " Containment Pressure Analysis code (C0CO)," WCAP-8327 (Proprietary) and WCAP-8326 (Non-Proprietary), June 1974 15.6-47 Draft Version

CPSES/FSAR

9. Bordelon, F. M., et al., "LOCTA-IV Program: Loss of Coolant Transient Analysis," WCAP-8301 (Proprietary) and WCAP-8305 (Non- l Proprietary), June 1974.

10. Bordelon, r. M., et al., " Westinghouse ECCS Evaluation Model -

Supplementary Information," WCAP-8471 (Proprietary) and WCAP-8472 (Hon-PMprietary), April 1975.

11. "Wastirighouse ECCS Evaluation Model - October 1975 Version,"

WCAP-862? (Proprietary) and WCAP-8623 (Non-Proprietary),

November 1975. l

12. Letter NS-CE-924 dated January 23, 1976. C. Eiche!dinger (West hghouse) to D. B. Vassallo (NRC).

13. Prosching, T. A., Murphy, J. H., Redfield, J. A. and Davis, V.

C., " FLASH-4 A fully implicit FORTRAN-IV Program for the '

Digitai Simulation of Transients in a Reactor Plant," WAPD-TM-84, Bettis Atomic Power Laboratory, March 1969.

14 Esposito, V. J., Kesavan, K. and Mtul, B. A., "WFLASH, A ,

F0f<TRAN-IV Computer Program for Simulation of Transients in a Multi-Lc'op PWR," WCAP-SP0O, Revision 2 (Proprietary) And WCAP-8261, Revision 1 (Non-Proprietary), July 1974.

Q212.80 6 14a. Skwarek, R. , Johnson. W. , Meyer, P. , " Westinghouse Emergency Core Cooling System $ mall Break October 1975 Model" WCAP-8970 (Proprietary) and WCAP-8971 (Non-Proprietary). April, 1977.

Q212.80 6 14b. Letter NS-CE-1672, dated January 1978, C. Eiche1dinger (Westinghouse) to J.F. Stolz (NRC).

i l

l l

Draft Version 15.6-48

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

a 1 CPSES/FSAR Q212.80 !

15. Kelly, R.O., Thompson, C.M., et. a'., " Westinghouse Emergency 6  !

I Core Cooling system Evaluat ton Model for Analyzing Large LOCA's During Operation with One Lonp Out of Service for Plants Without Loop 1 solation Valves," WCAP-9166, february 1978. l 0212.80 l

16. Eiche1dinger, C., "Westinghet se ECCS Evaluation Model February 6 1978 Version " WCAP-9220-P-A (Proprietary Version) WCAP-9221-P- -

A (Non-Proprietary version), February 1978. I Q212.80 16a. Eiche1dinger, C., " Westinghouse ECCS Evaluation Model - 1981 ORAFT I

Version," WCAP-9220-P-A (Proprietary Version) WCAP-9221-P-A l (Han-Pror,rietary Version), 1981, Revision 1.

17. Letter from T. M. Anderson of Westinghouse Llectric Corporation 6 to John Stolz of the Nuclear Regulatory Commission, letter

-number NS-TMA-1981, November 1,1978.

Q212.80 17a. Letter from T. M. Anderson of Westinghouse Electric 6 Corporation to R. L. Iedesco of the Nuc' ear Regulatory Commission, letter- number NS-TMA-2014, December 11, 1978.

I

18. " Westinghouse ECCS Evaluation Model Sensitivity Studies " WCAP-8341 (Proprietary) and WCAP-8342 (Hon-Proprietary), July 1974.

19. Salvatori, R., " Westinghouse ECCS - Plant Sensitivity Studies "  ;

WCAP-8340(Proprietary)andWCAP-8356(Non-Proprietary), July ,

1974.

20. Letter from T. M. Anderson of Westinghouse Electric 6- -

Corporation to John Stolz of the Nuclear Regulatory Commission, letter number NS-TMA-2030. January 1979.

20a. Johnson, W. J. , Massie, H. W. and Thompson, C. M.,- DRAFT

" Westinghouse ECCS-Four Loop Plant (17x17) Sensitivity Studies,"

WCAP-8565-P-A (Proprietary) and WCAP-8566-A (Non-Proprietary),

July 1975.

- .'15.6-49 DEAfT

CPSES/fSAR

21. DiNunno, J.J., Anderson, E. D., Baker, P.E. and Waterfield, R.

L., Calculation of Distance f actors for Power and Test Reactor Sites " T10-14844 U. S. Nuclear Regulatory Commission, Division of Licensing and Regule. tion, March 1962.

22. Postma, A. K., and Pasedag, W. F., "A Review of Mathematical Models for Predicting Spray Removal of fission Products in Reactor Containment Vessels," WASH-1329, U. S. Nuclear Regulatory Commission, Accident Analysis Branch.

23. Meteorology and Atomic Energy " U. S. Nuclear Regulatory _

Commission, Division of Technical Information,1968.

24 Murphy, K. G. and Campe, Dr. K. M., " Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19," V. S. Nuclear Regulatory Commission.

36 25. -Johnson, W. J. and Thompson, C.M., " Westinghouse Emergency Core Cooling System Eva luation Model - modified October 1975 ,

version," WCAP-9168 (Proprietary) and WCAP-9169 (Non-Proprietary). September 1977.

70 20.- Scatwright,,W. J., Maier, S. M. and Lo, S. S., " Design Basis Analysis of a Postulated Steam Generator Tube Rupture Event for -

Comanche Peak Steam Electric Station, Unit 1", RXE-88-101, TV Electric, March, 1988.

DRAFT- 27.- Chelecer. H. , Boman, L. ii. and Sharp, D. R. , " Improved Shermal' Cosign Procedure, " WCAP-8567 -July 1975.

28. Letter from E. P. Rahe of Westinghause Electric Corporation to Robert-L. Tedesco of the Nuclear Regulatory Commission. Letter number HS-EPR-23"R, December 1981.

Draft Version 15.6-50 ,

CPSES/fSAR

29. Letter from T. M. Anderson of Westinghouse Electri- DRAFT Corporation to Denwood F. Ross, Jr, of the Nuclear Regulatory i Commission, letter number NS-TMA-2354, December 1980,.

i

30. Letter from T. M. Anderson of Westinghouse Electric DRAFT Corporation to Denwood F. Ross, Jr. of the Nuclear Regulatory Commission, letter number NS-TMA-2379, January 1981.

31. Meyer, P. E., "NOTRUMP, A Nodal Transient Snell Break and ORAFT General Network Code," WCAP-100079-P-A (Proprietary), and WCAP- ,

10080-P-A (Non-Proprietary), August 1985.

32. - Ruprecht S. D., et al, " Westinghouse Small Break LOCA ECCS ORAFT Evaluation Rodel Generic Study with the NOTRUMP Code," WCAP-11145-P-A (Proprietary), and WCAP-11373-A (Non-Proprietary),

October 1986.

33. Lee, H., et al " Westinghouse Small Break LOCA ECCS Evaluation ORAFT ModelusingtheNOTRUMPCode,"WCAP-10054-P-A(Proprietary),and WCAP-10081-A (Non-Proprietary). August 1985.

4 4

)

I l

15.6-51 Oraft Version l

WMu u CPSES/FSAR

. TABLE 15.6-1 (Sheet 1 of 6) 11ME SE00EFfE OF EVENTS FOR INCIDENTS WHICH CAUS[_A DECREASE IN RE ACTOR COOL ANT INVENTORY Accident Event Time (sec)

Unit 1 Unit 2 84 Inadvertent opening of a $afety valve pressurizer safety valve opens fully 0.0 0.0- 84 _

Overtemperature N-16 5 reactor trip setpoint 5 reached 13.3 32.1 84 Rods begin to drop _ 14.3 34.1 84 Minimum VNBR occurs 14.0 34.6 84 5

Steam generator tube 70 rupture Start SGTR, Loop 4 5.0 5.0 DRAFT Reactor trip, turbine 70 trip, loss of offsite 70 power, AFW initiation 405 405 DRAFT 78 -

Close.MSIV, Loop 4 785 785 ORAFT Isolate AFW, Loop 4 905 905 ORAFT DRAFT l

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

CPSES/iSAR

. TABLE 15.6-1 (Sheet 2)

TIME SE0VENCE OF EVENTS FOR INCIDENTS WHICH CAUSE A DECREASE IN REACTOR COOLANT INVENTORY l

Accident Lygni Time (se.r)

, l Unit 1 Unit 2 DRAFT Begin RCS cooldown 1205 1205 DRAFT End RCS cooldown 1632 1632 DRAFT Begin RCS depressurizat'.on 1752 1752 DRAFT  !

