ML20137D735

Responds to Request for Addl Info Re TS Change Request 96-01 on Conversion to Framatome Cogema Fuel

ML20137D735
Person / Time
Site:	Sequoyah
Issue date:	03/20/1997
From:	Shell R TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-M95144, TAC-M95145, NUDOCS 9703260286
Download: ML20137D735 (32)

Text

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RA Tennessee Valley Authority, Post Office Box 2000. Soddy-Daisy, Tennessee 37379-2000 i

March 20,1997 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:

In the Matter of ) Docket Nos. 50-327 Tennessee Valley Authority ) 50-328 SEQUOYAH NUCLEAR PLANT (SON)- SUPPLEMENTAL RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION - TECHNICAL SPECIFICATION (TS) CHANGE REQUEST 96-01 ON CONVERSION TO FRAMATOME COGEMA FUEL (TAC NOS. M95144 AND M95145)

Reference:

1. TVA letter to NRC dated February 7,1997, on the above subject

2. TVA letter to NRC dated March 17,1997, on the above subject The purpose of this letter is to provide supplemental information to NRC for the TVA j responses to NRC's request for additional information : ? Ripferenne 1. Reference 1 l provided responses to 36 NRC questions. NRC detenmned that no additional information to these responses was necessary with the exception of the responses to Questions 1, 2, 7, 8,11,12,15, 20, 23, 25, 27, 28, 30, 31, 32, and 33.

Reference 2 provided the revised responses to each of these questions with the exception of Question 1. The enclosure provides the revised response to Question 1 as requested by NRC during a meeting between TVA and NRC on February 27,1997, in the NRC Rockville offices.

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ADOCK 05000327 PDR QC3Qjfk 1 200087 )

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. U.S. Nuclear Regulatory Commission Page 3 March 20,1997 i

cc (Enclosure):

Mr. R. W. Hernan, Project Manager Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike ,

Rockville, Maryland 20852-2739 NRC Resident inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy-Daisy, Tennessee 37379-3624 Regional Administrator U.S. Nuclear Regulatory Commission >

Region ll 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323-2711 l

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9 U.S. Nuclear Regulatory Commission Page 2 March 20,1997 Please direct questions concerning this issue to Keith Weller at (423) 843-7527.

Sincerely, .

R. H. Shell Site Licensing and Industry Affairs Manager Enclosure cc: See page 3 l

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. s ENCLOSURE REVISED RESPONSE TO NRC QUESTION 1 i 1

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REOUEST FOR ADDITIONAL INFORMATION TENNESSEE VALLEY AUTHORITY SEOUOYAH NUCLEAR PLANT. UNITS 1 AND 2 j DOCKET NUMBERS 50-327 AND 50-328

1. Revision 2 to the BAW-10168 evaluation model was modified L during the staff review with regard to the Moody break flow i l model and discharge coefficient. Please verify that the j analysis that was used to support this Sequoyah Nuclear Plant (SON) ; fuel change and Technical Specification- (TS) amendment was performed using the approved model.

Response

' Since the time that the small break LOCA analysis of BAW-10220 l was performed, several modifications were made to FCF's small l break LOCA evaluation model, BAW-10168, Revision 2. The changes l were made in response to the NRC review of the document. Chief l l- among these modifications - with respect to the SON small break l L .LOCA analyses - are changes to the break discharge coefficient l and the nodalization of the pump suction piping. The original l evaluation model implemented a variable 0.7/1.0 discharge l l coefficient, the approved model requires the use of a constant l l

1.0 discharge coefficient and the upflow side of the pump suction l piping is more finely divided to better predict the occurrence of l loop seal clearing. l l

As a result of FCF's efforts to bring the small break analysis l

! into compliance with NRC-approved methodology, additional l l problems with leak node modeling were identified and resolved l (see the response to NRC question number 8 in this set). The l small break spectrum was rerun using consistent homogeneous leak l junction inputs, a nonequilibrium fictitious leak control volume, l and the containment and leak volume areas set equal to the-cold l leg pipe area. l l

