ML20073M245
| ML20073M245 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/30/1983 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML20073M234 | List: |
| References | |
| TAC-49798, NUDOCS 8304210456 | |
| Download: ML20073M245 (31) | |
Text
e
- O Docket No. 50-336 Attachment 1 Millstone Nuclear Power Station, Unit No. 2 Reload Safety Analysis April,1983 8304210456 830413 PDR ADDCK 05000336 p PDR
l l
TABLE OF CONTENTS !
Section Title Page
1.0 INTRODUCTION
AND SU MARY l 1.1 OBJECTIVES 1 1.2 GENERAL DESCRIPTIW2 1
1.3 CONCLUSION
S 2 2.0 MECHANICAL DESIGN 3 2.1 GENERAL DISCUSSION 3 3.0 THERMAL AND HYDRAULIC DESIGN 4 4.0 NUCLEAR DESIGN 5 5.0 ACCIDENT ANALYSIS 6
5.1 INTRODUCTION
AND SUMARY 6 5.2 ACCIDENT EVALUATION 6 g
5.2.1 KINETICS PARAMETERS 7 5.2.2 SHUTDOWN MARGIN 7 5.2.3 CEA WORTHS 7 5.2.4 CORE PEAKING FACTORS 7 5.3 INCIDENTS REANALYZED 8 5.3.1 CEA WITHDRAWAL AT POWER 8 5.3.2 COMPLETE LOSS OF REACTOR COOLANT FLOW 8 5.3.3 STEAMLINE RUPTURE 9
6.0 REFERENCES
11 G
s .
r LIST OF TABLES ;
1 Table Title Page 1 Millstone Unit 2 Cycle 6 Core Loading 12 2 Millstone Unit 2 Kinetics Characteristics 13 3 Shutdown Requirements and Margins 14 4 Parameters Used in the CEA Withdrawal 15 Analyses 5 Sequence of Events Loss of Coolant Flow 16 6 Sequence of Events - Steamline Rupture at 17
~
No Load e
r e l
t e
l 11 l
l r, - -- , , , - .. -. -- - - , - -
LIST OF FIGURES Figure Title Page t
1 Core Loading Pattern 18
. ~
2 Millstone 2 - Safety Analysis - CEA Withdrawal 19 at Power 3 Millstone 2 - Safety Analysis Loss of Flow - 20 Reactor Coolant Flow Versus Time J
4 Millstone 2 - Safety Analysis Loss of Flow - 21 Nuclear Power and Heat Flux Versus Time 5 Millstone 2 - Safety Analysis Loss of Flow - 22 DNB Ratio Versus Time 6 Millstone 2 - Safety Analysis, Steamline Rupture - 23 Steam Generator Pressure Versus Time 7 Millstone 2 - Safety Analysis, Steamline 24 Rupture - Inlet Temperature Versus Time 8 Millstone 2 - Safety Analysis, Steamline 25 l Rupture - Pressurizer Pressure Versus Time 9 Millstone 2 - Safety Analyis, Steamline 26 Rupture - Total Reactivity versus Time 10 Millstone 2 - Safety Analysis, Steamline 2"/
Rupture - Core Average Heat Flux Versus Time iii l
1.0 INTRODUCTION
AND SUPMARY I
)
1 1.1 OBJECTIVES 1 .
1 This report presents an evaluation for Millstone Nuclear Power Station '
~
Unit 2,' Cycle 6, which demonstrates that the core reload will not adversely affect the s'afety of the plant. This evaluation was I
accomplished utilizing the methodology described in Reference 1.
those incidents Based analyzed upon the above and reported referenced in the Basic methodology, only(2) (BSR) whic Safety Report could potentially be affected by fuel reload and tube plugging have been reviewed for the Cycle 6 design described herein. The results of new analyses are included and the justification for the applicability of previous results for the remaining incidents is provided.
i 1.2 GENERAL DESCRIPTION The Millstone II reactor core is comprised of 217 fuel assenblies arranged in the configuration shown in Figure 1. Each fuel assembly has a skeletal structure consisting of five (5) zircaloy guide thinble tubes, nine (9) Inconel grids, a stainless steel bottom nozzle, and a stainless steel top nozzle. One hundred seventy-six fuel rods are arranged in the grids to form a 14x14 array. The fuel rods consist of i slightly enriched uranium dioxide ceramic pellets contained in Zircaloy-4 tubing which is plugged and seal welded at the ends to encapsulate the fuel.