End RCS depressurization 1907 1907 DRAFT $

Terminate ECCS flow 1967 1967 DRAFT I

Large break LOCA 5 1.-DECLG CD = 1.0 Start 0.0 -

Reactor trip signal 0.793 -

6 Safety injection signal 1.05 -

6 ,

Accumulator injection 6 begins 12.8 -

6 End-of-bypass- 23.49 -

6 End-of-blowdown 24.73 -

6 Pump injection begins 26.05 -

6  :

Bottom of core recovery- 36.83 -

6 Accumulator empty 47.02 -

6 ,

2.. DECLG CD = 0.4 Start 0.0 0.0 DRAFT Reactor trip signal 0.841 0.53 ORAFT Safety injection signal 1.65 1.62 DRAFT-Accumulator injection 6 begins 20.4 19.6 DRAFT'  :

l.

l ORAFT

. - . - - . . . - . - . - . . - . . ~ . - . . . . . . . - . . . . - ,- . .

[ CPSES/fSAR

..' TABLE 15.6-1 (Sheet 3)  :

TIME SE00ENCE OF EVENTS FOR INCIDENTS WHIC1LCAVE_A QECREASE IN REACTOR COOLANT INVENTORY .

Accident EyRD1 Time (utl ,

Unit 1 Un11_2 DRAFT Pump injection begins 26.65 26.62 DRAFT End-of-bypass 33.65 35.73 DRAFT ,

End-of-blowdown 37.66 35.73 DRAFT Bottom of core recovery 47.68 48.25 DRAFT Accumulator empty 56.91 54.08 DRAFT i 3. DECLG CD = 0.6 Start 0.0 0.0 DRAFT Reactor trip signal 0.815 0.52 DRAFT Safety injection signal 1.32 1.3 DRAFT Accumulator injection 6 begins 15.6 15.2 DRAFT End-of-bypass 25.52 28.69 DRAFT End-of-blowdown 28.65 28.69 DRAFT Pump injection begins 26.32 26.3 DRAFT Bottom of core recovery 39.07 40.7 DRAFT Accumulator erpt' 49.88 47.8 DRAFT DECLG CD = 0.6 Start -

0.0 DRAFT (Maximum $1) Reactor trip signal -

0.63 DRAFT Safety injection signal -

1.3 DRAFT Accumulator injection DRAFT begins -

15.0 DRAFT End-of-bypass -

28.69 DRAFT End-of-blowdown -

28.69 DRAFT Pump injection begins -

26.3 DRAFT l

l I

DRAFT

, , _,:..m _ . . ,___.._,,,v__ . - _. . , , ,r . , , ,,i., . - _ . . . _ _ , _ , , _ _ , . , , _ , , . - , , . . _ . . . ,-._-...-m- . , _

CPSES/FSAR

. TABLE 15.6-1

-(Sheet 4) ,

i TIME SE00ENCE OF EVENTS FOR INCIDENTS WHICH CAUSE A

, DECREASE IN REACTOR COOL ANT INVENTORY )

i Accident Event Time (sec)  :

Unit 1 Unit 2 - DRAFT Bottom of core recovery -

40.4 ORATT Accumulator ehoty -

48.3 DRAFT 4.-DECLG Co = 0.8 5 tart 0.0 0.0 DRAFT Reactor trip signa' O.802 0.51 DRAFT Safety injection signal 1.15 1.12 DRAFT Accumulator injection 6 >

-begins 13.5 12.6 ORAFT ,

End-of-bypass 23.72 24.96 DRAFT ,

End-of-blowdown 25.43 24.96 Pump irjection begins 26.15 26.1 DRAFT Bottom of tore recovery 37.03 36.5 DRAFT Accumulator empty 47.65 44.7 DRAFT Small break LOCA 6 6

1. 2 inch Start N/A 0.0 DRAFT t

Reactor trip signal N/A 62.9 DRAFT Safety injection signal N/A 73.9 DRAFT ,

Top of core. uncovered N/A 2381.2 ORAFT-Accumulator injection ORAFT begins N/A N/A DRAFT j Peak clad temperature. ORAFT occurs N/A- 4062.6 DRAFT Top of-core covered N/A -5512.5- ORAFT ORAFT

a CPSES/fSAR l TABLE 15.6-1 l (Sheet 5)

JIME SE00EPCE_QF EVENTS FOR INCIDENTS WHICH CAUSE A I DECREASE IN REACTOR CCO. ANT INVENTORY Accident Lyr01 Time (sec)

I Unit 1 1101L2 DRAFT

2. 3 inch Start 0.0 0.0 DRAFT Reactor trip signal 33.1 21.6 DRAFT Top of core uncovered 622.7 990.5 ORAFT Accumulator injection begins 2594.2 1999.8 DRAFT Peak clad temperature occurs 1363.9 1841.8 DRAFT Top of core covered 2308.6 3263.9 DRAFT

3. 4 inch Start 0.0 0.0 DRAFT Reactor trip signal 20.8 12.7 ' DRAFT Top of core uncovered 325 623.5 DRAFT Accumulator injection begins 795 887.6 DRAFT i Peak clad temperature occurs 836.2 948.0 DRAFT Top of core covered 848 1342.2 DRAFT

'i i

DRAFT

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

. ,a - CPSES/FSAR TABLE 15.6-1 (Sheet 6) >

TIME LSE00ENCE OF EVENTS FOR INCIDENTS WHICH CAUSE A ,

QECREASE IN REA(20R COOLANT INVENTORY Accident Event Time (serl Unit 1 Unit 2 ORAFT

4. 6-inch Start 0.0 N/A DRAFT Reactor trip signal 13.1 N/A ORAFT Top of core uncovered 133 N/A DRAFT Accumulator injection begins 335 N/A DRAFT L

Peak clad temperature '

occurs 212.6 N/A DRAFT 4

Top of core covered 367.7 N/A 0 RAFT i

E o

l 4

DRAFT l1

! i-

~~-

M CTSE8/r539 n3Lt 25.6-5 9 IWJT 4APAMETEPS 17?CD JN 3 W LOCA ANALYSf

\

i t! NIT 1 f74 Liceneed ccre power TWIT 1 Oste) LAPCE FFAJUt . UN7T 2 hs M or Coolant Prmp Seat  ;

34118 SHM1 WAAK , TMIT 2 \

(F5f)- J g EPEAT fCPA?T 3411a Peek linear poorer, C 3412e SKL4 BPUE DPAFT laclades 192* facter pwifel Ob 3,1* e Total peaking factor, FTg 12.8% ob f yajyy 11.44 oc 2.32 12.54 jcparT Power Shap. 2.32 12.88 2.32 l CPAFT 2.32 W esiae See Ftguze 15.6-44 fOFArt Chopped Costaa I See TApre f fCfAFT Fuel assembly array 15.6-69 lCFATT Acaumulador water volume, 17 a 17 aminal 2ft3/acccantiator) 17x17 Accusmletor tank volumme, e50 nomunal (ft3 950 Ortsmised 17 s 17 Accumulator gas pressure. /accamuletor) 135C minimum spet41 850 Optimized 17 a 17 f CFArt 1350 650 Safety inject ion pomped flow 600 1350 f DupT 600 1350 Fee Figure 15.6-47Ad s 4C0 lD5 AFT Containment parameters Figure 15.6-4?t i 60 0 See Figure 15.6-47Ad fCFAFT Inzt1=1 loop' flow See Fiore (1b/see) See Section 6.2 NA fOFIFT Veseel inlet 9902 15.6-47Ts4 te v rature fors *743 S**

Sectar.n 6.2 MA fDSAFT vessel outlet temperature (or) $56,5 9981 564.8 fLFAFT 98!66 Reac*er coolant pressure tpsia) 617.2 $59.3 (27.6 l OPATT Stems pressure (pria) 2250 5(4.1 415.7 l DParr 224C 623.3 Steam generator tube plugying level 965 2260 W 2000 fDFAFT Od 2290 994.7 f DFATT C4

• Two percent Od 1000.0 is m<ited to this power to .5 fJFAFT

> Pump heat mo3esnt is aestmed to t.e generated in thee. corfor caloriswtric error. f D$Aff ca..> e.dme y S= ,.

