ML20205T603

Comparison of Std Westinghouse PWR & McGuire Nuclear Station After Deletion of Upper-Head Injection Sys Re Expected Results of Large Break Loca

ML20205T603
Person / Time
Site:	Mcguire, McGuire
Issue date:	06/30/1986
From:	DUKE POWER CO.
To:
Shared Package
ML20205T600	List: ML20205T596 ML20205T603
References
NUDOCS 8606160097
Download: ML20205T603 (44)

Text

. _ .

I' l

l i

g A COMPARISON OF A STANDARD W PWR AND THE McGUIRE NUCLEAR STATION AFTER DELETION OF THE UPPER READ INJECTION SYSTEM IN REGARD TO THE EXPECTED RESULTS OF A LARGE BREAK LOSS OF COOLANT ACCIDENT DUKE POWER COMPANY JUNE 1986 is kDR fyCK05000369 97 860609 P

PDR  !

e

. e e .

I) INTRODUCTION Upper Head Injection (UNI) was added to the design of the McGuire Nuclear Station in the mid-1970's in order to compensate for the negative impact of the ice condenser containment upon the ECCS Evaluation Model. The UHI System was designed to provide additional core cooling during an accident involving a large break of Reactor Coolant System cold-leg piping. The additional cooling '

was expected to offset the slower reflood of the. core associated with the low containment pressure and reduced steam venting rate which result from the ice condenser design. Without UHI, the reduced reflood rate and very conservative treatment of the reflood portion of a LOCA by the Evaluation Model which existed at the time McGuire was being constructed would have resulted in more restrictive operating limitations for ice condenser containment plants.

Research and development sponsored by the NRC and the nuclear industry since the 1970's have resulted in a better understanding of nuclear plant behavior during a loss of coolant accident. The application of this knowledge has led to improved ECCS evaluation models and "best estimate" models. The improvements in prediction capability tended to demonstrate that UHI was less beneficial in the mitigation of a LOCA than may have been thought initially. The reduced observed benefit of UHI in limiting the calculated peak clad temperatures coupled with the many operational problems and delays experienced at McGuire related to the UHI System led Duke Power Company to investigate the possibility of removing the system.

In order to determine if UHI could be removed without adverse impact on plant safety or the requirement of overly restrictive operating limits, a scoping analysis was completed by Westinghouse using the BART Evaluation Model in late 1984. Results of the scoping BART analysis demonstrated that regulatory requirements could be satisfied and reasonable operating restrictions could be expected with the improved analytical models and deletion of UHI. Meetings were held in early 1985 with the NRC staff and ACRS ECCS Subcommittee to discuss the proposed deletion. Subsequently, Duke decided to proceed with the project and contracted Westinghouse to perform the required safety analyses.

The Westinghouse analyses demonstrated that the large break LOCA and other affected safety analyses continued to satisfy all safety and regulatory requirements after deletion of UHI. Reports submitted to the NRC in October 1985 and March 1986 provide the detailed results of these analyses. Also addressed in the March 1986 submittal is a comparison of results of analyses performed assuming UHI operable and UHI deleted and an explanation of the differences.

The NRC contracted Sandia National Laboratories to perform additional analysis with the "best estimate" TRAC computer code. The results of the Sandia evalu-ation were presented to the ACRS ECCS Subcommittee in March 1986. Comparisons of the Sandia TRAC results for McGuire with other TRAC analyses for standard Westinghouse dry containment plants showed that the predicted peak clad temperatures were similar; however, there were differences in the calculated clad temperature response during reflood. In its April 16, 1986 letter, the ACRS concurred with Dcke's proposal to operate McGuire without the UHI System. The ACRS also suggested that further calculations be considered to show the operating conservatisms associated with McGuire are similar to non-UNI units before the UHI System is removed in an irreversible manner.

4

O e * *.

This report will use the existing analyses and data from various calculations and experiments to demonstrate that the response of McGuire without UHI to a large break LOCA is similar to non-UHI plants. Parameters such as peak clad temperatures are provided for various cases and are adequate to show that the operating conservatisms of McGuire without UHI are comparable to non-UHI plants.

5 i

0 0

II) WESTINGHOUSE EVALUATION MODEL ANALYSES The nature of the ECCS Evaluation Models and the regulatory requirements, guidelines, and criteria which govern their use are intended to ensure plants of various designs are operated in such a manner as to satisfy minimal safety requirements. The fact that the McGuire analyses performed to support UHI removal used NRC approved methodologies which have also been applied to non-UHI plants and the results are compared to the same acceptance criteria I supports the conclusion that similar levels of conservatisms exist at McGuire and non-UHI plants.

The required conservatisms in the Evaluation Model analysis result in predictions of peak clad temperatures occurring during the reflood portion of the LOCA for all types of plants. "Best estimate" and experimental results show that peak clad temperatures probably occur during the blowdown phase of the LOCA. The "best estimate" and experimental results are discussed in more detail in the later sections of this report. Comparisons of McGuire without UHI and non-UHI plants using Evaluation Model calculations are provided in the remainder of this section.

Although the UHI System will be removed from McGuire and some plant changes will be made which result in McGuire more closely resembling non-UHI plants, several significant differences will remain. The major differences between McGuire and non-UHI plants are:

1) Ice Condenser Containment

2) Increased Upper Head / Core Flow Paths

3) Increased Upper Head /Downcomer Flow Paths The Ice Condenser Containment results in a lower back pressure against the Reactor Coolant System during the LOCA than does the dry containments of most Westinghouse plants. The reduced back pressure reduces the rate at which steam is vented out of the break and thus reduces the refill /reflood rates which determine the clad temperature predictions. Therefore, the Ice Condenser Containment is a factor which, if all other parameters were equal, would tend to result in higher clad temperatures predicted for McGuire thar.

would be predicted for dry containment (non-UHI) plants.

The control rod guide tubes and support columns between the upper core plate and support plate combine to form a larger flow path between the upper head and the reactor core at McGuire than exists at plants with standard (non-UHI) vessel internals (Figures II-5 and II-6). The increased flow communication which exists with the UHI internals produces enhanced thermal-hydraulic conditions within the core during a large break LOCA blowdown. Figures II-1 and II-2 show the core flows during blowdown for McGuire without UHI and a plant with standard vessel internals. The core flows are similar for the first few seconds, but from five seconds onward the UHI internals are clearly beneficial. Between 5-10 seconds the UHI internals provide a greater water delivery into the upper plenum which produces a notably higher positive core flow rate for McGuire. Likewise, at around 20 seconds, the enhanced water delivery from the upper head at McGuire permits a much greater negative core flow surge than is exhibited by the standard plant. These greater core mass flow rates result in a significant benefit in calculated peak .

clad temperatures for McGuire relative to the standard plant at end of blowdown.

O The increased flow area between the upper head and downcomer was added to McGuire to ensure the upper head region temperatures -emained at cold-leg values during normal operation and thus reduce the conce*n related to the flashing of the water in the upper head impeding the injection of UHI. The bypass flow to the upper head (Figures II-5 and II-6) was increaseed from approximately 1% of the total RCS flow to approximately 4% of the total RCS flow. During a LOCA, this flow area provides a path for steam from the core to flow directly to the downcomer, thereby allowing for an increase in the refill /reflood rates. Therefore, the larger flow area associated with the upper head bypass paths is beneficial with respect to the plant response during a LOCA.

An approximation of the integrated effects of the above factors are provided in Figures II-3 and II-4. As shown, the UHI vessel internals result in a lower clad temperature at the end of blowdown. The blowdown peak temperature is about the same for the standard and UHI internals analyses. Figure II-4 shows that the case with UHI internals experiences end of blowdown and beginning of refill /reflood several seconds later than the standard plant.

The slight delay is due to the lower containment pressure and water from the upper head region entering the core and maintaining a slightly higher pressure in the Reactor Coolant System. The extension in the blowdown period is thus attributed to the same factors previously discussed and those whose net effect is a lower clad temperature at the beginning of refill /reflood. The heatup of the clad during refill /reflood is similar for the UHI internals and standard cases but the earlier stabilization for the UHI internals analysis is partly due to the BART model.

The comparison of McGuire without UHI and a standard Westinghouse plant using Evaluation Model analyses shows the results are similar. The operating margins for McGuire without UHI are comparable or perhaps slightly higher than non-UHI plants using the conservative Evaluation Model calculations.

o 7000 5000 -

2500 -

n  ; h \\ -

v < e I

(

N

-2500 -

-5000 I

~7UUU o lo 20 30 40 W TILC (SEC)

Figure II-1: Core Flowrate with UHI Internals (CD"

• 7000 I

i 5000 -

o tu u

2500 -

J TOP

. W O 4 t ,

E 3

u.