Analysisoof a spectrum of small breaks predict a modest clad l temperature excursion, approximately 1162 F, for the limiting, 'l 2.75-inch, break case. The work fully complies with the l l NRC-approved SBLOCA evaluation model, as documented in BAW-10168, l Volume II. No evaluation model exceptions have been taken and no l evaluation model difficulties were encountered in the process of l analyzing this event. -l

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5.9'.2 Small Break LOCA Evaluation Model l l

RELAPS is used to predict the reactor coolant system l thermal-hydraulic responses to a small break LOCA. The code l has been approved by the NRC for licensing application and l t

is documented in detail in Reference 5-3. RCS nodalization l l is based on the model described in Volume II of the NRC- l l approved BWNT RSG LOCA EM, Reference 5-1. Nodal diagrams of l the SON small break LOCA model are presented in Figures 5.9- l 1 and 5.9-2. The small break LOCA model is similar to that l l used in the large break analyses. l l

The reactor core is divided radially into two regions l similar to that of the LBLOCA model; one region represents l the hot fuel assembly and the other represents the remainder l of the core. The core is further divided into twenty axial l segments. Cross-flow junctions connect hot assembly fluid l ;

l nodes with the adjacent " average" assembly nodes. This  ;

arrangement allows the computation of hot assembly cladding l and vapor temperatures with limited influence by coolant l from the average core and provides resolution of the mixture l level to within approximately 0.5 foot. Initial fuel pin l parameters are calculated with the TACO 3 computer code l (Reference 5-6). The reactor vessel downcomer and upper l l plenum regions are represented in finer axial detail than l those of the large break LOCA model to give a better l representation of the void distribution that affects the l l system hydrostatic balance. l

! l In the small. break LOCA model, the RCS is subdivided into i two flow loops. One loop represents the broken. loop and the l other represents the three intact loops. The pressurizer is l attached to the hot leg of the composite intact loop. The l

nodalization is similar to that of the large break LOCA l model. l l

The steam generator tube region is divided into two radial  ;

regions. One region represents the shortest half of the l tubes and the other region represents the remainder of the l tubes. This provides sufficient modeling accuracy to  !

simulate tt i draining effects; tube draining can be l sensitive t tube length. l

! l l The reactor coolant pump suction nodalization has been l l altered relative to the large break LOCA model to produce an  !

accurate hydrodynamic representation of loop seal clearing. l Two additional nodes are added to the downside of the pump l suction pipe and three nodes to the riser section. This  ;

allows finer resolution of the void distributions and l FCF Non-Proprietary

elevation heads that control the occurrence and timing of l loop seal clearing. l l

The bottom elevation of the lowest node of the intact loop  ;

pump suction piping is artificially extended one foot below l the corresponding node in the broken loop. This l preferentially promotes the clearing of only the broken l loop. An RCS configuration characterized by a single clear j loop and three intact loops conservatively restricts steam  ;

flow to the break. The added restriction can result in l worsened core conditions and the potential uncovering of the  !

core. l l

Both the broken loop and the intact loop reactor coolant l pump discharge piping are modeled as four nodes. In the l large break model, the intact loop is modeled as one node. l Using four nodes provides an accurate simulation of the l hydrodynamic effects of the ECCS injection. l l

The computer code options and generic input requirements l used in the small break portion of the BWNT RSG LOCA EM are j summarized in Volume II, Tables 9-1 and 9-2, of Reference 5- l

1. Consistent with the RCS loop modeling, the phase non- l equilibrium option is selected for the artificial leak l volume, node 276. The break path is treated homogeneously  !

for both critical and non-critical break flow predictions. l In addition, the volume area of the break mass sink volume  !