+a2.e.% no . f76 l7 IvFArr J ru,T

!l1liI I.Il1i e T 2 T T T T T T T T T

- F F F & F F F F F F F A A A A A A A A A A A R R R R R R R R R R R D D c D D O D D D D D l li5l  ;

i v

a 5

5 5 9 2 5 8 2 3 .

4 5 7 9 7 1 7 C 9 0 I 4 <

S -

= 1 m

p u c m i

, n C i L M C

E D

5 I 4 S 5 0 5 3 9 0 0

m s 7 I 7 o 4 7

= u o t 7 E m e T pi l L c x U a S , M E C R L C

D E N D A

T U

6 P t

)

M S R. 6 2

I 6 2 5 S m 9 5 3 a 7 F 5 S 0 u 5 t

/ 1 I m i 7 0 2 6 e

S

+

S - i 4 7 < 7 E E Y n 0 S L h L s 7i 9 P B ( A CM 1 C A N T A ,

C .

- L C

K E ,

E D E

R B 5 8 T 7 5 3 6 2 E S 5 9 2 G 0 o 5 7 R m 5 7 1 7 < 7 A = u 7 L m 7 Di 1 C n i

, M c

L C

E )

D s

) d F ) ) n )

• t r t o t

( e e 4 c e e t

• e e e f a f r s f r ( w ( e ( (

u / t t c d n a e n a o a o w m o -

r l l i / z i 1 e t c t d t t p e a e a t m c l c l t c i e o a ) o c ) s o n t L c % L  % r L U t ( e ( u d I r b a n o n e l mo c o d t c u i i o l m t l +. r u t i c a c s r x a t a t e o a e o e o R P M r T r H l i i ! l llll; llt 1l,

CPsts/rsAR TABLE 15.6-6 (sheet 1 of 2)

LARGR BPEAR . analyst 3 typpy ggp pg3pgyg Quantities in the Calculationet Q212.134 Licensed core power rating 4 102t of 3411 MWe Tots 1 core peaking factor 6 2.32 Peak itnear power 6 1024 of 12.43 KW/ft for Unit 1, ORAFT Accumulator water volume 102% of 12.62 RW/ft for Unit 2 D P. A F T Accumniator presenre 850 cubic feat par tack 6 600 paia Number of Safety Iniection Pumps Operating 3 6

Steam Cenerator Tube Flugging Level 6 C percent ruel Faramstore - Cycle 1, Region 1 78 76 Reeults. Unit 1 PECLC, C p = 1.0 DECLG, C = 0.3 p pretc, c 0.4 f6 g DECLC, C D

" 0+4 Peak clad temperature (*F) 1969.2 1945.2 2010 7 1577.P Location (feet) 7.5 7.5 7.5 l6 7.5 l6 Hesimum local cled/ water reaction (t) 6 3.s 3.06 3. 92 0.72 6 Lacetion (feet) 7.5 7.5 7.5 7.5 6

• etal care clad / water reaction (t) 6 3.3 0.3 0.1 J.3 6 Est rod burst time (seconds) 26,2 29.0 27.5 w& 6 Location (feet) 6 . C* 6.0 f,0 NA 4 l

76 76 DPET m ,

.' CPSES/FSAR TABLE 15.6-7 S!!ALL BREAK LOCA RESULTS FUEL CLADDING 06]A Pipe Break Size 3 Inch 4 Inch 6_lDIh Unit 1 Results ORAFT Peak clad temperature (OF) 1499.2 1787.5 1343.5 6 i

Peak clad temperature location (ft) 11.25 11.50 11.0 6 Local Zr/H 2O reaction, maximum (%) 0.683 3.21 0.382 6 Local Zr/H2 O location (ft) 11.25 11.25 11.0 C Total Zr/H 2O reaction (%) <0.3 <0.3 <0.3 6 Hot rod burst time (sec) N/A 758.1 N/A 6 Hot rod burst location (ft) N/A 11.25 N/A e

Pipe Break Size l DRAFT 2 Inch 3M 4 Inch ORAFT Unit 2 Results DRAFT Peak clad temperature (OF) 1005.3 1433.8 1290.9 DRAFT -

Peak clad temperature location (ft) 11.5 11.75 11.5 ORAFT Local Zr/H O 2 reaction, maximum (%) 0.05 0.60 0.11 DRAFT Local Zr/H2 O location (ft) 11.5 11,75 11.5 DRAFT Total Zr/H 2O reaction (%) <1.0 <1.0 <1.0 DRAFT Hot rod burst time (sec) N/A N/A N/A ORAFT Hot rod burst location (ft) N/A N/A h/A DRAFT DRAFT

i 4

4 l ti 946-3 w L F

0 C

L T

A A

S H

CORE PRESSURE, CORE FLOW, HIXTURE LEVEL, (UNIT 1)

(UNIT 1)

AND FUEL R0D POWER HISTORY N

0 T L 0 < TIME < CORE COVERED R 0 U y C M  : T P A (UNIT 2) (UNIT 2)

COMANCHE PE AK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNITC 1 and 2 Code Interface Description for Small Break Model FIGURE 15.6-6

~

l

.. l l

2500 l

2000-I f

1

.1500-I

$3  !

1000-a 500-0 , ,

0 -50' 100 150 200 250.300 350 400 450 500 550 FLOW RATE (LB/SEC) .

4 COMANCHE PEAK S.E.S -

FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2 Large 8 teak Safety injection Flow Rate oc7 N824-Figure 15.6-47A

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

l a

2500 2000-n 0

05 b 1500-A3 b

,n. 1000-500-

, 0 i -- i i- i i i i- i 0 50 100 150- 200 250 300 350-400 450 500 550-FLOW RATE (LB/SEC)

COM ANCHE PEAK S.E.S FINAL SAFETY ANALYSIS REPORT UNITS 1 and 2 '

Smail Break Safety injection Flow Rate

'006 38824 Figure 15 u-478 n, : , , . , _ . , - _ . . ~ , . _ . . , _ . . . _ . . . . . . . ~ . _ . . _ _ _ . . . _ . _ - . . . . _ . . . . _ - - . . , _ , . _ , . .

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

9 4 .-

-

• 104674 p 2500 0,

O 7.5 FT 1 g- @ 2000 -

h V

. w- _

1 S 1500 - -

y

- 1 E

j 1000 - 6.75 FT

-W

-O

- g-

-<[ 500' -

O-

. .a u 0 l l l l

-0 ~50 100 150 200 250 TIME (SEC) 1 COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT-UNIT 2 Peak Clad Temperature DECLG (CD ' O.6). Min SI FIGURE 15.6-49 mw- -

10457 6 C 2500 C.

O O ---

7,5 FT 2000 -

Y 1500 -

7.0 FT 2

y 1000 -

8 cc

$-500 -

O d 0  !  !  !  !

~

0 50 100 150 200 250 TIME (SEC) 1 COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 Peak Clad Temperature DECLG (CD = 0.6) Max St FIGURE 15.6-49 A

~ _ . _ . + . . ._ - - . ,

104$74

-2500' 20nD --

2 1500 -

g e.

w e

o 8

w.

g 1000 -

500 -

0 I I 0- 10 20 30 TIME (SEC) i, l

! ' COMANCHE PEAK S.E.S.