-2500 ---

DOTTOM tb

-5000 -

I  !  !  !

-7000 O 10 20 30 10 11 0 TIMLI (SEC)

Figure II-2: Core Flowrate with Standard Internals (C D "

• t nm.o FOUR-LOOP PLANT CLAD TEMPERATURE TRAfiS!ENTS i TilR00Gil BLOWDOWN:5.75 FI. CORE ELEVATI0ri E imo.o ,

E n.

om.o h

[

iLS l E " * ~#,

, - I soa.co . _ _ . _ . . _ _ . _ _ . . . . . . _ . . _ _ . _

0.0 ,

o

.:  : a n .sie.

F-Figure II-3: Comparison of Peak' Clad Temperatures During a LBLOCA for a Standard (non-UllI) Plant and McGuire Without UllI

noo.o -

RESAR-1 i, 1981 MODEL

- acm.o , _

~

MODIFIED lHI, BART tlih

3500.o ---

5 S

e - .

Y, i o sooo.o y

Soo.co .. ....

PfAK CLAD TEf1'ERAIURE

', C0fiPARIE0rl

>! 1!  :: Si ',i

<: 1:  :: u-  ::

o - .a. .- ,,

l'lMt. (SLC) i Figure II-4: Comparison of Peak Clad Temperatures During a LBLOCA for a Standard (non-UHI) Plant and McGuire Without UHI i

u m.

N

in .

I li p; -

I eel n n

-~~

F, * . 'i

\ i ,i.) 1,' c:a

)l ,tj*), k ' ~ ' ~

! i r- O

}h ;g n l= T' m ss ;

r ll ,

lii i F' -

l ' A. =. Ml t.1

%& o. w- _m ti.\

ik J _M M +l "

O 9}y , k hidiM&&<

_u

> ,F a,! ,.9 t R _

g Bypass Flow

$ $  ! dl, R Nozzle m == m -._

[. ,

, y y' Dt "- '

/ 9 9 1 -

i

~ .... .. , /;

%, , +, , _

i.

1. , .A , - s y

a t u ....

e axx,q a i ,

k[ bk[ bk1A 1 .

xh f "ontrol C

" ' " " ' "Rod 4 M[

LN@77fI iiRR F

$ lR$$ '

! M Support Column b~\

l t--

,4 T7 r-i:h ll..i.l

, , , ,i r1 tp

.I l

,.. ,7

,/i x\ .\

!

• d

,, s

,h ,,nd

,'g g

l ,' ',,; ,

( .

-l  ;

I- f )- ,

$ . h .

f E ,

}

l-

- lj i :d t (( - '

.N D N

',f' i -

4 ,

lJ)

(

ii! ... . .-

g) x i e$llsh)Mj fW$M,. f.ci?A7 cd dm N

kx 4

(.

> s(i q;, rig-@f.Wr:q( l N

x u,l , maDJ=g N

x\ ,

q( . e.d l

g H i

j ' * -

N

?j t' -

{

\

\

v 1 hje

{d A

)m\,1 J

i

..O N

,4H J'i k Figure II-5: McGuire Reactor Vessel Internals

i si al Jj -

i 1am 0 T

,$.,d,b@me+{h,-  !

hd5hd5hd5h ' ,j l 1 @s EE":5h 5 l lE3f_0,Nc.;<-

iI 5 j ' I l -

Control Rod "f .

=$

j.

i%

Ei " iAl-d Q r

lWnQ_

'/,/

5 jii ff.&jd)A i

i f(

$[(f!t, ll'l lH li,! l . !N

' Hf -

CCLL LEG  : - h[ h -[; -  : HCT LEG

%)}g (

gf' T

.{ '

?

yu hII $dhh' -

-*--LCaCC"EP MiflE t -

Ejan lE -

darum l0l l  :

i l

l Figure II-6: Standard Plant Reactor Vessel Internals (From NUREG/CR-4183 (Ref. 5))

III) "BEST ESTIMATE" CALCULATIONS In support of the NRC review of Duke Power Company's proposal to delete the UHI System, Sandia National Laboratories performed an analysis with the TRAC-PF1/ MODI computer code in which McGuire was modelled without UHI and with the minor modifications to other plant systems which are associated with the elimination of UHI. The results of the Sandia analysis of McGuire without UHI were compared to existing TRAC-PF1/ MOD 1 analyses for McGuire with UHI assumed operable in an attempt to determine the impact of UHI removal in "best estimate" calculations. Results of an analysis performed with TRAC-PD2 for a standard Westinghouse plant were also provided in the Sandia presentation to the ACRS to allow a comparison of McGuire (with and without UHI) to a plant with the standard design. The apparent difference in plant response between McGuire and the standard plant led to the ACRS and subsequent NRC inquiry regarding operating margins.

The standard plant evaluation included in the Sandia presentation to the ACRS and used in this evaluation was performed with TRAC-PD2 and included as Appendix E to NUREG-0896, Safety Evaluation Report Related to the Operation of Seabrook Station, Units 1 and 2, March 1983. Table III-1 provides the input assumptions used in the NUREG-0896 and the Sandia McGuire without UHI analyses. As shown in the table, except for the differences in plant design (containment, vessel internals, etc.), the only significant variation in the analytical assumptions involves the value for peak linear heat generation rate (LHGR). The impact of the conservative heat generation rate assumption by Sandia upon clad temperature predictions will be discussed in detail in the next section of this report (LOFT Experiments).

Figures III-1, III-2, and III-3 provide the liquid volume fractions for the lower plenum and core for the NUREG-0896 and Sandia McGuire without UHI analyses. The general behavior of the two plants are similar except the McGuire refill /reflood occurs several seconds later than is predicted for the standard plant. This delay in refill /reflood was also predicted in the Evaluation Model analyses discussed earlier. As in the Evaluation Model cases, it is believed that the McGuire vessel internals and related enhanced core flows during blowdown compensate for the delay by reducing clad temperatures prior to refill /reflood.

Figures III-4 and III-5 give a comparison of the clad temperatures predicted by the Sandia McGuire without UHI and NUREG-0896 standard plant analyses. The comparison does not demonstrate the lower end-of-blowdown clad temperatures for McGuire or the general similarity in behavior between the two plants due to the different assumptions regarding linear heat generation rates. Figures III-6 and III-7 provide results of power distribution measurements for the beginning of cycles McGuire 1 Cycle 3 and McGuire 2 Cycle 2. As shown, the steady rate peaking in the McGuire cores can be bounded by a total peaking factor (Fg) of -1.75 (-9.5 kw/ft) since the beginning of cycle peaking is larger than the expected peaks later in core life.

A comparison of McGuire without UHI and a standard Westinghouse plant using "best estimate" techniques demonstrate that the refill /reflood behavior is similar and agrees with the Evaluation Model predictions discussed earlier.

The significant differences in the clad temperature predictions between the Sandia analysis and the NUREG-0896 results are due to the different

J assumptions regarding linear heat generation rates. LOFT test results will be

{

discussed in the following section of this report to support the belief that the cause of the clad temperature variations between Sandia and NUREG-0896 predictions are due to the differences in assumptions for linear heat generation rates.

1 l

I I

)

1 i

1

)

[

l

-w-w---_ , c - - -- _-------,-3 - - , . --- ,-m7 g- ., .,m. wy

y TABLE III-1 Comparison of Sandia Analysis for McGuire Without UHI and NUREG-0896 Analysis Sandia TRAC-PF1/ MODI TRAC-PD2 Parameter McGuire w/o UHI NUREG-0896 Plant Type W 4-Loop (UHI) W 4-Loop (Std)

Containment Ice Condenser Dry Core Power (MWt) 3411 3411 Average LHGR (Kw/Ft) 5.58 5.44 Peak LHGR (Kw/Ft) 12.12 9.13 Cold Leg Temperature (F) 558.1 557.6 RCS Flowrate (Ib/s) 38889 38938 CLA Water Volume (Ft**3) 950 1046 CLA Pressure (psia) 640 600 I

1 i

I

15 r---- . . . .

IDWER PLENUM

. . . gag 5 2 ' -

U ,

Beginning of . 's l refill .

'... . . l' Su O

s

.i 1 >  ; /.

$ G ,

', 'l *

.l *.:

g  :: ~.

O k meinnans .c r.n e O to ED W 48 sp Q num Arnat summa (.)

Figure III-1: Liquid Void Fraction in Lower Plenum and Core During a LBLOCA for a Standard W Plant (From Ref. 4)

1. 3 Z - . 2 O O

- a

• Is >

u .8 -- ..

0.8 u

< c 2 . ~

w -

C W a I r D O

+6 .. --

e w . c-O N

Mo-UHI >

~ ~

O

- .=.