is set equal to the volume area of the artificial leak node l to preclude the development of any contribution to the l momentum flux gradient at the break junction. l l i The small break LOCA model is fully compliant with the NRC-  ! )

approved guidelines of the BWNT RSG LOCA EM established in l Reference 5-1. It has been developed utilizing the RELAPS l l large break LOCA model, described in Section 5.2 of this  ! ;

report, as a basis. The small break model adheres to the l requirements of 10CFR50 Appendix K and contains demonstrated l conservatism for the evaluation of ECCS mitigation of a l postulated small break at SQN. l l

5.9.3 Inputs and Assumotions l l

The major plant operating parameters used in the SQN small l break LOCA analyses follow. These inputs are similar to l those utilized in the large break analyses. l

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1. Power Level - The plant is assumed to be operating l in steady-state at 3479 MWt (102% of 3411 MWt). l l

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2. Total System Flow - The initial Reactor Coolant l System (RCS) flow is 348,000 gpm. l l .

3. Fuel Parameters - The initial fuel pin parameters l are taken from TACO 3 (Reference 5-6) runs l performed for BOL fuel conditions. j  ?

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4. ECCS - The ECCS flows are based on the assumption i of a single active failure. A single train of l ECCS is modeled as described in Volume II, Section l )

4.3.2.2, of the BWNT RSG LOCA EM, Reference 5-1.  !

For the case of a centrifugal charging line break, l charging flow is assumed to be spilled to the l l containment. l l l

5. Total Peaking Factor (F o ) - The maximum total l j

peaking factor assumed by this analysis is 2.5. l ,

The hot assembly peaking for small break analysis l )

is illustrated in Figure 5.9-3. l l

6. The moderator density reactivity coefficient is l based on BOL conditions to minimize negative l reactivity. l

7. The, cladding rupture model is based on NUREG-0630. l ,

I l 5.9.4 Analysis Results l l

l In the SQN small break LOCA analyses, seven break cases were l considered independently to predict core and system l  ;

responses over a spectrum of break sizes. Small break l spectrum results in Volume II, Appendix A, Reference 5-1, l indicate that break areas corresponding to 2- to 6-inch l  !

diameters produce the most severe core depression. Breaks l l of 1.5 , 2.5 , 2.75 , 3.0 , 3.25 , and 5-inch diameters in l the bottom of the reactor coolant pump discharge piping were l analyzed for SON. In addition, a 1.34-inch diameter l centrifugal charging line break, located in the top of the l piping, was analyzed. l l

Table 5.9-1 presents time sequence of events for each of the l small break LOCA cases. Fuel thermal responses for the hot l pin are included in Table 5.9-2. Parameters of interest to l j

the small break analyses are shown in Figures 5.9-4 through l l

5.9-28. There are seven sets of figures, each set contains l five plots. The five figures of each set show (1) the RCS l pressure, (2) the break flow rate, (3) loop seal levels in l the pump suction downflow and upflow pipes, (4) core  !

collapsed level, and (5) hot spot cladding temperature. l FCF Non-Proprietary

A 'relatively slow depressurization rate occurs in the 2.75- l inch break case. The core does uncover, making the 2.75- l inch break the most limiting case of the small break l spectrum analysis. The resulting peak cladding temperature j (PCT) is 1162 F. Core metal-water reaction for the 2.75- l inch break is negligible because the cladding oxidation rate  ;

is not significant below about 1500 F. l l

For breaks smaller than 2.75 inches in diameter, core l cooling is maintained by a combination of steam relief at l the break and reflux cooling in the steam generator. The l core does not uncover for these smaller breaks. For breaks l larger than 3 inches in diameter, the rapid depressurization l rate following loop seal clearing has two positive effects  ;

on the core level. One is increased ECCS flow, and the l other is increased core level swell. No core heatup was l predicted for these break sizes. l l

The centrifugal charging injection line break is postulated l to allow examination of a small break LOCA that is l characterized by a degradation of high pressure injection. l  !