FINAL SAFOY ANALYSIS HEPORT UNIT 2 Core Pressure DECLG (CD = 0.6) -

FIGURE 15.6-50

.. . . . . . -. .. -.....v,..... . . . . ~ . ~ . . - _ _ . . - _ _ . .

.- . .~.

4

• 10%M 20.0 17.5 -

15.0 -

I U 12.5 -

a

' w.

uj.10,0 -

x N 7.5 -

3l 5.0 -

2.5 -

I 0  !  !  !

-0 100 200 300 400. 500 TIME '(SEC) ~

1 COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 Reflood Transient - Core & Downcomer Water Levels DECLD (CD = 0.6) Min St FIGURE 15.6-51

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

7- 10457a 20.0 -

17.5 -

15.0 -

P-E 12.5 --

a w

3 -10'.0 -

a

$ .7.5 -

~^

f .

, 5.0 -

' 2.5 0 l I I l

. 0 100 200 300 '400 500- ,

TIME (SEC).

b l

l COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 Ratiood Transient - Core & Downcomer Water Levels DECLD (CD a 0.6).Mu St FIGURE 15.6-51 A

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

t: ,;

,q .

-.104674

. .7 1

4 u -

T

~ 2.00 '

1.75 -

3

, 1,50.

o-

- =

= g 1.25

=

' N 1.00 -

<t cc O .0.75 o: -

N L

' O.50 - 'N.

l.

0.25 -

o- -I I l l

.0 100 200' 300 400- 500 TIME (SEC) ,

[

.-(.: '

l COMANCHE PEAK S.E.S. -

Li FINAL SAFETY ANALYSIS REPORT.

L, UNIT 2 l

i Reflood Transient Coro inlet Velocity t DECLG (CD - 0.6) Min S1 FIGURE 15.6-52

10457-10 2.00 e

1.75 -

I

_ 1.50 -

S

-e

. g ' 1.25

=

1. _1.00 -

4

.x

- @ 0.75 o

a-

' O.50 -

0.25 --

0 l l l l-01 100 -200 .300 400 500 TIME (SEC) l-l It CDMANCHE PEAK S.E.S.

! FINAL SAFETY ANALYSIS REPORT l- UNIT 2 Reflood Transient Core intet Velocity DECLG (CD = 0.6) Max Si p FIGUflE 15.6-52 A

.)

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

,,' - 10457 11 2.00 1.75 -

1,50 .

1.25 -

6 E

e 1.00 -

w B:

o a

0.75 '

0.50 - -

0.25 - - .

O I i 0: 10_ 20 30-TIME (SEC) l l

l COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPOP.I-

[.-

UNIT 2

! ~ Core Power Transient DECLG (CD = 0.6)-

FIGURE 15.6-53

e

• 10457-12 30 25 -

3 03 9.

w 20 - -

c D

E

a. 15 - -

2 w

E E 10 - -

5 z

O O

5--

0  !  !  !  !

0 100 200 300 400 500 TIME (SEC)

C0MANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPOP.T UNIT 2 Containment Pressure DECLG (CD = 0.6) Min St I

FIGURE 15.6-54 i '

at

, 104h713 25 5 20 - -

- Di S:

w-

. =.

h.15-w E

H

-z

- g 10 .'_

a ,

-- g-

2 y

O - - -5 -

- 0' .  !

0 - 100 200 300- 400 500 TIME (SEC) 4 COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 -

Containment Pressure DECLG (CD a 0.6) Max Si FIGURE 15.6-54 A

. 10457It, 7500 5000 -l

_ 2500 - -

f 1

S f .

j--TOP

$ /'

/

N / L  %.?

3 L. BOTTOM

)

+

N 2500 -

-5000 - - -

7500 I I o 10 20 30 TIME (SEC)

COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 Core Flow (Top and 53ottom)

DECLG (CD = 0.b) Max Si flGURE 15.6-55 I l

~

(G, -  !

.s 10457 15 2 j 3

l 10 3 _

=

5 -

z 2 E (-

u. )

2 (E _ 10  : ,

, u. u. 1 0& 5  :

b"H I ~~I

~ \! 7.0 FT -

ge 2 - x _

z3 H

c: 8 101 1 F --

i.

Q 5  :.

w I __

7.5 FT 2 -

100 i l l l d 0 50 100 150 200 250 TIME' (SEC's l.

i

{

l COMANCHE PEAK S.E.S.

l' FINAL SAFETY ANALYSIS REPORT UNIT 2.

i:

Heat Transfer Coefficient DECLG (CD = 0.6) Max SI FIGURE 15.6-56

.. l

\MI*lO i

i 2000 --

1750 -

[1500 w

$1250 F- 7.5 FT J

a 1000,I- -

2 .I -

p

,$-750

.o: k 5- J a

-: 500 7.0 FT 1250 -

-: o l l l } i OL SO - 10n 150 200 250 TIME (SEC)

.i

.i CDMANCHE PEAK S.E.S. 1 FINAL SAFETY ANALYSIS REPORT I UNIT.2 q i

Fluid Temperature _ '

DECLG (CD = 0.6) Max St FIGURE 15,6-57 il

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

f -

10457 17

.g . -

3 10.0 7.5 --

5.0 -

'oe "

.X g 2.5 -

a 3- 0:

3:

- O -

.J N

w 2,5 -

4 -

w CC c3

-,5.0 -

'l

7.5 -

10.0 t I O 10 20 30-4 TIME (SEC) v CDMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2'-

Break Flow Rate OECLG-(CD = 0.6) Max Si FIGURE 15.6-58

)

- s l 10457 18 e

5 e

3 -

o e-X U

$ 1 -

3

~

> 0 0

x y.,1 -

w M

<:t .

w

'cc 3 -

1 5 l '!

0 10 20' 30 TIME (SEC)

COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 Break Energy Released to Containment (CD = 0.6) Max SI FIGURE 15.6-59 w

10457 19 1.50 1,25 - ~

- 7.5 FT 3

1.00 -

9 I 6

o a a 2 g 0.75 -

b 7.0 FT a

y 0.50 -

0 0.25 -

0  ! ! ! !!!!!  ! ! ! !!!!!  !  ! ! !!!!!  !  ! ! !!!!!

0'l 2 5 10 0 2 5 10 1 2 5 102 2 5 103 TIME (SEC)

COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 Fluid Quality DECLG (CD = 0.6) Max Si FIGURE 15.6-60 l

1

is --

. 1005700

.4 m

i 1.0 l

l l

0.8 -

"o e

va -

's 0.6 -

d

- 3:

O Z

O 0.4 -

4 J

O E

D 0

4 ' O.2 -

O l  !

0- 10- -20 30 TIME (SEC) l L

- COMANCHE FEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT-UNIT 2 4 Accumulator Flow (Blowdown)

DECLG (CD = 0.6) Mar St

>- FIGURE 15.6-61 l-t

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

s. 1 . . . . . .. . _. - . . _ . . ..m ... . . . . . . . . . , _ . . . . . . . . _ . . . _ . . _

1046701

l

.g 7

i

.- c V

300 e

U'.200 -

E.'

.N-

[ 7.5 FT

\

._ -100 g.

c o.

d y

N 7.0 FT

.a 2 -~.100 -

s 4200 I ! l lllll l l llllll l l l lllll l l l lllll

. 10'l '2 5 10 0 2 5 101 2 5- 10 2 2 5- .103 b TIME (SEC)

^

l t.

COMANCHE PEAK S.E.S.-

FINAL SAFETY ANALYSIS REPORT-

. UNIT 2 l

Mass Velocity DECLG (CD = 0.0) Max Sl FIGURE 16.6-62 L-t ;-

l ':

(l- 8, .

1045702 I

< 20.0 ,

17.5 -

U w

[r15.0 -

S Q .

O .-

O g 12.5 -

w x

0 2

2 10.0 -

-o ,

-Q

. g- / '

-O

$ 7.5 -'

8 '

0 W ,

@ 5.0 -

c E

o

- c.

2.5 -

0  ! I l 0 100 200 300 TIME (SEC) .

COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 Pumped ECCS Flow (Reflood)

DECLG (CD = 0.6) Max SI I

FIGURE 15.6-63 L - -

(c u ..

10457-2J

- _ 2500 u.

o_ .

O <- 7.5 FT

• -2300 -

H '

-O

= y- _

- w-

-5[1500 g -

<=

c:

E 7.25 FY fh1000 _

w-o ,

-x y :500 -

O-g -g- l , g

--u- o r i I i 1 0- 50 -100 150 200 250

. TIME -(SEC)

[--

t . - COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT

.. UNIT 2 Peak Clad Temperature - .

DECLG (CD = 0.8) l .

! FIGURE 15.6-64 t

i.

M _ _ .

4 e

10467 2f i

2500

?D00 -

2 \, '

(1500 d E

1000 -

E \

i 600

}

'~~

O 0 6 10 15 23 25 TIME (SEC) i

\

. COMANCHE PEAK S.E.S.

t FINAL SAFETY ANALYSIS REPORT UNIT 2 1 l Core Pressure DECLG (CD = 0.8)

FIGURE 15.6 65 l .o .

O

. 'WL 7-26 i

l l

l i

j 4

20.0 17.5 -

15.0 -

G S 12.5 -

d d 10.0 --

a

((

Y 7.5 -

4 ]

a 50 -

l 2.5 -

0 0 100 200 300 400 500 -

TIME (SEC) i i

I l

l 1

! COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 Rettood Transient - Core & Downcomer Water Levels DECLG (CD = 0.8)

FIGURE 15.6 66

- .-,.. - - . -.-.-..--.~... - ....- . -..-..-- .~... . . . - - , . - - - . . . - -.

.I.-

e

.E' 10467 76

)

2.00 1.75 -

l

_ 1.50 -

o w-

, p 5

z 1.25 r

C-W r 1.00 -

4  ;

. E O 0 75 -

8 a.

W 0.50 -

0.25 -

b O

o- .100 200 300 400 500 .j TIME (SEC)

P i:

l I.

p-l' y

COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT-UNIT 2 i

Reflood Transient Core Inlet Velocity  ;

DECLG (CD a. 0.8)

FIGURE 15.6 67 l.

-- , ..~ -

4 10067 77 h

I 1

2.00 h

j 1.75 -

1 1

1.50 -

6 1.25 -

A E

g 1.00 l-w t

@ 0.75 0.50 - -

0.25 -

f f I f  ! >

.0 ' ' ' ' '

o :5 10 -15 20 25 TIME (SEC) 1 COMANCHE PEAK S.E.S. .

FINAL SAFETY ANALYSIS REPORT - (

UNIT 2 -

Core Power Transient DECLG (CD = 0.8) .

, =  !

l .- FIGURE 15.6 68 i

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

, -. _ _ .- ...._.__e . ._ .. .-.~ _.

10457-78

..~

.I

%0 l

l 40 - l 9 i 2 l

(

W ,

'i g

D 30 - 1 8

w

)

' r-p 2 I

- w 20 -

lE .

E H

2

- 8 10 -

0 I l l l 0 100 200 300 400 500 ,

TIME (SEC) i r

E COMANCt!E PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 Containment Pressum DECLG (CD = 0.8) .

FIGURE 15.6 69

( ..

. . - . _ . . . _ . . . .. . . _ _ . . . . . . . - . . . . , _ . . . , _ . . . . . , . .....-..._u....,...-- _ , . . . , . _ . - _ _ . - ~ . . - . . . . . . . . . _ _ _ _ . , _ . . . , .

to457a l

g 2500 -

L o

O

[2000 -

7 .5 FT e

f

& 1500

< N ,,,,,

cc E 7.25 FT 2

L y 1000 8

4 tt y 500 -

C 0  !  !  !

O 50 100 150 200 250 TIME (SEC)

COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 Peak Clad Temperature DECLG (CD - 0.4)

FIGtlRE 15.6-70

J. ., l Q l 1005740 I

• l l

i 1

2600 2000 t

<t -

-G 1500 -

6 w

CC D

d

$1000 n.

P 500 -

0  !  ! I '

0 10 20 30 40- ,

TIME (SEC) l  ?

l. I i

l COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT -

UNIT 2 Core Pressure DECLG (CD = 0.4)

FIGURE 15.6 . . . . . _ _ . _ _ - . . _ . . , _ _ _ _ _ _ _ _ _ . _ _ _ , . , _ _ . _ - _ _ . . _ _ , . . . _ . . . _ . . _ . . _ . . _ . _ _ - - _ . - - , _ , - . .

.. - -. . . - ~ . .

?e o i e 1045741 i

l 20.0 -

17.5 15.0 -

P E 12.5 -

a w

uj10.0 -

x

1. 7.5 -

C ,

5.0 -

i 2.5 -

0  !  !

o 100 200 300 400 500 TIME (SEC)

I

~

COMANCHE PEAK S.E.S.

l- FINAL SAFETY ANALYSIS REPORT ,

UNIT 2 Reflood Transient - Core & Downcomer Water Levels DECLG (CD = 0,4)

DGURE 15.6-72

[ ..

1046742 2.00 1.75 -

1.50 --

8v>

]= 1.25 N 1.00 -

4 C

@ 0.75 O

d 0.50 -

0.25 -

0 I I O 100 200 300 400 500 TIME (SEC)

COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 Reflood Transient Core inlet Velocity DECLG (CD = 0.4)

IlGURE 15.6 73

104 743 I

2.00 1.75 -

1.50 -

1 ,

i.25 -

3 c.,

E

~ 1.00 c:

w C

O a 0.75 -

i l

0.50 - -

s

'O.25 --

?

g i i 1 0 10 20 30 40 TIME (SEC) i F

COMANCHE PEAK S.E.S. .

FINAL SAFETY ANALYSIS REPORT UNIT 2 Core Power Transient DECLG (CD = 0.4)

FIGURE 15.6-74

+

e

. iO4b7J4 25 g 20 -

G b

w CC h

w 15 -

E H

2

• 10 ' -

a H

2 85 . . .

N O l l l l 0 100 200 300 400 500 TIME (SEC)

COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 Containment Pressure DECLG (CD - 0.4)

FIGURE 15.6-75

2400.

2200.

• 2000.

1000.

~

31600.

E w 1400. .

g 1280.

lee 0.

800. ,

600. - v'

488 1580. 2889. 2SM . 5800. 5580. 4888.

1. 588. 1980.

TIME (SEC) l COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 RCS Detiressurization Transient (3 Inch Break) rinunt 15.6-76

4 t

6 4

e 4 l

'\ .J s3. -

\ '

I 1 i i

% l l

29. i

, !i I .

as. -

I w

l i 5

. y 2.s .

i ---

]~

z TOP OF CORE s'

- - - - - - - - - - ' - - - - ~ ~ - - - ------> -

f '

22.

(

==---*l---- -

\

23.

\

i r ,

- J \

19 2000. 2500. 50ife. 5500. 4000.

27 500. 1000. 1500.

TIME (SEC) i l

l i-COMANCHE PEAK S.E.S.

f FINAL SAFETY ANALYSS REPORT (NT2 .- -

Core Mixture Height (3 Inch Break) o ___

noune 15.6 77 l

L l.,+ .-,w.y,. . . ,n

--, _ , , , . . . . . . . , , _ , . . . _.,,,.m......r_m, ,_ _, _.,..%,, .,.-_...,,.y.,,_,.,.,.,yy....

15C0 J l

1400 ,

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3 u 500. I E --

w 7en, _ _ - -

E 9 600. --

I 6 _ _ _

I I

2400. 28C0.

500 1200. 1600. 2000.

00.

TIME (SEC) l.

i- COMANCHE PEAK S.E.S.

l FINAL SAFETY ANAC(SIS REPORT .

UNIT 2 Clad Temperature Transient l (3 Inch Break)

L .