3 O .4 -- --

0.4 C J

I - -

r D =

z 2 C.

.2 -- --

0.2 bc.

2 - _ .

~

w -

u

• 2 O ' ' C a 0. ' ' ' ' ' ' '

O.0 a 0 10 20 30 40 50 60 TIME (S) l i

Figure III-2: Liquid Volume Fraction in Lower Plenum During a LBLOCA for McGuire Without URI (From Ref. 3)

1.

g

.8 -

0,9 g_

U -

y G - g e ~

' .6 - .. c,5 2 W u r r D . D I .

o =

0,4 s

.4 ' ..

C

- _=.

o -

o 5

p' J .m __

w .2 - ' e'.. 0.2 ..

x n Y =

o 5 u . 1:o-uni . y O. r 0.0 0 10 20 30 40 50 60

TIME (S) i l

l l

Figure III-3: Liquid Volume Fraction in Core During a LBLOCA for McGuire Without UHI (From Ref. 3) t I

l t

C 1300-L A

D 1200- MNS w/o UHI Maximum Average T Rod (Ref. 3)

E M 1100-P O

R 1000- ll A r i'

T .

'i U

R 900 - ' _,}

E \

-:'A

/

'\

D 800 - 's /

E \/ 'm j G lNUREG-0896 Standard

""'"E "*#

F 700 - i Region (Ref. 4) i I i i I i 8 1 l 4 I I i 10 20 30 40 50 60 Time (s)

Figure III-4: Clad Temperature Response for Standard Plant and McGuire Without UHI "Best Estimate" Analyses I

i C 1400-MNS w/o UHI A

Hot Rod (Ref. 3) 1200-T E A M 100& (Ne/s , , _ _ - - - , s,_

P E \

R \s '

A 800 - i T l i

U R e NUREG-0896 E

600 - i Standard Plant, t

i Hot Rod (Ref. 4)

D j

E 400 -  ! ,

G s i

F i i i i i i i i e i i e i i 15 30 45 60 Time (s)

Figure III-5: Clad Temperature Response for Standard Plant and McGuire Without UHI "Best Estimate" Analyses l

l l

l 34-F R 32-E q 30-U E 28-N C 26-Y 24-U M 22-B E 20-R Sandia Analysis for McGuire 18- w/o UHI (Ref. 3) Assumed 0 F = 2.17 Q

F 16-A NUREG-0896 (Ref. 4) Assumed 14 S F = 1.68 9

S E 12-M B 10-L I 8-E S 6-4-

2-0.8 1.0 1.2 1.4 1.6 1.8 2.0 2.2 Maximum Peaking Factor (F ) Per Assembly Figure III-6: Maximum Peaking Factor (F ) Census for McGuire1,BeginningofChcle3

l l 4

)

l l

\

l l

l 34 -

F R 32 - l E i 30 -

l E N 28 -

C l

y 26 -

N 24 -

U l M 22 -

B

~

Sandia Analysis for R

McGuire w/o UHI (Ref. 3) 0 18 - Assumed qF = 2.17 F

16 - NUREG-0896 (Ref. 4)

A Assumed F = 1.68 S 14 -

S E 12 -

M 10 -

f8 -

S 6 -

4 -

2 -

0.8 1.0 1.2 1.4 1.6 1.8 2.0 2.2 Maximum Peaking Factor (F ) Per Assembly Figure III-7: Maximum Peaking Factor (FO) Census for McGuire 2, Beginning of Cycle 2 l

l

IV) LOFT EXPERIMENTS ,

In order to show that the variation in the results predicted by."best estimate" calculations for McGuire withcut UHI and a standard plant are due more to input differences related to linear heat generation rate than actual significantdifferences in plant behavior during a LOCA, experimental data from the LOFT facility have been reviewed and key points will be presented in this report. LOFT Power Ascension Experiments L2-2 and L2-3 were' tests of large break loss of coolant accidents with similar initial conditions'except L2-3 was performed with significantly larger linear heat rates than existed' for L2-2. Table IV-1 provides a summary of the initial conditions for the two LOFT experiments.

The following figures were extracted from References 1 and 2 in order to demonstrate that the general thermal-hydraulic behavior of the LOFT facility was similar for the L2-2 and L2-3 experiments.

Figures IV-1, IV-2: Flow Rate (liter /second) in ECCS Accumulator A discharge (high range)

Figures IV-3, IV-4: Flow Rate (liter /second) in ECCS LPIS Pump A discharge Figures IV-5, IV-6: Flow Rate (liter /second)'in ECCS HPIS Pump A discharge Figures IV-7, IV-8: Momentum Flux (Mg/m.s ) in reactor vessel downcomer stalk 2, 1.13m above reactor vessel bottom 2

Figures IV-9, IV-10: Momentum Flux (mg/m.s ) in reactor vessel above upper end boxes of fuel assembly 1 Figures IV-11, IV-12: Differential Pressure (MPa) in reactor vessel core from upper end box to lower end box of fuel assembly 1 Figures IV-13, IV-14: Differential Pressure (MPa) in reactor vessel intact loop cold leg inlet to reactor vessel bottom Figures IV-15, IV-16: Pre sure (MPa) in blowdown suppression tank under downcomer 1 Figures IV-17, IV-18: Coolant Temperature (K) in reactor vessel at fuel assembly 5 lower end box Figures IV-19, IV-20: Coolant Temperature (K) in reactor vessel at fuel assembly 5 upper end box The plots of the various parameters show that the ECCS performance and system behavior during the L2-2 and L2-3 experiments were very similar. Additional information regarding the experiments may be found in NUREG/CR-0492, Experimental Data Report for LOFT Power Ascension Test L2-2, February 1979 and NUREG/CR-0792, Experimental Data Report for LOFT Power Ascension Experiment L2-3, July 1979.

The linear heat generation rate initial conditions are detailed in Tables IV-2 and IV-3. Given the similarity between other initial conditions (Table IV-1) and the thermal-hydraulic behavior (Figures IV-1 through IV-20) for the L2-2 and L2-3 experiments, it seems reasonable to conclude that variations in the clad temperature response will be due to the differences in linear heat generation rates. The clad temperature responses for several core locations are provided in Figures IV-22 through IV-31 and discussed briefly below:

Assembly 1 Rod F7 (Figures IV-22 and IV-23):

The location of Assembly 1 Rod F7 and its relationship to Rod C7 for which the LHGR are provided in Tables IV-2 and IV-3 is shown by Figure IV-21. The L2-2 blowdown clad temperature peaks at -620 F (600K) at 7 seconds while the L2-3 blowdown peak reaches -690 F (640K) at 6 seconds. The clad temperatures for L2-2 decrease after the blowdown peak with several periods lasting several seconds in which the clad reheats by 40-80*F. While both L2-2 and L2-3 experiments had decreased to -430 F (495K) at 20 seconds, the L2-3 clad reheats between 20 and 40 seconds to a temperature of -560*F (565K) before core quenching occurs.

Assembly 2 Rods F7, F8, and F9 (Figures IV-24 and IV-25)

Clad temperatures decrease from the. accident initiation until 20 seconds for both L2-2 and L2-3 tests. Between 20 seconds and 40 seconds, L2-2 clad temperatures rise from 430*F (495K) to 530 F (550K) while the L2-3 clad temperatures increase from 430 F (495K) to 595 F (585K) during the same period. L2-3 test also included a clad reheat between 12 and 16 seconds in which the temperature increased from -525 F (547K) to 620 F (600K). Core quenching occurs at -40 seconds in both experiments.

Assembly 3 Rods B11, B12, and C11 (Figures IV-26 and IV-27):

The L2-2 blowdown peak reaches -710 F (650K) while the L2-3 clad temperatures reach -900 F (755K) at the same period in time, -6 seconds. From -18 seconds to core quench at -40 seconds, the L2-2 clad temperatures increase from ~485 F (525K) to -600 F (589K). The clad temperatures during the L2-3 tests increase from -530 F (550K) to -820 F (711K) during the period between -16 seconds and

-40 seconds.

Assembly 5 Rod 18 (Figures IV-28 and IV-29):

Rod 18 is adjacent to the center location H8 for which LGHR values are provided in Tables IV-2 and IV-3 (see Figure IV-21). The clad temperature peak during blowdown for the L2-2 experiment was -925 F (770K) while the L2-3 blowdown peak reached -1150 F (895K). The L2-2 clad temperatures decrease af ter blowdown with minor reheats occurring at 12-18 seconds and 30-40 322-

r seconds. Clad reheat during refill /reflood during the L2-3 experiment resulted in temperature increases from ~535 F (552K) at 11 seconds to -945 F (780K) at -40 seconds.