The break size is insufficient to allow significant l >

depressurization of the RCS. Coolant addition in the l progression of the transient is, therefore, governed by the l high pressure injection alone. With a broken injection l line, a large portion of the ECCS flow associated with the l centrifugal charging flow is directed to the break. To  !

ensure a conservative result, all of the charging flow is l assumed lost to the break, and the transient is mitigated by j safety injection pumps only. The results of the charging l i line break indicate that the core remains covered by the  !

mixture level and that no core heatup occurs. l l

The small break analyses are terminated when the break flow l rate is exceeded by the ECCS flow rate. Note that the l collapsed liquid level at the end of the transient may still l be below the top of the core. The core mixture levels at i the end of the analyses are, however, above the top of the l active core and the RCS pressure is still falling. A steady l increase in ECCS injection and continued core cooling is l therefore assured. l l \

5.9.5 Comoliance to 10CFR50.46 l l

l The small break calculations directly demonstrate compliance l to two of the criteria of 10CFR50.46 and serve as the basis -l for demonstrating compliance with two others. As seen in l the figures and in Table 5.9-2, the highest peak cladding l

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l temperature, 1162 F, and the highest local oxidation, about l i

0.004%, are well below the 2200 F and 17% criteria. l  ;

I The whole-core oxidation criterion of 1% cladding reaction l t is met as well in_the small break LOCA analyses. Whole-core l l oxidation will be much'less than the peak local oxidation l l figure of 0.004%. Whole-core oxidation associated with l  ;

small break LOCAs, utilizing the assumptions and inputs as l  !

documented above, is negligible. l l I

The. fourth acceptance criterion of 10CFR50.46 states that l l calculated changes in core geometry shall be such that the l  !

core remains amenable to cooling. The calculations in l  !

Section 5.9 directly assess the alterations in core geometry l  !

that result from the LOCA, at the most severe location in l  !

the core. These calculations demonstrate that the fuel pin l  !

cooled successfully. Further, for SQN, no hot assembly l i cladding ruptures occurred in any of the small break LOCA l j

. cases. Therefore, .the assembly retains its pin-coolant l  ;

channel-pin-coolant channel arrangement and is capable of l being cooled. l l

The.fifth acceptance criterion of 10CFR50.46 states that the  ! ,

calculated core temperature shall b2 maintained at an l  !

acceptably low value, and decay heat shall be removed for l j the extended period of time required by the long-lived  ;

i radioactivity remaining in the core. Successful initial  !

operation of the ECCS is shown by demonstrating that the l ,

, core is quenched and the cladding temperature is returned to l i near Saturation temperature. l  !

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Compliance to the long-term cooling criterion is l demonstrated for the systems and components' specific to SQN 'l  ;

in the FSAR and is not related to the fuel design. The l

' initial phase of core cooling has been shown to result in l

-low cladding and fuel temperatures. A pumped ir.jection. l system capable.of recirculation is available and operated by l the plant to provide extended coolant injection. Therefore, !. '

compliance with the long-term cooling criterion of l  !

10CFR50.46 has been demonstrated. l

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. Table 5.9-1 Small Break LOCA Time Sequence of Events l j i

i 1.5-Inch 2.5-Inch 2.75- 3.0-Inch 3.25- 5.0-Inch CCIIine l Events, Seconds Break Break inch Break Inch Break Breuk l Bnak Break l \

Break Initsatson 0.0 0.0 0.0 0.0 0.0 0.0 0.0 l Reactor Scrum 91.5 33.4 27.6 23.4 20.2 10.2 115.1 l RC Phmp Coastdown 91.5 33.4 27.6- 23.4 20.2 10.2 115.1 l MSlY Closed 91.5 33.4 27.6 23.4 20.2 10.2 115.1 l 51 Signal 104.6 40.7 34.0 28.6 24.4 10.1 129,7 j MFWIsolation 101.5 43.4 37.6 33.4 30.2 20.2 125.1 l Psmped ECCS Iqjection 141.6 77.7 71.0 65.6 61.4 47.1 166.7 j Loop Seal Clearing NA 1267.8 989.5 784.9 636.1 247.3 NA l Top of Core Uncovers NA NA 2370 2070 1690 210 NA l Peak Cladding NA NA 2914 2295 NA NA NA l Tensperature l Accumulator Iqjection NA NA .2895 2395 1830 490 -NA l l