MOURE 15.6-78

. . , , - , --, ,,,,, .. , . . . .,n < . . . , , , . -,n.-,- , , - , - - . . , , , , , . . , , , . . . . ,, - r

O I

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m 5

=

_2 -

l10' y

((

5 L3 -

t2 -

l10-2 y O E

5

=

n --

a 4

2

!O 10-8 me-E z 5 =

2 -

10-' ' I I !IIII I ! I III!! I ! !IIIII I I IIIIII  ! I ! IIIII 10-2 2- 5 10' 2 s 10* 2 5 10 2 2 5 10 2 2 5 10" TIME AFTER Trip (SEC1 R

l-COMANCHE PEAK S.E.S.

l- FWAL SAFETY ANALYSIS REPORT +

v LWT2 Core Pcwer After-Reactor Trip 1

nount 15.6-79 l: ,

220. -

l 200.,

190.-

l!

l I

- 160. ,, n- ,, ;

l C I g 143.1 3 d

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g 60.[-

i :

4a. -

4l

20. - --

0

2. 500. 1000. 1500. 2000. 2500. 5000. 5500. 4000.

TIME (SEC)

COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 Steam Flow (3 Inch Break)

F6GURE 15.6-80

4 10 __

b 7 m 1 1

~

C s

E 10 3 S -

• i e i 5 -

U t

u.

v I

5

" 102 1 _,,

n g __

g . .

= ) _ - t I

""' N - ...ml. ._ m l'

10 1 800. 1200. 1600. 2000. 2400. 2800.

TIME (SEC) l l

l

..y COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT l_

UNIT 2 Rod Film Heat Transfer Coeff;cient (3 Inch Break)

FIGURE 15.6-81 l

!200.i ,

l I, , .

t l

l M --

i 00. j'  ;

l

/

/ ,

1000.

/

I

/

/

n ( ,

' )

w 400. ,I l

W i j

er j N

g 900. [

l (I w 1

• l $

ca 700. i

[3 u.,

600. j

\.

500.

I \1'.-

400 ~

3. 500. 1000. 1500. 2000. 2500. 5000. 5500. 4000.

TIME (SEC)

P

_d COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 Hot Spot Fluid Temperature l.

(3 Inch Break)-

FIGUFIE 15,6 82 l

2488.

2200.

2000. --

q 1800.

E b

$$1600.

5?

d E i 1400. -

O ac I I Ny.

w 1200. i

\ -

1880.

N% '

~

880 2989. 5888. 4ese. 5884. 6800.

B. 1998.

TIME (SEO COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 RCS Depressurization Transient (2 Inch Break)

F60VRE 15.6-83 ,

l 2496.

2200.

2000.

I 1880.

1600.

Q E A

$ 1400, w

h1290. 3 1200. -

s S

880.

N 688.

A 480.

299 M. 298. 488. Gee. 800.1996.1290.14N.16N. lese. 29N.

TIME (SEC)

COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 RCS Depressurization Transient (4 inch Break) nounE 15.6-84

$1.  ;

I 50.

\ l l

24

! l I

29.

N =-

,\ d- } i' 27.

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,\

g 26. ,

b Q 25.

W 24.

\

25.

TOP OF CORE 22.

~ ~ ' ~ - - ~ ~ ~ ~ ~ - ' ' ~ ~ ~ '~ ~ - ' " - ~ ' ' - - - ' ~ ~ ~ ~ ~ ' ~ - - - ~ ' ' ~' - 'd - ~ ~

~

r W,g /

20 6000.

B. 1000. 2000, 5000, a000. 5000.

TIE (SEC)

COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 Core Mixture Height (2 Inch Break) nounE 15.6-85

52.

l t

53, S

\ ,

I l

- l

28. c 1.h ) i

(

l

20. ,

e~

24

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TOP OF CORE

[4 4 I

%s YNj f

20. ,

\ '

I.

ia.

Miy ,

F i

' B. 200. 400. 600. OfuB . 1000. 1290. 1400, 1600. 1800. 2000.

TIME (SEC)

COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 Core Mixture Height (4 Inch Break) nounE 15.6-86

l 1050.

' ^

10:0.

050-N/

c "w 900.

r E

lw 850 J f x f

b,_ 800 -

l

(

• g 750.

Yac 700.

v g 650.

u

>= '

@ 600. q 550.

000. 2500. 3000. 3500. 4000. 4500. 5000. 5600. 6000.

TIME (SEC)

COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 clad Temperaturc Transient (2 inch Break) )

nount 15.6-87

1300. ,

l i  !

12^0.

f i  !

,1100.

I 1

i

• 1000.

E zW

! ~

g 900. /)

w d I --

g 800.

W j 700. -

u I 600. -

i-- .'  !

8 NJ 500.

L ~m

- m q l

200. 400. 600. 800. 1000. 1200,1400. 1500, 1800. 2000.

TIME (SEC) l l

l

[

COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UMT2 i Clad Temperature Transient (4 Inch Break)

FIGURE 15.6 88 ,

I i

13 -

~

12-11-10- j E.

9-

"c c

8-w 3 7 O  :

n.

k o 5- >

I 4- ,

s 3-2-

1 , , i i , , , , ,. , ,

0 1 -2 3 4 5 6 7. 8 9 c10 11 12 i CORE HEIGHT (FT) ,

l l

COMANCHE PEAK S.E.S.

FINAL SAFETY ANALYSIS REPORT UNIT 2 Small Break Power Distribution l Fiount 15.6-89

- , :, , u . . :- - .-- . . ., ; _ u. . . _ . . . _ . . . . _ _ _ _ . . . . . _ . . . _ . . _ _ _ _ _ _ _ . - _ _ _ . - . _ _ _ _ . . - . _ _ ._

I CPSES FSAR AMENDMENT 85

. DETAILED DESCRIPTION Page 1 FSAR Page (n imcDded) GmVP Drscripti00 4.3-33, 74 2 Expand discussion to recognize that operations support is required in addition to testing and to identify that NRC approved TV Electric methodology is employed for both testing and operstion support.

Revision:

NRC has approved TU Electric Control Rod Analysis methodology for generic application. The TV Electric Steady State Physics methodology has been approved for use on Unit 1. The FSAR has been revised to reflect the intent to apply the methodology to Unit i reload cycles and Unit 2 initial startup and reload cycles.

FSAR Change Request Number: 91-141.99 SER/SSER 1mpact: Yes Safety evaluation written by NRC on TV Electric Steady State Physics topical only identified applicability to Unit 1. Applicability is being expanded to include use on Unit 2.

Table 4.3-2, B 2 See Sheet No(s):1,2 Correct Nuclear Design Parameters Correction:

The boron coefficient for boron dilution was not re-analyzed for a change in the Unit 2 shutdown margin until af ter Amendment 84 was incorporated into the FSAR. This number is being corrected for the reanal-ysis, in addition, a typographical error was corrected for neutron lifetime (20.7 microseconds vice 21.7).

FSAR Change Request Number: 91-141.35 SER/SSER 1mpact: No 15.0-12 3 Correct Statement Concerning Axial Power Shapes Correction:

In Amendment 84, changes were made to reflect the Unit 2 differences, but a statement concerning the axial power shapes used in the DNB calculations was not changed to teflect Unit 2. The Unit 2 analyses do not use a 1.55 chopped cosind shcpe. The axial power shapes for each unit are discussed in section 4.4.

FSAR Change Request Number: 91-141.05 SER/SSER 1mpact: No Table 15.0-2 2 See Sheet No(s):3 Correct Reactivity Coefficients Used in the Rod Withdrawal at Power Accident Correction:

In Amendment 84, the reactivity coefficients were mod-ified to reflect the Unit 2 analyses and the reanalyses associated with the change to a positive moderator co-

, CPSES FSAR AMENDMENT 85 DETA! LED DESCRIPTION Pege 2 FSAR Page (ts amended) EEWP h1Er_lption efficient in Unit 1. The value for the moderator den-sity coefficient was inadvertently left out of the Amendment. The correct value of +.43 delta k/gm/cc has been added to the Table for the Rod Withdrawal at Power Accident.