Assembly 6 Rods HIS, I2, and 114 (Figures IV-30 and IV-31):

The peak clad temperatures during blowdown were -875*F (741K) and -1095 F (864K) for the L2-2 and L2-3 experiments respectively. The L2-2 reheat occurs between -14 seconds and -37 seconds involving a temperature increase from

-510 F (539K) to -640 F (610K). The L2-3 data includes a clad reheat from -12 seconds to ~42 seconds and a temperature rise from -530 F (550K) to ~870 F (739K).

The experimental results of LOFT tests L2-2 and L2-3 demonstrate that the effect of increasing the initial condition LHGR's for otherwise similar plant designs and accident conditions is to increase the peak clad temperatures which occur during blowdown and introduce or amplify clad reheating during refill /reflood.

TABLE IV-1 L2-2 & L2-3 EXPERIMENT INITIAL CONDITIONS L2-2 L2-3 Parameter Experiment Experiment Prima ry:

Flowrate (kg/s) 194.2 199 Pressure (MPa) 15.64 15.06 Cold Leg Temperature (K) 557.7 560.7 Hot Leg Temperature (k) 580.4 592.9 Boron Concentration (ppm) 838 679 Power Level (MW) 24.88 36 Maximum LHGR (kw/m) 26.37 39 Pressurizer Level (m) 1.09 1.19 Secondary:

SG Water Level (M) 3.14 3.11 Water Temperature (K) 553 482.1 Pressure (MPa) 6.35 6.18 Flowrate (Kg/s) 12.67 19.5 ECC Accumulator A:

1.68 1.71 WaterVolumejnjected(m) 0.96 Gas Volume (m ) 1.05

Pressure (MPa) 4.11 4.18 Temperature (K) 300.8 307.2 Boron Concentration (ppm) 3301 3281 I

l J

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

I TABLE IV-2 LINEAR HEAT GENERATION RATE PRIOR TO TEST L2-2 BLOWDOWN Linear Heat Generation Rate for Core Position (kW/m)

Height Above IC7 3C7 5H8 SM3 Core Bottom (m) 0.152 7.74 7.74 10.44 10.65 0.305 12.25 12.25 19.95 19.96 0.457 12.93 12.93 21.05 21.08 0.610 15.38 15.38 24.37 24.68 0.762 14.86 14.86 23.54 23.84 0.914 12.71 12.71 20.14 20.40 1.067 11.23 11.23 17.79 18.02 1.219 8.44 8.44 13.37 13.54 1.372 4.64 4.64 8.31 8.14 1.524 2.20 2.20 3.99 3.74 1.676 1.03 1.03 1.86 1.74 TABLE IV-3 LINEAR HEAT GENERATION RATE PRIOR TO EXPERIMENT L2-3 BLOWDOWN Linear Heat Generation Rate For Core Position Height Above (kW/m)

Core Bottom IC7 3C7 5H8 SM3 (m) 0.152 8.94 8.94 16.00 15.69 0.305 19.26 19.26 31.38 31.41 0.457 20.29 20.29 32.15 32.57 0.610 22.98 22.98 36.12 36.88 0.762 22.49 22.49 34.09 34.44 0.914 17.94 17.94 27.19 27.51 1.067 16.11 16.11 24.42 24.70 1.219 12.57 12.57 17.97 18.30 1.372 7.33 7.33 11.11 11.24 1.524 3.35 3.35 5.59 5.25 1.676 0.95 0.95 2.13 2.00 I

1 l

1 75 .

1 . .grf. Pia 0 l

. ... ....  ;. . r l.. .... . .. .

' J Figure IV-1: '.a 50 y.

LOFT Test L2-2 - * - - - - + + - - ' "* * -- - - - - -

. . +.4 . . .... . . .

j . .

, . . . . , - . _~a. .4... ,. -. . . . . - . . . . . . .

a .- . . . . . . .- - . . . . - . . . . .

3 2S.

o .. . .. .. . . -.... .... .

....4

..4%

L{I . ....

U - . . . .

g . . . _,+. . + + . . . . , - . .

0 f .r + ~

4 . . . -

~ ~ .

- g' _ _ . - . . g ._ _ . _ . . . .... . . . .

-Es.

O. 20. 40 60. 80 00

-20.

TINC arfCR RUPfuRC as e Fig. 64 Flow rate in ECC'. Accus J1ator A discharge, high range (FT.Pl?O.36 11 fitestrained data was f orced to agree with the accumulator liquid level data.)

75.0 i  ! . , ,

i j j j j jrr-piao-3s-i[

f ' i ! i t i!  ; e a  ! ili  !! !! i i i i

' ' ' ' ' i ' i! i ( i i ' .

Figure IV  ?- 50.0 1 LOFT Test L2-3 ] / -( j ,!l; ,

lll

, ;7 From Ref. 2 .-

(  ;  ;

i

, 8 i

i 2 / -  ! i !i i I i j i 7 2 25.0 I I I iU'i I i i i i i O  !'!  !\d i i ! i i 4 ,,

'  ! ' !l[ l ' i j i  ;

j t o I  ! W!  !

I II ! I A  !! I 1 i , ,

g I ' ' ' ' iH I i i i ' i I C 0.0 -

I I !  ! !  ! i , , ,

E  ! i ! i- ^

3 I l l j I

!I !  !!i I i i ,

.J . : ~

O  !

.! 1 .  !. !. I I I i j j i i

! ! l  ! !! l ! ! i i! j i . i j

~

' 'I i I ' i ' ' ' ' I '

-25.0

-20. O. 20. 40. 60. 80. 300, l

t

' TIME AFTER RUPTURE ts:

IM.80 Flee receie tcC5 Arcsam381er A downerge.high rangeirr.el:W.givQtt on.

t - _ - -

r-

. _ y lFI-Pl20.SS j . .. . ... . ..g 9,

. ... . . .. . . . ... . ..I,

, . ... .... . . .. . .., e

.9 i

} .

g.. .... .

..! dtt j ;*t ;e e*

Figure IV-3:

LOFI Test L2-2 w

s.o

, .s. ..

~

,. ii t I

8 ' '-

From Ref. 1 * '

i ,,. i.., . .

i} t j[- tilf g l f-l-

! j

, i ,.

l

. - .. L ;.< -j i --

o

-- es I}. .

},}t

};,t- y; i 4 -

!- {

t,-

i i

- h-_ - .i i.-HlI-i ii i i i i E ti;i iI '* t 6 y  : -i,i _;  !, t. L, _t-.

_. t . . .4e._

. .i I 4. . .._

j,; Ii i, j.i.5 j. , 47

p --

3

- 6 8y-

- ..; .,. g r. . .-.IIL ? !-T7 -M _

O.0

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

tint arfra nuptunt ioe Fig. 66 Flow rate in ECCS LPIS Pump A afiscfiarge (FT-P120-85).

(OEUD.)

10.0 ' '

lFT-P120-085 l i i i  ! i i ,

- i > j jii i i i ,

! i i

• i ' ' ' i ' ' ' i ' ' ' '

Figure IV-4: 7.5 LOFT Test L2-3  :

' i ' '

I ' I '

' !I '

t From Ref. 2 .u. i ii i - ! i ii*'

!!ii iA-"~ i' I y i I i! i 4 i fii.r.

i i 5

50 i i i i i i . ! i !

iii m, , 6

. ,/ i f,f, d  ! I i i i i iA 1! /i i  !

I  ! i i ! i/! NAi  !! '

3 I I i i. i !(! i Vi i ' iii ' '

E I I i ii l i /i i iI e ii t 2.5 y  !  ! ! I i ! !! ! 1! -  !

a i l i i i j 'i i i i i i i i , i 1 0 I I II ll l I i'

> j ! ! i ! ii i i

- i,i ,

I i i i i ii i ! i i iie i+ ,

-20. O. 20. 40. 60. 80. 100.

TIME AFTER RUPTURE (s)

Fis. Il Flee re,e in f CCS LPts Pump A di.cherse EFT.PI:0-485h0EL DI.

O e oo ~

.[rv. ease.~eow ]

. .s- . . . . . . . . . ....

i So

.::p# ~~ e q ...

Figure IV-5: 3 LOFT Test L2-2 . . . . .... . . . . . . . . .

l From Ref. 1 - . . .. . . . . . . . . . . . l

^

s e ao 3 . . . ... . .... .... . . . . Y...

s. . . . ... . .... .... . . . . ....

. . . ... . .... .... e . . . ....

~

e.se

. . .. . .... s . .. . . . . ....

o.oo

.ao. e. as. we. so. es. see.

tant artca nuetuac e..