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Results 2.75-Inch 3-Inch Break l Break l Peak Cladding Temperature,"F 1162 828 l Peak Temperature location, ft 10.9 l l .6 l Rupture Time, sec NA NA l.

Rupture location, ft NA NA l Maximum local M/W Reaction, % ~D.004 -D l Total M/W Reaction, % . <0.004 ~D l l

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i 5v

' ' ..c

• E, ,

' e l

15- +

, i 9 l z

% l V

-20 O 500 1000 1500 2000 2500 3000 35J0 4000 A TIME, S

• Figure 5.9-17 2.75-inch Pump Discharge Break Core Collapsed Level 16 ,

4 i i i  !  ! ,

jf. ..4-- , 4... . t i

j i 3

j 12- +- 4-t-

i--

i '

i k 10- --

t- t-- i y-8- --

j- .

h i

3 6- t-i f- -

i.  ! i a

4- + 4-r-- -L- -- j  !.- .

i

! , l

. l 2- -

' +- - - - - - - -

, i 0 >

0 500 1000 1500 2000 2500 3000 3500- 4000 TIME, S Figure 5.9-18 2.75-inch Pump Discharge Break Hot R0d Clad Temperature 2400 I i i  !

1 .

1 I l 2000- 4- J- - -

-4 -

i  !

.! I i 1600- +- + -~- ,

i

-a a- I I1200-

+- F  ;-

i i i  ! ,

i  !

(

2 800- ---

5 L -

c i i 400- l +- '

- ' a- , . . - - . - $

i i e c

O i Z O U 0 500 1000 1500 2000 2500 3000 3500 4000 W TIME, S

4

.' Figure 5.9-19 3-inch Pump Discharge Break Primary System Pressure 2400 , .

i  !

I 2000- +- - - -

+- -4

-i-l ,

i 1600-  ; .

4 - -- -

t-- r-l  !  !

E  ! i i

! t 1200- --

4 y- y  ? -

i 800- r- l -- - A -- 4-i' i

400- +- .,

+-- -;-

I 1

O l 0 500 1000 1500 2000 2500 3000 3500 4000 TIME, S i

l  !

l i

Figure 5.9-20 3-inch Pump Discharge Break Leak Flow Rate 1600 i  ! ,

2

, i jf . . . . . . . . . - - . . . . . . . . . . ..u. , . . . . . .

I  ! .

t i e i  ! i 1200- r += 4- -- e i-E  !

j I i 8

f 1000-  ;

e-

.N  !

$ 800-

+

I i i- --

t i

i 3:  :

i o i ,

j  ! -

m 600- - +- 4- +-- 4- F- +-

e 4 g 400- - - -

+-

i i -

]-

u 2

l i A 200- -

h i

i

, W -

0 V O 500 1000 1500 2000 2500 3000 3500 4000 A TIME, S

Figure 5.9-21 3-inch Pump Dischargle Break Pump Suction Loop Seal Leve s 20 LEGEND 15 .". INTACT LOOP SEAL DOWN,

, BROKEN LOOP SEAL DOWN O- INTACT LOOP SEAL UP 10- ......._._;. +. ------ BROKEN LOOP SEAL UP 5- -- =

i- -

[-

.i 4-30

"-- +- - '- -

20 0 500 1000 1500 2000 2500 3000 3500 4000 TIME, S Figure 5.9-22 SequOyah 3-inch Pump Discharge Break Core Collapsed Level 16 ,

i 14-  :

• * < > - < +

12- 4 -

j

?

k 10- i~ --

8- -' - -

0 3 6- 4 -

- ' - --- * "+" * -

i b

4 . . .

c

. 4 8

' O.