FSAR Change Request Number: 91-141.05 SER/SSER 1mpact: No Table 15.0-2 2 See Sheet No(s):5 Add Unit 2 Computer Codes and Initial Conditions for the Large and Small Break Loss of Coolant Accidents Addition:

Table 15.0-2 summarizes the Computer Codes and the range of initial conditions assumed in the analyses.

The Table has been revised to include the Unit 2 codes and initial conditions.

FSAR Change Request Number: 91-141.05 SER/SSER 1mpact: No Figure 15.0-2, B 2 Correct Doppler Power Coefficients Used in the Accident Analysis Correction:

In Amendment 84, figure 15.0-2B was inserted for the doppler coefficients used in the Unit analyses. The Figure inadvertently inverted the upptr and lower dop-pler curves. A new figure 15.0-2B has been provided.

FSAR Change Request Number: 91-141.05 SER/SSER Impact: No Table 15.1-3 2 See Sheet No(s):3 Parameters for Postulated Main Steam Line B-eak Accident Revision:

The Unit 2 parameters have been added tc Table 15.1-3.

Unit 2 has a different parameters because the steam generators are different from Unit 1. The parameters have also changed for Unit 1 because of a change in the relief valve setpoint.

FSAR Cnange Request Number: 91-141.99 SER/SSER Impact: No 15.4-33, 34 2 See Page No(s):37, 38 Add Unit 2 Parameters and Results for Boron Dilution '

Event During Startup and Power Operation Addition:

The Boron Dilution Event has bee reanalyzed for Unit 2.

The Unit 2 input parameters (e.g., RCS volume, boron boron worth, etc.) and results (e.g., time for operator

. CPSES FSAR AMENDMENT B5 DETAILED DESCRIPTION Page 3 FSAR Page (as amended) Gr0UP DriLI.lPUQD action, etc.) have been added to  :,e discussion.

FSAR Change Request Number: 91-141.52

'alated SER Section: 15.2.3.1 SER/SSER 1mpact: No Table 15.4-1 2 See Sheet No(s):4, 5 Add Unit 2 Sequence of Events for Boron Dilution Event During Startup and Power Operation Addition:

The Boron Dilution Event has been reanalyzed for Unit

2. A new column has been added to Table 15.4-1 to add the times for the Boron Dilution Event during startup and power operation.

FSAR Change Request ' tun.ber: 91-141.52 Related SER Section: 15.2.3.1 SER/SSER 1mpact: No 15.6-14 2 Remove Specific Referance to the DNBR of 1.50 Addition:

Each occurrence of "DNBR is less than 1.30" (departure from nucleate boiling rat',o) has been replaced with a generic term, "less than the limit value" 5ecause the DNBR is different for each unit and may change in the future if the fuel type changes, or if a different anslytical method is used in the analysis. This change does not have a material effect and has been made to preclude repetitive changes in the future. The DNBR limit is listed in Technical Specification 3/4.2 for each unit.

FSAR Char.ge 16 quest Number: 91-141.99 SER/SSER 1mpact: No 15.6-22, 24 2 See Page N3(s):26 thru 35 Add Unit 2 Assumptions, Initial Conditions and Results for the large and Smali Break Loss of Coolant Acchients Addition: "

Tite large and Small Break Loss of Coolant Accidents were enalyzed dif f crently for Unit 2. The large Break LOCA was analyzed using the approved 1981 ECCS evalua-tion model. Tha Small Break LOCA was analyzed using the May, 1985 N0lkUMD ECCS evaluation modst. The assumptions, initial ccnditions and results, which are different for Unit 2, have been added.

FSAR Change Request Number: 91-141.01 SER/SSER Impact: No 15.6-32 1 Changes to Peak Clad Temperature (PCT) Fenalties and Final Limiting PCT for Unit 1 Large Break LOCA

7i -,

, CPSES FSAR AMENDMENT 85 DETAILED DESCRIPTION Page 4 FSAR Page (M AEndfd) DIDi!D QCitr_lD110D

, Revision:

An error in the ECCS calculation for Unit i resulted in a PCT penalty of 7.2 degrees Fahrenheit (F) which increased the total PCT penalty to 55 degrees F and the final limiting PCT to 2055.7 de'grees F. The 7.2 degree F penalty was for steam generator tube collapse due to concurrent seismic and LOCA loads. TU Electric noti-fied the NRC that the total PCT penalty exceeded 50 degrees F, in accordance with 10CFR50.46, via letter TXb91230 dated July 31, 1991, and provided a schedule for reanalysis.

FSAR Change Request Numbei: 91-141.99 Related SSER Section: SSER23 15.3.8 SER/SSER Impact: Yes The large Break PCT is stated as 2058.5 degrees F.

( 15.6-32 1 Changes to Peak Clad Temperature (PCT) Penalties and 3 Final Limiting PCT for Unit 2 Large Break LOCA Revision:

Errors in the ECCS calculation for Unit 2 result in a total PCT penalty which exceeds 50 degrees Fahrenheit (F). TU Electric notified the NRC that the total PCT penalty exceeded 50 degrees F, in accordance with 10CFR50.46, via letter TXX-91270 dated 'uly 31, 1991, and provided a schedule for reanalysis.

FSAR Change Request Number: 91-141.99 SER/SSER Impact: No 15.6-34 1 Changes to Peak Clad Temperature (PCT) Penalties and Final Limiting PCT for Unit 1 Small Break LOCA Revision:

A correction to the ECCS calc'J1ation has been made to account for zirconium-water reaction, and safety in-jection and auxiliary feedwater flow adjustment. The correction increases the total PCT penalty to 247 de-grees Fa,~enheit (F) and the final limiting small break PCT w dF i degrees F.

FSAR ;.a @ Auguest Number: 91-141.99 Related 55th Section: SSER23 15.3.8 SER/SSER Impu t: Yes The Small Break PCT is stated as 1895.5 degrces F.

15.6-49, 50 2 See Page No(s):51 Add Unit 2 References to the Reference List Addition:

The source materials for the Unit 2 anelyses have been added to the refereace list.

FSAR Change Request Number: 91-141.02 SER/SSER Impact: Ho

I' CPSES FSAR AMEN 0 MENT 85 DETAILED DESCRIPTION Page 5 FSNt Page 1(n imm@d) (Gg EncrhiFD Table ib.6-1 2 See Sheet No(s):01 thru 06 Add Unit 2 Times to Sequences of Events for the Loss of Coolant Accidents and Steam Generator Tube Rupture Addition:

The loss of Coolant Accidents have been analyzed for Unit 2. A new column has been added to Table 15.6-1 for the Unit 2 sequence of events. The times for the Steam Generator Tube Rupture event are identical to Unit 1 because the same analysis is used for both units.

FSAR Change Request Number: 91-141.04 SER/SSER Impact: No Table 15.6-1 3 See Sheet No(s):5 Table 15.6-1, Time Sequence of Events for Decrease in Reactor Coolant Inventory

. Correction:

Incorrect duplicate entries have been removed for the 3 inch break. Duplicate entries for Accumulator Injec- -

tion, Peak Clad Temperature, and Top of Core Covered, stated times which were the times for the 6 inch break.

FSAR Change Request Number: 91-141.99 SER/SSER Impact: No

' Table 15.6-5 2 Add Unit 2 Parameters and Results of Loss of Coolant Accidents to Table 15.6-5 Addition:

Table 15.6-5 summarizes the input parameters for the LOCA analysis. The large and Small Break LOCAs have been analyzed differently for Unit 2. The input pa-rameters for Unit 2 have been added to the Table.

FSAR Change Request Number: 91-141.45 SER/SSER Impact: No Table 15.6-5 3 Table 15,6-5, Input Parameters used in the LOCA Analysis

• Correction:

l The LOCA analyses are run at 3651 megawatts for the

-thermal hydraulic analyses and at 3411 megawatts to determine the cladding heatup.

FSAR Change ..equ?st Number: 91-141.99 SER/SSER Impact: No Table 15.6-6 2 SeeSheetNo(s):01 and 02 Add Unit 2 Parameters and Results for Lerge Break LOCA to Table 15.6-6 Addition:

Table 15,6-6 summarizes the input parameters and re-

=. CPSES FSAR AMEN 0 MENT B5 DETAILED DESCRIPTION Page 6 FSAR Page (as Lmgoded) Grpup Descript100 sults for the large Break LOCA analysis. The Large Break LOCA was analyzed differently for Unit 2. The Unit 2 input parameters and results have been added to the Table.