Fig. 67 Flor rate in ECCS HPIS Pump A dischage (FT.Pl?8104).

(Otte. )

2.000 i e i i .. , , , ,

^

i i i ii

+;

i il 4 i i i i lFT-PI28-10% (

! ! l  ! ! l i  ! ! I I i i I 1

! !! i !i i n! l ! i I ,

i i

^

i i r .

^

Figure IV-6: $ .500 I I i ' i ' ' h M-

' ' ' ! lV! ! ' ' ' ' ' ' '

LOFT Test L2-3 -

From Ref. 2 g

!'h^[ ! !ll!

! i  ! i i t ! !

!',!l i

l i l l 6 iiii i ,

2 1.000 O I I I I  ! 1 ! ! l ! . ! ,

d I i  !  ! ' I l  !! ! i i - i u

I  !

!!'! !I !

i i e I

l l  ! I ! 1 i i E i i i i i i '

i 8 i i , ,

C 0.500 ' ' ' ' ' '

I  !'  !  ! ' ' '

3 i i l i  ! l 4 a

.2 '

0  ! l !I i 6

1. i i  ! I  ! I i  ! i i

• i!

' i i i i i '

i i i i '

0. M

-20. O. 20. 90. 60. 80. 100.

TIME AFTER RUPTURE (s Fig. 82 Flee raie 6n FCCS HPl5 Pump 4 discharge iFT P128-104'iQF L Os.

I I

8% U _ _

.... .... . . . . . . . g ,,,g ,.., ,

00I I

. . .. . . . . . .. . . ..a E

. e .. .... . . . . . . . .... . ...

gg gg gy_y. - -

e0 o -

LOFT Test L2-2  %,

p . . . . .... . . . .... . . . .

4 0 -

8, .... .... . ' . .... . ...

m .. .... . .... . .

m. a . .

} OO e . . . ....

. . . . f. . . . 454 4. .... . e. .

. . . . .1. .. .... . .44

.h .... 1....

2 .... 4... ... . ..

Y .S O . ... .... . . . .

o . . . . .... ....

E . . . . .... .... . ... . . . . .. &

. . . . .... .... . ... .... . t .

..,. .4 .

Ij t e.f

-se o'

-20 0 20 we 80 SO 800 t l eet atTER AuPtumE eee t . ,3, 10 Prene,t.c flup en reAsthF VP% s. l fopwSF teWher il 4 lb 2, ! . l I m abnes re.s t nr ve..P' .. tr t te+ f Pl .?'.T.00l l. f ir en<l. '

10.0 r> ,j , , jME-aST-00t {

.i  ! . . i  !

1 ! . ' i i i ll U e

!  ; i j j i t i i i , . . i i , i Figure IV-8: .. . In r i i lll LOET Test L2-3 n

• ll , ,  ;

l From Ref. 2 e mLa I,-

i ! !

i ! !

Im1 IEI i

i l

l 1 0.0

' ' I #^ ' I

% um I w y wwawi r  ! I 1 b 1

- r

! i i de . .l  !  !  !

x  ! i i i 'i i i i . ' IPEl I I  ! i o i ! i i i e !  !  ! !E  !  !

J i , , , , , , .

iI . I i w -5.0 1 i E i ' i  ! -

r i  !  !  ! i i  ! E  ! I !

i i i ! i 6 3  ! i i E z '

i , i .

i .'

< 1 i 6 i i i  !

w i , ,

r -10.0 '

f!

Ik i  !

i!

i! '

! i .

i ! '  ! i ' i ! i i i  ! !

-15.0

-20. O. 20. 90. 60. 80. 100.

TIME AFTER RUPTURE ts)

F 63 101 M.m.atum net la re.ct.e .ennel Domec. met $isik 2. l.13 m ab..e reactor .eenet b.et.m tME.hi40lHresarsened, m.,

e.s indicate meg.eende .i d.e ec.mer n..L

1 S . .

.... .. . . .. . . .,e..oo,

.... .. . . . .. O ME-30P-ool

.... . . . . . .. . t l,'.I'.l.... l

~~ g a Figure IV-9: .

• - 8 LOFT Test L2-2 From Ref. 1 I.

tg

• lflN I

[ Ill 5 .... .... .... ,

,, e3 _-_

3 .... g .. .... .-

. i . -4 d .... . .

W

. . 4 <-

g .& . . g -4 2 - - .  ; A _

l r 1 0 0 II III! IIIIIi1 y '..y 4 j .I 8

. . , I

!I l o .  ; .. a ... & ...

. -h - -N Ie II- 1 La

• . . .  ; 4 ... ....' '

, p.

, t

. .... . ..i .l.}-., 1 .i . . ! ' - -f t ;- i.

1 t t

-&.s 80- 100

o. 20 *o. 60.

-20 tant artEn Ruptung s,i Fig.10 Momentum flun in reactor vessel above upper end besti, of fuel Assemblies 1 and 3 (ME.luP.001 and ME.3UP.001). (Irend.)

4.0 '

fffk'

,i kkNi i

-f ~d

, i.

l lME-luP-001 {

' ' i i ii i i . iii iiii i ' ' ' ' ' ' ' ' ' iii Figure IV-10: 3.0 LOFT Test L2-3  %

!!' , ! i li,,

l,.' !' ll ii ll i!

From Ref. 2 e i ,

!, ', , ' 1

! ', i , i ; , i

• 2.0 i  !

1 . i i . 1 i t  !

! ! i ' i ! i i 'i i i i i  ; i I

' ' i  ! i ' '!  ! l i i i e i

x '

I a 1 t 1 i i

' i i d 1.0 '. . ,

i ii l .

r ' ' . ' m_ ' i i ii! i i .

l i i~b ' ' i iI e I n <IL i 3  !

i f; ' '\ i i

i L.  ! i i .i m '  !~'= ==i i z  ! ~

~ ' ' ' ' 'T '

I' b Y 0.0 /r ' t i t t l o  ! _

' ' i , i ! .  ! , , ,

r J'i l i  !  !

l'ifl

l. l! i i .

l,'

!ll

-l.0

-20. O. 20. 40. 60. 00. 100.

TIME AFTER RUPTURE (s)

Fig. let %=essem neu se reacter es # nee, epper end bein of Ieet Anwebly I eME-It P.06u tresarsined, gennenette ensgassedet l

l l

l

F.

o.co

.f f NZ_ ZZZ.~l'OE-****08 l I ,' f _ l

., 6 & :

...y.._7 u

} ..p -g

.y I -_- ._

._g__

. . .__, q +

.4 Figure IV-11:

• o.oS LOFT Test L2-2 I 'I~~

..._ _.__ .fh~~ 1 ~_ -~ ._ . '_ f From Ref. 1 iisii iiiii i l EEE 5 !!!!!!!!E-

_I.._, .. t.

a ._ ._ ___ _ _

s-o.oS '

!, _ . l' . & . l n --., 6. . .

i r .4

. r -I ._ __._.__ ._.1 . .

g&,.

.e.+..

g .6 s ,--

_,- L-f ... .- -.

.g I-u #_

,. i1 to

,,J.

+!st r.

. . . F l- 4_ . - . .,. f }- { . -*I -

. ,t } ' r.,

• c. 8 o .._,t-+--

-o.IS

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

-30. o. 30. wo. so. so. 800.

TIME arTCn muerunt see Fig. 96 Otfferential pressure in Reactor vessel core from upper end box to lower esul box of Fuel Assembly 1 (PDE-RV.002). (Trend.)

0.05 ll l ll! lflPOE-RV-002 Figure IV-12: I lll LOFT Test L2-3 $ l From Ref. 2 0.00 i

!  ! 'R ! I fl l II!! II I s , il l ll ll

-0.05 l Ill illi l ll

-20. O. 20. 90. 60. 80. 100.

TINE AFTER RUPTURE (s1 Fig.129 thtferenssel pressur,6e resciev .enet core from espee end bot so lowee end bos of Feel Auembly I (PDE RV-802J treussened. dynames tape.

i i

0 88 i

.. -4.l.i,. . 1.rt _.i }__.q (,0s.nv.0Os g.

y 1 ,

?

, * - * )s-.!_. _g. _l lt

. . ..r--- g p - -

Figure IV-13:

1 0.05

'4

.+. p i. I IIIIII ._._.._ __ _..

LOFT Test L2-2 {= { -}---- '

7 ( ll,lj l l l -

r

1. ---

From Ref. 1 3 t i g 4, _ _

g ._ _ _._ . . ._.. . . .

y 0.00 MW l ui-= = = ==

__{_g_

J g =_=_=.