2- t- - -

h Z

w 0 U O 500 1000 1500 2000 2500 3000 3500 4000 W TIME, S

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

5 Figure 5.9-23 3-inch Pump Discharge Break Hot Rod Clad Temperature 2400 1

j  !

i t 2000- -t  !-

f' - 1 -

1800- b, -- 4 ,- , -- -

- r -

. i i

'1200- r=- -

• I 800- 1 -+-  ;-

i

( i e 9he^us- tme De- tae i aus lie d e g$t e e -

i i  !

i' i

0 O 500 1000 1500 2000 2500 3000 3500 4000 TIME, S Figure 5.9-24 3.25-inch Pump Discharge Break Primary System Pressure i

2400 '

I f i  : i I  !

1

! 1 i l l 2000- -+--

r--

i

---

-f -f- 1 i'

i i

1600- 7- 4- --

7-i  !

E l 1200- -j -

+- ,

i i 800- ' - - - - - - - - -

• i-- '

u c

.  : i O.

400-

+  ;

+--

+ -- t-2 t, - s A,

! c i o i

Z 0 U 0 500 1000 1500 2000 2500 3000 3500 4000 W TIME, S

4

, Figure 5.9-25 3.25-inch Pump Discharge Break I Leak FIOw Rate l 1600 i

?

i 1400< -

N- - -

+-

! i i  !

1200- -l-' f -- l-e  ;

1000- - - -

4- - k- -- '

g i i  !

i 5 800- -+. L-

{-

m 600-

\/i

\./ i-

+- +-

i M i I

400- -~+ 4- ,

I 200 ...L. . . . . . . , .

w 0

0 500 1000 1500 2000 2500 3000 3500 4000 TIME, S Figure 5.9-26 3.25-inch Pump Discharge Break Pump Suction Loop Seal Levels 20 LEGEND 15- --. ,  ! , INTACT LOOP SEAL DOWN. ,

BROKEN LOOP SEAL DOWN )'

i -

O- INTACT LOOP SEAL UP 10, BROKEN LOOP SEAL UP tC 5- -- 1 d  ! i b O ;O0000 ^^

g 5- --..

.+.. -~

.+..

~v - e -- -

'C i i i i 4

s@

4-j- +- ,

g Z

-20 w

0 0 500 1000 1500 2000 2500 3000 3500 4000 A TIME, S

,' Figure 5.9-27 3.25-inch Pump Discharge Break Core Collapsed Level 16 + ,  ;

~

I t

i i 14- - - -

r-- --~~- < ~  ;  !

i . i.

12< -r 4-- 4

---

f e-k 10- t-

+- t-i i

8- - -

h  ;

l I i

3 6- t-E 4

i t- 4 -;

i i f i I I . I

,i 4 .i. . e. .. j.: 4.- . .

k i f

. I 2- - - -- + --

r- -+

3

,  ! . I-

! i t i i ,  ; e 0

O 500 1000 1500 2000 2500 3000 3500 4000 TIME, S i

1 l

Figure 5.9-28 3.25-inch Pump Discharge Break Hot Rod Clad Temperature 2400 i

! l 2000- - -- 1- -

j-  !

l i I

! .  : I

,1600- --  ?- e f-- +- -i- -*~

i i

! i

]

+

l1200' 800- + 4- -l- l- - -- - + &'

V L c

c.

t 2

-400-  :

- + - -i - - <

% 1

! s i O

Z 0 0 0 500 1000 1500 2000 2500 3000 3500 4000 %

TIME, S

,' Figure 5.9-29 5-inch Pump Discharge Break i Primary System Pressure 2400 .

t i i i  ;

i I I .  ! .