FSAR Change Request Number: 91-141.45 SER/SSER Impact: No Table 15.6-7 2 Add Unit 2 Small Break LOCA Results to Table 15.6-7 Addition:

Table 15.6-7 summarizes the results for the Small Break LOCA analysit,. The Small Break LOCA was analyzed for Unit 2 with a different methodology. The Unit 2 re-sults have been added to Table 15.6-7.

FSAR Change Request Number: 91-141.45 SER/SSER Impact: No Figure 15.6-6 2 Add Unit 2 Figure for Small Break LOCA Computer Code Addition:

The Unit 2 Small Break LOCA analysis uses the NOTRUMP code. Figure 15.6-6 has been changed to reflect this difference from Unit 1.

FSAR Change Request Number: 91-141.47 SER/SSER Impact: No Figure 15.6-47, (ALB) 2 Update Existing Safety Injection Flow Rate F'igures Revision:

The Safety Injection Flow Rate Figures for the Large and Small Break LOCAs have been updated to account for plant changes to the Residual He.t Removal system.

FSAR Change Requet. i Number: 91-141.47 SER/SSER Impact: No Figure 15.6-49 2 Add Unit 2 Figures for Large Break LOCA Additica:

FSAR Change Request Number: 91-141.47 l SER/SSER Impact: No l

Figure 15.6-49, A 2 Add Unit 2 Figures for Large Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER 1mpact: No

, Figure 15.6-50 2 Add Unit 2 Figures for Large Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER Impact: No

.. CPSES FSAR AMENDMENT 85 DETAILED DESCRIPTION Page 7

'FSAR Page 4 (11AmcDdsd) Group Enscription  !

Figure 15.6-51 2 Add Unit 2 Figures for Large Break LOCA Addition:

FSAR Change Request Number: 91-141.47  !

SER/SSER Impact: No I Figure 15.6-51, A 2 Add Unit 2 Figures for large Break LOCA Addition: 1 FSAR Change Request Number: 91-141.47 l SER/SSER Impact: No l

1 Figure 15.6-S2 2 Add Unit 2 Figures for Large Break LOCA 1 Addition:

FSAR Change Request Number: 91-141.47 SER/SSER 1mpact: No Figure 15.6-52, A 2 Add Unit 2 Figures for Large Break LOCA Addition:

-FSAR Change Request Number: 91-141.47 SER/SSER Impact: No Figure 15.6-53 2 Add Unit 2 Figures for Large Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER Impact: No

' Figure 15.6-54 2 Add Unit _2 Figures for Large Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER impact: No Figure 15.6-54, A 2 Add Unit 2 Figures for Large Break LOCA l Addition:

L FSAR Change Request Number: 91-141.47 SER/SSER Impact: No l

Figure 15.6-55 2 Add Unit 2 Figures for large Break LCCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER Impact: No "

Figure 15.6-56 2 Add Unit 2 Figures for large Break LOCA Addition:

l

Page 8 i CPSES FSAR AMENOMENT 85 DETAILED DESCRIPTION FSAR Page Etnup DnctlRL100 (as amtided)91-141.47 FSAR Change Request Number:

SER/$SER 1mpact: No 2

Add Unit 2 Figures for Large Break LOCA figure 15.6-57 Addition: 91-141.47 FSAR Change Request Number:

SER/SSER 1mpact: No Add Unit 2 Figures for large Break LOCA 2

Figure 15.6-58 Addition: 91-141.47 FSAR Change Request Number:

SER/SSER Impact: No 2

Add Unit ? Figures for large Break LOCA Figure 15.6-59 Addition: 91-141.47 FSAR Change Request Number:

SER/SSER 1mpact: No Add Unit 2 Figures for Large Break LOCA 2

Figure 15.6-60 Addition: 91-141.47 FSAR Change Request Number:

SER/SSER 1mpact: No 2

Add Unit 2 Figures for Large Break LOCA Figure 15.6-61 Addition: 91-141.47 FSAR Change Request Number SER/SSER 1mpact: No 1

l 2

Add Unit 2 Figures for Large Break LOCA l Figure 15.6-62 Addition: 91-141.47 FSAR Change Request Number:

SER/SSER Impact: No 2

Add Unit 2 Figures for Large Break LOCA Figure 15.6-63 Addition: 91-141.47 fSAR Change Request Number:

SER/SSER Impact: No 2

Add Unit 2 Figures for Large Break LOCA Figure 15.6-64 Addition: 91-141.47 FSAR Change Request Nunter:

No SER/SSER Impact:

.. . ...im

1 i

. CPSES FSAR AMENDMENT 85 DETAILED DESCRIPTION Page 9 FSAR Page (as amended) GEDua ECICI.1Rt100 Figure 15.6-65 2 Add Unit 2 Figures for Large Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER Impact: No Figure 15.6-66 2 Add Unit 2 Figures for Large Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER Impact: No Figure 15.6-67 2 Add Unit 2 Figures for Large Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER Impact: No figure 15.6-68 2 Add Unit 2 Figures for Large Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER 1mpact: No Figure 15.6-69 2 Add Unit 2 Figures for Large Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER Impact: No Figure 15.6-70 2 Add Unit 2 Figures for Large Break LOCA

~

Addition:

FSAR Change Request Number: 91-141.47 SER/SSER Impact: No Figure 15.6-71 2 Add Unit 2 Figures for Large Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER Impact: No Figure 15.6-72 2 Add Unit 2 Figures for Large Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER 1mpact: No Figure 15.6-73 2 Add Unit 2 Figures for Large Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER Impact: No

q

( CPSES FSAR AMENDMENT 85 OETAILEO DESCRIPTION Page 10  !

FSAR Page (as Emcoded) GrnVp Descr1otion Figure 15.6-74 2 Add Unit 2 Figures for Large Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER Impact: No Figure 15.6-75 2 Add Unit 2 Figures for Large Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER Impact: No Fi5ure 15.6-76 2 Add Unit 2 Figures for Small Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER Impact: No Figure 15.6-77 2 Add Unit 2 Figures for Small Break LOCA Addition:

FSAR Change Request Number: 91-141,47 SER/SSER Impact: No Figure 15.6-78 2 Add Unit 2 Figures for Small Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER Impact: No Figure 15.6-79 2 Add Unit 2 Figures for Small Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER-Impact: No Figure 15.6-80 2 Add Unit 2 Figures for Small Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER impact: No Figure 15.6-81 2 Add Unit 2 Figures for Small Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER impact: No Figure 15.6-82 2 Add Unit 2 Figures for Small Break LOCA

j. Addition:

y --

t-3 CPSES FSAR AMENDMENT 85 l OETAILED DESCRIPTION Page 11 FSAR Page (Al McDdfd) Gr_Qup Description FSAR Change Request Number: 91-141.47 SER/SSER 1mpact: No Figure 15.6-83 2 Add Unit 2 Figures for Small Break LOCA Additlon:

FSAR Change Request Number: 91-141.47 SER/SSER 1mpact: No Figure 15.6-84 2 Add Unit 2 Figures for Small Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER Impact: No figure 15.6-85 2 Add Unit 2 Figures for Snall Break LOCA l Addition:

l FSAR Change Request Number: 91-141.47 SER/SSER Impact: No figure 15.6-86 2 Add Unit 2 Figures for Small Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER Impact: No Figure 15.6-87 2 Add Unit 2 Figures for Small Break LOCA Addition:

FSAR Change Request Number: 91-141.47 SER/SSER Impact: No Figure 15.6-88 2 Add Unit 2 Figures for Small Break LOCA Addition:

FSAP Change Request Number: 91-141.47 SER/SSER 1mpact: No i

Figure 15.6-89 2 Add Unit 2 Figures for Small Break LOCA l Addition:

! FSAR Change Request Number: 91-141.47 SER/SSER Impact: No l