___g_

_ . . . .m__

g . 6. ) . ___ _. _,.

is-0.05 a W j . ., , t_4_

-=

4 4. 4

4 - -- i i

,, inggy ii, ,g,g i i --<- -t -

44 t- 1 7 -

it o

l**- ,

. . . . . .. 4. ' ,

, i.i lg t , -t- -- f 7

-0.ie

-20. O. 20. 40. St. St. 100.

- tant artta nuptunt se Fig. 97 Olfferential pressure in reactor vessel intact loop cold leg inlet to. reactor vessel bottom (PDE-RV-003). (Trende goot for initial conditions only.)

O.15 ,

! i .'  !

i

' i +

! I 1

2 t

(

iPUE-Av-003

- [

1i i > I I i ,, i.,

Figure IV-14: E 0 30

!i , ,.

LOFT Test L2-3 E l,' ,[j, ',

lll l ll'l ll' w i1 u II I i, ,,,i From Ref. 2 ' i i  !

i -

1

' ' I

$ 0.05 m

i i i a a mi l, I '

f i f i i, i i

a . i ,

, i si i l J 0.00 W '--

f l 5  !!# ,

• e ,,i i , ,

(

z y i . . . . .

-- ii

$-0.05 O

(

I - , ! - i i  ! > I i'l 1111II IIE EIII i

! ! ! ' 1 I i 8 l' I I IINIlli  !

O i i  ! ' I I I I I E II'I  !

'

• I -

F 1 i

-0.10

-20. O. 20. 40. 60. 80. 100.

l TIME AFTER RUPTURE <si i1 3. 134 Differenteel pmsere en reactor eenet antact loop co4d les en6et to reecloe mwl bestom a PDE.R % .00h IQEt DI.

l l------- _______ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

o.%0 ,

i --- - -- -

Czec-sv-oos [

O'3'

{ '_ .'_b_ . .

~

Figure IV-15: l .

/ ,

ei, LOFT Test L2-2

"- ~-~

, /

From Ref. 1

, o,3o , , l -f h~ .[ -)

~

g' --- -

Z _

l

_ o.25 ,.

y ---

7 a -.,_

y 0 20 .-_

o.55 i

o.no-

-ao. o. 20. wo. so. 80. 808.

vinc arism mueruns see Fig.113 Pressure in blowdown suppression tank under Downconer I

( Pt-SV-003 ). 10luo.)

l 0.5

' 0 PE-SV-004 A PE-SV-003 l l 1 4 ! t 9 i 0 . ie Figure IV-16:

l l l llll ll

$o"ma2I~ - b,Tldti l I i Nl)N g 0.3

~

_ac i

w i (t: . l !

!o2 r i i ,

e a

0.8

..'^^ I :j!

i! i l l!l t

. Il 0.0

-20. O. 20. 40. 60. 80. 100.

TIME AFTER RUPTURE (s)

Fit.142 Pressere 6e Moede=a suppressee Isak onder Deencemer i sPE.9 401 and 40h (QEL Dh

soo.

. Tc- ..sei g a TE-sLP-sea

, a Tc-ste-ees o Tc-ste-se,

. %L Figure IV-17: x. sso. g LOFT Test L2-2 w V From Ref. 1 g (

~

1

• soo. ' '

a 1 z 1 t

.- t E t

< =se. t J l 8 h -

u i m y _ , - -

woe.

-ae. o. as. we. so. es. oe.

T Ltat AF Tgm muPf umE eee Fig.178 Coolant teaserature in reactor vessel at F al Assembly 5 1cwer ved bos ( TE-5LP-001. -002. -003, end .004 ). 10EUO.)

600. ' '

o TE-5LP-001 -

' I! '

- !! a TE-5LP-002 -

! !'! 5!' '

e TE-5LP-003 -

!  !-^'- II iA,'  !  !' .i :i O TE-5LP-004 -

Figure IV-18: j 550.

i ii

, ,,  ! ,g; i

, i i i LOFT Test L2-3 y i ii i i !q  !

! ! i i i i '!

' i i i i i From Ref. 2 g i ! ii

!!i\

I i i !T

! ' I i 1 i !

i ! i i i i !

I

$ 500.

y i e

! !! T! ! ! i '  !

' ' i $

I i  ! I 'i! l!j i l i i i ! i i ' i U i ' ! I i i ii Mii i i !i!i I i i r

- i i i  !! ' ' It ! !  !,i  : i i i i

a z i i 16 i i e i i iiie J ' '

! ! ' !  ! !l!  ! ! i ! t  !

8 i i i ! i i !M_ l i !!  !! i i u , i i i , I I?QIAal i i i -

i-i !

- ' i mi . i ! i !

i i !  ! i a i i i ; i i 400.

-20. O. 20. 40. 60. 00. 100.

TIME AFTER RUPTURE (s) flg.

• Coolset semperature se rescier eessed si Feet Aswebly $ lowe, end bos eTE SLPal. 402. 403. sad #4t aQEt DL

1 l

l l

I l

l

)

6s0. ,,

~

is e it-sue-006 n o it-sue-coa um a vt-sue-eas Figure IV-19: = o it-sue-ee% .

s0o.

LOFT Test L2-2 g x tt-sue-oes -

Form Re f . 1 w n

i 4

ss0. 'N I I 1 11 h a I I i b.

500. ___,

ll

( ___

• I I L 11 1 k

s i 1 o

o

%50.

_____ __ __'_i q m ____ _____ _____

o L__ __ ____

q gggg ggggg ggg-_ _

I ' '

woe. 30. 100.

-20. O. 20 %0. 80.

tint AWttR Ruetunt ieE Fig.179 Coolant tenverature in reactor vessel at Fuel Assembly f apper end bon f TE-5"*-001. -002, -003. -004, and 0051 'Ottl0. !

700. i , , , ,

o TE-5UP-001 h:r ; ji i iiii i ,,, ,,,

i iii ,i i i a TE-5UP-002 -

i i ijjj e TE-5UP-003 -

Aijl ,i i i.~ ,

O TE-5UP-004 j jm ,j;  ;,,,

• TE-suP-005

$ 600- ,;,; _

Figure IV-20: ,,! j , ,

, ! ,i, LOFT Test L2-3 w i !! i . I! i i! i !iii i i ! i i i i i Form Ref. 2 $  !!!i N i - ii i ! i i !!! iiii

• - '! 8 ! !iT i !!.. i i ! i i !! i i i i a: 500*

' ' ' -M I I I I

• I ' i ' i E  !! ! !  ! '+* \ km i  ! ! i i i e i r i j i j i ,jj im g' i j i i j i

~

j j i U i i i i !i! i  ! i" L/II I i i l! i 6 , i

,_ I ! ! I 'ii i!lP - _

_;i.; a ' j i z . i e i i i I ii! 'l i i jg , , , ,

4 400, iiii

!{i i e iii i i i , i a i i ' i e i o _I { i I i ! I, ij Iiii i , i i i  !

u i i i 3

,i , , ijii j ,  ; i , ,

I i 8

, I j i i l l t i 300.

-20. O. 20. 90. 60. 80. 100.

TIME AFTER RUPTURE ts:

Fig.198 Ceoiset temperature se reactor teswi at f see Anwebly $ uppee end bes (TE $L P401,402,403,404. sad 405HQEL Db.

Broken loop E cvs :i: H O - H s . 4 Broken loop cold leg !5 '

• i"4'* *N j o4 3.r. .g' C'OC oo.3agg. G = :i " ". ."g hot leg

/LM* 9'<p3 8P'* **@ eT 99T59'* N

/ Ma ..g gx 6 ecve@ v4 r H O %~Mrefv ea.. 4, .e.<2.;<i.. aq a.a.....q  %

eGe9@9ssa e@CX3/ges,,6B e .- e f N6iEb2B3 9 s&66 tcyac. .a .wa . c ... 3....Cess . ..c. a90 g2earsit2 444. s as B @@ a ,

10. ese8696064uG444A004essene,Os e s 809 e6eMe=3e006-"900666699 990SQ;-

v()

62 me wx xr,;.a.r.

- x w., e, a.me,a,;,4 g ,

e se eta'<30de,*, <*e@g a

e >Gev. - ga dGee ge,@.;,sse

e. em xd@eee~eee(

x w xn p s' 9

• s

..... a,r*1a.eM..u..OOO99999.g3.m.

.> g3ae .y . .

emoeeeee _

, ,r. 2 aa geeg

.;.s.oooges.wp.3g acgyaaop e gos e., e.34 3 as 39..y . 4 .. -ai.r5n.,gga.3geg6;R.,3f.se*

yp.ge..e..p;..yg..34 93.a... y jge g.3,.3. ge.3e gg g, b pg.g p g...yrg,6@ EGG 44 6$008089508909e Oe90se@G98t$8e'30 9 c3 e 8090EQ999s9730e88860te +9 @+-

l99990 m.