2000- '

+- '

+- --

1 i  !

l+

! i i i 1600- l i- -- f- p- 4- +- -

i i i i. i 1200< -

v-- h--- *-

l 800- +---- t --

l l i i

l 400- r- -r- - i~ d~ +

l i

i

~

i i l j '

i l 0  ;

! O 500 1000 1500 2000 2500 3000 3500 4000 4

! i l TIME, S Figure 5.9-30 5-inch Pump Discharge Break '

t Leak Flow Rate 1600 ,

i  !

j f,4Q . .. . .

.7 , . - . . .

4... . . . .

t l

: i 1200- -?-- 4- - -

-i-E J-( i i i 1000- +- 6

! +

i i i 800-

• is l l 0 i  ! 2 I

j

, 600- -t- +- e-- -+-- -i- -+-- - - -

2 l

i I

k ti 400- - --

-r - - - --

'Q

- }-

200, 4 ...- . . . ..- .. e

• s O

i Z

m 4

0 V O 500 1000 1500 2000 2500 3000 3500 4000 W TIME, S

^ '

• ^ * ' '- '

Figure 5.9-31 5-inch Pump Discharge Break Pump Suction Loop Seal Leve s 20 i LEGEND i

15- -t---- 0 INTACT LOOP SEAL DOWN.

, BROKEN LOOP SEAL DOWN j -

0--- INTACT LOOP SEAL UP 10-  ;- r {- ------ BROKEN LOOP SEAL UP i

I I  !

t 5- - -r-i -

i- .- --}  :-- i -..

.r,

i  !

i 3 .

Og

--t- '

- 4- y -- 4-  !.

l

. i  !

.s. . . .

q._. 3 ..

30 ., (4 _ _-.- .... . ...

+

tra !  !

i

., s . - .+-> s

. . - . . . . - -+- 1 i i i i' l 1

~

20 0 500 1000 1500 2000 2500 3000 3500 4000 TIME. S Figure 5.9-32 5-inch Puma Discharge Break Core Collapsec Level 16

!. i

! l-14- -- -- !- - -k - ,

{ -- - ? - ---4 ---+ -

I  !

}  !  !  !

1 i i 12- -+-  ?-- r -4 i- i i i i i i'

i  ! i k 10-

?-

?

4-I 5 8- --

-y a- 1 +

6- - -

--r -; - - T-  ;---

• 5 4 ....{.... .. ... _

7 g

2- +--

-i - t- i -- -

A

o I

Z i i i i m 0

O O 500 1000 1500 2000 2500 3000 3500 4000 W TIME, S

t Figure 5.9-33 5-inch Pump Discharge Break Hot Rod Clad Temperature 2400 .

i"  !

2000- i- - --  ? - -- --- 1 i i i

1 i

jggg. . . . . . _ , . -

3-....... [. . ... . . ~ . . .

u. j  ;

! 1  :

1200- -----

i --

1 i

!  !  ! i 800- i- 4 -+ 4-- 4 -- -

. i t i 8'. I

- i I 400- -b-- +- +- -

-r -

---+--- ,

i I I  !

l i

i i 0

O 500 1000 1500 2000 2500 3000 3500 4000 TIME, S Figure 5.9-34 CCI Line Break Primary System Pressure 2400 i i i I  !

2  !

l 1

2000- --

+

i-2 I

1600- - +- - - -

I

. i ,

I 1200- - c-- '--

i i

E i i 800- -  !--

i

' - ~+-

t -

(

cc .

I O

400 - " - - - ~ -

-- +- -

5 i b

O Z

O N

0 U

1000 2000 3000 4000 5000 6000 m TIME, S

a

• I* Figure 5.9-35 CCI Line Break Leak Flow Rate 1600

+

1 r  :

i ,

1400- 4- l f-1200- + '-

10 1000- +-- ~~ r -- - ----+-

. I I 800- 4 --  ? -- --+ +

d i 600 . .. -- 4 1

400- 7 r +-

l ii

5 200- f -- -j- f-T 0

0 1000 2000 3000 4000 5000 6000 TIME, S Figure 5.9-36 CCI Line Break Pump Suction Loop Seal Levels 20 LEGEND 15- -

4 --

A INTACT LOOP SEAL DOWN.

, BROKEN LOOP SEAL DOWN j -

G- INTACT LOOP SEAL UP 10- '

4 ------ BROKEN LOOP SEAL UP

. LLL LLLLL LL _ m i

. i N i . . i  %

C 5- -e-  ;-

d ,! ,

b OM 0 0 0 0 0 0 0 0 0 0 0 0; O O O i [0 0 0 f .s. +-- ~ 7..

.p

(

i Y

c:

c.

i

: 2 1 ,

m

--? - r- t &

o l i 2

20 0 1000 2000 3000 4000 5000 6000 p.,

TIME, S 3

i 4

• #**~*

Figure 5.9-37 CCI Line Break l Core Collapsed Level l

16 . .

i  !

! I i

! 2 14- 4' -- -+ b - * -4 i  : I a

i i  ! i

- i-12- b-i k. 10- - - ;- - - - -t-t-

- -t t --

i 8- .  ? *-

i 1

! h l 3

6- 1 -I- 4- I-l  !  !

.  !  ! i l 4-  ! f 4- r-

+-

i i i i

,.  ! t j ,

4  : i

' e -- -

2- ,- 4 i

+ - - -

. I l

0 O 2 4 6 8 10 12 14 16  ;

c c:~ - - TIME, S 1

Figure 5.9-38 CCI Line Break  !

Hot Rod Clad Temperature 2400

. i I

i

. i t

., i i  !

1 2000- ----i- -- ---!- b- i- - -

i  ! l l

i 1- -+-

• 600-1 - - --

1 I

N i I i I1200-

&4 --

I ,

i  !

800- * * <

5

(

% 'co.

o w

4%. ..i.. .

.p. .. . . . . . . <. . 4

' c:

' O Z

, W O U 0 1000 2000 3000 4000 5000 6000 W TIME, S i

I

l

• <c,f*

5.'12 References l l

l 5-1 BAW-10168P Revision 02, BWNT Loss-of-Coolant Accident l l Evaluation Model for Recirculating Steam Generator l l Plants, October 1992. l l

5-2 BAW-10168P Revision 03, BWNT Loss-of-Coolant Accident l Evaluation Model for Recirculating Steam Generator l Plants, November 1993. l l

5-3 BAW-10164P Revision 03, RELAPS/ MOD 2-B&W - An l Advanced Computer Program for Light Water Reactor l LOCA and Non-LOCA Transient Analysis, October l 1992. l l

5-4 BAW-10171P Revision 02, REFLOD3B - Model for Multinode l Core Reflooding Analysis, January 1989. l l

5-5 BAW-10166P Revision 04, BEACH - A Computer Program for l Reflood Heat Transfer During LOCA, October 1992. l l

5-6 BAW-10162P, TACO 3 - Fuel Pin Thermal Analysis l Code, October 1989. l l

5-7 BAW-10092P, CRAFT 2 - FORTRAN Program for Digital l Simulation of a Multinode Reactor Plant During Loss-of- l Coolant, April 1997. l l

5-8 BAW-10174P Revision 1, Mark-BW Reload LOCA Analysis for l the Catawba and McGuire Units, September 1992. l l

5-9 V. H. Ransom et al., RELAP5/ MOD 2 Code Manual, Volumes 1 l and 2, NUREG/CR-4312, EGG-2396, 8/85. l l

5-10 BAW-10177P, Mark-BW Reload LOCA Analysis for the l Trojan Plant, October 1990. l l

5-11 BAW-10184P, GDTACO - Urania Gadolinia Fuel Pin Thermal l Analysis Code, February 1995. l l

5-12 BAW-10172P, Mark-BW Mechanical Design Report, July l 1988. l l

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