6 9aggpeam.org,EiheM66'3086 rs 5See66&$de*6 BO 99908 i .

Me3 fs9.g6..SGOsmG.ogeG

..pgor.AO.ef+4 m Bee 4 ss*3 6 0e 3e@86eOOO*G 0G 5@550eeeef00Ed00*4 0 60e 0 s0999De G O +e e 9

ep960 m OFee000000000495 5 ee6 moorhg;,Q.seee.~L.6egggggasegogqqs4

• 6 sv.y y4

. 0000a9ee9aaGQO goggse34ge%e O S 9-a Mo eh'* 6'O ewo$ $36000eee000se0999999080

(*O seee.

MO O6.pgree3s&

2 e00*M..ee,, gee ewe 99%!j09.4e%9'094909r Geed.8@ea e2 <3 e se 3 e d09 hin.a a a a.o.mc20..uro.pa.jp_a Gs G s60i!*is G GG G OE.9 90 0 00 a 993rs'i&&betenee*

.:93 43 36d3a ,,..eee.e%.p.~.aeeGec a(3.a .a.i 3 . eu.ar.eG~sOG ers.cae.& 6.y,s' di s

sGa e ' ..a T. insp+

q *n .,c

,,a..,

,eieo ..... p. ...e s ...c. r . , ..+ . s . .. .. a sa ..

..n X n ei e e e n~$hf9*3'@b6D@*3855*E95N65E*6*Eb555$$45'*$b5 [ 9O' O Ne0 =*e***90e*M*e00eG e **e0*ie09st*6@ 990 #* 9a e* E 'V

'4*Nb r a.

c.3  % .. ap g"#*d***#0******a3,

"*3 3.. ".;.3 ees.3..p a c."w c.g g"we+ /

55&&'3 bGe';$5@ 5 3 0 c 56 ? E N40 &e~y. 1 2 s 9.i @ es6 , s jb a eba 9.

r (c.,<3 s.3

)ssej @a @ dde @ @ d 9@ J5@@d e &*.* es.an.i. af gi. i. s pp @.Ag gg@ g .j{* c'r e c) .? .i pg a 0.=,>

v wi. e.;*0,. @ " s'.' ..;; ,3 s .. *. .la s.4.(D@')b,i"(R

.. a r .-

N.; . i.3.T.G()g66(Of*h4, Intact loop N"p"00 6 0

9 3., .', @M*l'."%g. D., ,' %a intact loop g ,, ,,. ,, , ,pgggg hot leg cold leg INEL-A-6901 Identification key, 2

Instrumented h Guide tube 4 5 6 Q Pin 7 9 8

Un,nstrumented i - @ Pin Core key,

-@ Guide tube 1 N

Figure IV-21: LOFT Core Arrangement (From Ref. 2)

r eso. _____ ______

e T E- t F 7-e ls e TE-IFT=ett

& TE-IFT-ete Figure IV-22: see. o Tc- F1.-ese ,

LOFT Test L2-2 -

v From Ref. 1 u

g sse.

(

4 s

[ see.

w r

E I 9

.s u *se.

__ ___ _________ u gE E EEEEE EEy-_

____ __________ _ _i t ___

woe.

-re. e. as, we. so. es. see.

Tint AFYgn nuptunt iea Fig.186 Temperature of cladding on Fuel Assembly 1 Rod F1

( TI-IF1-015 -021, .026, and -030). (0(UO.)

650.

^~ ~ ~~

~

o TE-IF7-015 7 ,

. _ . _ _ . . . _ . _ _ ..4 - _ _

.i_- '

e TE-IF7-026 Figure IV-23: 600. .-

O TE-IF7-030 -

LOET Test L2-3 - '

-n -

x , -

\. s I From Ref. 2 -

, ip i: . ,

i. \

w '

e'na a 550- 1 N1 m ~ ~ ~ ~' *^

? _

• F i yk I I. 'I  !

j . . . _ . _

~_ , p . -

y . _ _.- _..

% 500. l y - . - ._._

g , , ,

4 _ . __. .._ . . . _ . _ .

.J 5e 5 0 . '

U .. _ _ . .- . . . + _ _ .

' i^-

. . .. ._ . _ - . - - +

-M!' T TTC 5e 0 0 .

-20. O. 20. 40. 60. 80. 100.

TIME AFTER RUPTURE is)

Fig. 20$ Temperature of risodsat se Feel Asse=My 1 Rod F1:TE.llT415. 421. 026. sad 4wugit Dt.

{ .

1

\

soo. II

,, e sc-art-oss e it-art-o37

" II a vt-are-eas h lI o gg.are-esa Figure IV-24: ,,,, 'L JI gg.,, .. ,  ;

4 LOFT Test L2-2 w

- q 2 -

'wi i

From Ref. I w r rum L, .i 5

" l a u I. _____

g .... _____ _____ ___ ____ _____ _____

g _____ _____ ___ ___ _

::::: ::: 55 5 5 5 5EE 8 8 5!::

-me. o. u. ... ... ... .. .

TINE ArTER RUPTURE 4ee Fig.188 Temperature of cladding en Fuel Assembly 2 Hods F7, F8, and F9 (TE 2F7-015 and 037 TE 2F8-028 and -032, and TE-2F9-026). (GEUO.)

650. , ~

,i  ; ,i i , i O TE-aF7-015 ~

! 'i  ! ! i  ! . . ! A TE-2F7-037 ~

l;  ! > ' ' . TE-aFe-Oae _

Figure IV-25: - - -

ll! i O TE-2FO-032 -

600. '

LOFT Test L2-3 -

s - iq ' ,

, TE-aF9-026 -

_h;w_r s From Ref. 2 2 t

lE .

f s/JE Ad ' I '

l !

i

' i i I  ! ! ! l

• ' I f ' ' ' ' ' ' '

$ 550. MJ 11 i '

3 9  ? I i e-I i 4 L f T I i I I ' ' '

1 500.

E I  ! I i 4 W i e 4 i i

O l e

'<  ; . i i U

8 450. I

'I

_ l t f_ _

! i 1 '

I  !

400.

-20. O. 20. 40. 60. 80. 100.

TIME AFTER RUPTURE tel Fig. 297 Temperstere of ctedding en f eet 4ewebly 2 Reds 97. 78. and 69 . TE.1F74t$. 437. TE 16 8428 and 431. and it.25 9426HQI L DI.

e 700. , , ,

~

i i I o TC-3Gil-03 ll o it-seia-Oas i i a TE-scis-Ori  ;

Figure IV-26: 650. i i o TE-scis-Oss LOFT Test L2-2 Il From Ref. I w ll

- 1 e00. I I j w

g -

-E

. __J.Z_.

4 g

550. ' e' == . _,.4 .*_ _.

S E N w m A

I ME . i_ ,,.e f ,4 j' --4 t

, um 1 ~~'

ene I, ME

. SOO. ._.

,, . -, _, ~ _;

O nE

--t-'-+

E ~

.8 E u 450. n,

[m I .

k' I E

=00.

-20. O. 20. 40 60 80 100 TIMC AFTER RUPTURE tol Fig.195 Temperat ure of riadd'nq nn Fuel Asseeht y 1 R o<t s A l ' , 8!'

and fil ( TE.3Bil .0??, IE.3812 076, md it .'C l .07! and ..'ln ' . U' 000.

. . + _ . . _ -

^

_ _ _ . . .~

~~ '

o TE-3BII-032

[." ,* ' ,*_ ' [. 2, '[ & T E - J ts 12 -026

. _ . . . - . ._ j ' e TE-3Ci1-02i

"- O TE-3Cll-039 -

Figure IV-27: 700. /2a, LOFT Test L2-3 _ _' o '

7.x- . rv

, From Ref. 2 1  ; y-* U

'Mh%jg . .

N 600* -

. IF :

i.

o _.

L t I

a:

._r._ .

h:

{, }

-+-- -

l g ._.. - _._ .

L_' (l1l ^

! r 500. g ;3 w - . . . . _ , . -

j ,!

O ~

' ' W A' -- 2:

~

_ _ - . . . - -L._.. _----

^^

} ~^ - .

d 400.

300.

-20. O. 20. 40. 60. 00. 100.

TIME ArTER RUPTURE tsi Fig. 214 Temperatere of casedsag ea Feel Assemmat 3. Rods 811. 812. and ClliTE 3811432. TF.3812 424. sad it.3Cll.021 and 439p eQt.L D).

l I

i

00o

~ b- , l j, + -- e IE-SIS-008 o ,c.si.-ons r ~ZZZ _. ~ .  : oa Tc-see-Osa it-ste-o*1 Figure IV-28: voo. .

LOFT Test L2-2 ,. . . m . _ m From Ref. 1 3 --

" 600.

l l$ l I

g .__

j

{g "

500. 1 u .

i

O 4 ,

.a u woo.

-, 1 300.

2o. o. ao. wo.

• so. so. 800.  ;

1IMC arfra muPfunt eee Fig. 215 Tewperature of ci soding on Fuel Anembly 5 Rod 18 l f it.5IS-008 -076, 041, and -058). (Qtuo.)

900. . ' s

! i i f i i . i  ! .

. . o TE-5Is-008

' ' l' ' '

! ' '!  ! A TE-5IS-026

' F i ! i  ! ' ! ! i i i i ! . TE-5:e-0%:

Figure IV-29: 800. , , l [ l l'; o TE-sie-Oss .

LOFT Test L2-3 -  ! ' ! ' '

! ie'M! ! ! , .

From Ref. 2 5

!l!4

[ '

, NY {l , ; i lil!

l y 700. ,,

i rYi

' ' ' ' ' ' I '

A7 bbi i. i . ! .

i 3

!! ii! i i.I

-  ! 'I l i _  ! i i  ! i

• i ! i fM l 7i -; i i , i , , . i

$ ' i'i . /II / iM ! 1 ~ ' i i j

! I 600. f'_ , f [ll , , l ,

% ' '  ! u.,  !! li i  !  ! l i i ! i i 4

' '  ! !  !'"I iL F I! ! +

! ' i i i

'  ! il i s",m.9! I I i ' i Q '

l i

! 7 ' '

i i

J 500.

' u y / e .

I i +

NP ' + . '  !  ! i

'  ! ' ' . . _.  ! ! i i  ! i +

8

! 1 C- _

gl a - -~-!

400.

-20. O. 20. 40. 60. 80. 100.

TIME AFTER RUPTURE tst Fis. De Temperasure of cledding on Feet 4mmber S. Rod isiTE 518408. 024. -041, sad essiIQf.L De.

800. , , - ____ -

E I e s c -ees i s-e* :

E i. e it-esoa-eas Ea a it-enoa-ese Figure IV-30: E rn o sc-se 4-oas 3ag. .

LOFT Test L2-2 ,

,y u TE-es 4-ese From Ref. 1 5 "

y soo. -- H l --

l si dE 22222 22222 22222 3 __, , _, ____ _ _____ _____ _____

s-g E I -_= 2 y E I f -

1 W

soo. C II k J i E I '

O E I

~ ~ -

• ~

I I j 400. ,,

II E I I i 300.

-a0. O. as. 48. St. St. 100.

f irst Arttr. Rupture Iee Fig. 228 Temperature of cladding on Fuel Assembly 6 Rods HIS. IF, and 114 ( TE-4H15 041. TI 6102 021 and -039, and TI-6l14-021 and -039).

(QEUD.)

900, m a TE-6Hl5-0%I Figure IV-31:

[i i a TE-6IO2-021 e TE-6102-039 800.  !  :

LOFT Test L2-3 l- 0 TE-61l*-021 -

• TE-6I14-039 -

, From Re f . 9. ,1 , , , ,

d 700. ] ,

! id lit g  ! i/

)e 600.

_ i r

i

^

p c

W I 7 - -  %.i . 2 _Fi i in  ; i a +  ! M J d 4 'at i E u m g,m1 ,

500. I f!

N

[ ._

!l O '

4 I i  ! I! .  ! . ! i '

g i u- ;1 900.

300.

-20. O. 20. 40. 60. 80. 100.

TIME AFTER RUPTURE ts)

Fig. 247 Tempeesseee of cleddsag on Feet hwati, 4. Rods HIS.12 sad 114 (TE4HIS441. TE4102421 and 439. and II.4114421 sed 439HQLt DL

[~

V) DISCUSSION OF ANALYTICAL / EXPERIMENTAL RESULTS Using the "best estimate" calculations performed by Sandia for McGuire without UHI, the NUREG-0896 analysis for a Westinghouse standard plant, comparisons of McGuire without UHI and standard plants using the Westinghouse Evaluation Model, and data from LOFT experiments, it can be shown that the operating conservatisms for McGuire without UHI are similar to non-UHI plants.

Figure V-1 provides the clad temperature histories for the Sandia TRAC-PF1/ MODI analysis of McGuire without UHI, NUREG-0896 TRAC-PD2 analysis of a standard plant, LOFT Test L2-2, and LOFT Test L2-3. Given that the LOFT experiments were performed with similar initial conditions except for LHGR's (L2-3 MLHGR was 50% higher than L2-2 MLHGR) and demonstrated similar thermal-hydraulic behavior except for clad temperature histories, it is evident that the differences in clad temperatures between the L2-2 and L2-3 tests are due primarily to the LHGR and related store energy variations. The higher LHGR's which existed in the L2-3 test resulted in higher clad tempera-tures during blowdown and the introduction or amplification of clad reheat during refill /reflood. The same relative behavior of L2-2 versus L2-3 was present for the various core positions monitored. The LHGR impact on blowdown peak and refill /reflood reheat is also demonstrated by comparing various core locations during the same test.

The Sandia analysis for McGuire without UHI and NUREG-0896 standard plant analyses were performed using best estimate input assumptions except the hcGuire LHGR's used were conservatively high (Sandia MLHGR was 33% higher than NUREG-0896 MLHGR). The differences between the results of the two analyses involve the McGuire case predicting a longer time period prior to the beginning of refill /reflood, a significantly higher peak clad temperature during blowdown, and a significant increase in the clad reheat during refill /reflood. The longer blowdown period was also predicted by comparing Evaluation Model results and is believed to be a real phenomenon related to the ice condenser containment and UHI reactor vessel internals. The longer period prior to refill /reflood is not considered to reduce operating conservatisms since the McGuire vessel internals enhance core flow and result in lower clad temperatures at the end of blowdown. The increased peak clad temperature during blowdown and increased reheat during refill /reflood predicted for McGuire result from the conservative LHGR's assumed. This is supported by the similarities in the comparisons of LOFT data for tests with different LHGR's and comparisons of the Sandia McGuire without UHI and NUREG-0896 analyses (Figure V-1).

1 TEgPERATURE DEg F cLAo

)

2 n 0 3 e 6

2 f

e L e w

(

R u e

t s . 0

)

8

.. 5 1 _

D .

O .

R 0

_ 5 g8 L N 4

• ,lI8I8 Y

L B

!,g )

2 , s M -) i gf 0 (

E 21 3

S e S L

. ,% m

_ A

( tf  %

i se  % T A eR T T( 0 A tai lEt8 s% '

2 D

T /, s s S I e E 4' r T

eat 0 u gije 1 t F

T 8 g:giI a r

O #s i 8,gg e L lgga,8a ,

p

~ m 0 e T

0 7 0 0 0 0 0 0 0 0 d 0 0 0 0 0 0 0 0 0 0 a 3 2 1 0 9 8 7 6 5 4 l 1 1 1 1 C 1

s 0 V I i ) 6 D r ) s 6 e

_ O oo3 yy9 r Mf/ l 8 u

) / w . ad0 g D l s f nr - i O Fi ee AaG

• 0 F

_ R P srR dE 5

_ - yi( 2 nR E Cl u D aU C A aG1 PtN A R ncH - S(

R TAMU C E ArR V R W 0 A Tf P 4

M l:It E I U \

_ M  % )

_ I s X 0 (

_ A 3 M e

( s m i

S T N

O I *

\ 0 T , 2 A ,

L

_ U #

C p L

A C

C g 0 1

A u

R T g t O 0

_. CLAD TEMPERAT RE DEG F 1y

i l4 ljj] ,i{ 4f.lji,i llll i l

'af , 1 iIll.!,l r !llll

f 4

o VII) REFERENCES

1. NUREG/CR-0492, " Experimental Data Report for LOFT Power Ascension Test L2-2", February 1979.

2. NUREG/CR-0792, " Experimental Data Report for LOFT Power Ascension Experiment L2-3", July 1979.

3. March 17, 1986 Letter from D. Dobranich (Sandia National Laboratories)'to  ;

J. J. Watt (USNRC) on "LOCA Calculation for a UHI PWR with the UHI System Removed",

t

4. NUREG-0896, " Safety Evaluation Report Related to the Operation of Seabrook  !

Station, Units 1 and 2, Docket Numbers 50-443 and 50-444, Public Service I Company of New Hampshire", March 1983.

5. NUREG/CR-4183, " Pressurized Thermal Shock Evaluation of the H. B. Robinson i Unit 2 Nuclear Power Plant", September 1985.

i i

l I

[

i l

I t

i i

.