[ Nominal core design parameters utilized for Cycle 6 are as follows:
Core Power (Mwt) 2700 l
, System Pressure (psia) 2250 ;
l Reactor Coclant Flow (GPM) 350,000* 1 Core Inlet Temperature (OF) 549 l- Average Linear Power Density (kw/ft) 6.065 i (based on best estimate hot, densified l core average stack height of 136.4 inches)
- Minimum guaranteed safety analysis flow-assumes plugging level of 2500 J steam generator tubes j j 1487F:6 1
Th2 core -loading pattern for Cycle 6 is shown in Figure 1. Twenty-four (24) interior feed assemblies containing 2.7 w/o U235 and forty-eight (48) peripheral feed assemblies containing 3.2 w/o U235 are replacing seventy-two (72) Combustion Engineering (CE) batch E assemblies. The batch B assembly in Cycle 5 is replaced by a batch B assembly that was discharged at the end of Cycle 1.
A summary of the Cycle 6 fuel inventory is given in Table 1.
1.3 CONCLUSION
S From the evaluation presented in this report, it is concluded that the Cycle 6 design does not result in the previously acceptable safety limits for any incident to be exceeded. This conclusion is based on the following:
- 1. Cycle 5expectedburnupof11,500[f000 MWD /MTV.
- 2. There is adherance to plant operating limitations as given in the Technical Specifications.
1487F:6 2
/
a 2.0 MECHANICAL DESIGN 2.1 GENERAL DISCUSSION 1
The mechanical design of the Cycle 6 fuel assemblies is essentially l
~
identical to that of the Cycle 5 assemblies. The Westinghouse fuel assemblies are designed to be fully compatible with all resident Millstone 2 fuel assemblies and core components (e.g. adequate i clearances for insertion of CEA's, plugging devices, etc.). Table 1 summarizes pertinent design parameters of the various fuel regions.
1487F:6 3
l l
l 3.0 THERMAL AND HYDRAULIC DESIGN A description of the thermal and hydraulic design of the Westinghouse Millstone II reload fuel assembly to be utilized in Cycle 6 is given in Chapter ,3 of the BSR.
As discussed in the BSR, the Westinghouse fuel assemblies have been designed and shown through testing to be hydraulically compatible with all resident Millstone II fuel assemblies.
The DNB analyses for Cycle 6 were perfonned for a minimum reactor coolant flow rate of 350,000 gpm and a radial peaking factor, rF * #
1.565. This is a decrease in the flow rate and peaking factor assumed in the Cycle 4 and initial Cycle 5 analyses of 370,000 gpm and 1.63.
As indicated by the power and flow sensitivities reported in the Cycle 4 Reload Safety Evaluation Report (Reference 6) a flow reduction can be offset by a power (or Fr ) reduction in a 2:1 ratio to maintain a constant DNBR. Thus the reduction in flow has been more than offset by the reduction in radial peaking factor and this has been confirmed in the Cycle 6 analyses. A reduction in flow from 370,000 gpm to 362,600 gpm and a conservative reduction in Fr from 1.63 to 1.597 was previously implemented during Cycle 5 operation. The Cycle 6 analysis takes a partial credit of 3.0% of the net conservatism which exists between convoluting and sunusing the uncertainties of various measured plant parameters in power space. Th'I partial credit was applied in previous cycles and is discussed in more detail in the Cycle 4 Reload )
Safety Evaluation Report (Reference 6).
1487F:6 4
l l
4.0 NUCLEAR DESIGN The Westinghouse nuclear design procedures, computer programs, and calculation models utilized in the Millstone II, Cycla 6 reload design
~
are presented in the BSR. Similar to the Cycle 4 and Cycle 5 ;
evaluations (6,7), Cycle 6 accident simulations take credit for the variable high power trip by terminating accidents 55 above the variable high power trip. Also P values g (see BSR Section 6.0) are computed only if the maximum allowed power density of 21 kw/ft is exceeded.
The Cycle 6 core loading results in a maximum linear heat rate of less than 15.6 kw/ft at all fuel heights at rated power. Table 2 provides a summary of changes in the Cycle 6 kinetics characteristics compared with the current limit based on the reference safety analysis.(2)(6)(7) It can be seen from the table that all of the Cycle 6 values fall within current limits. Table 3 provides the control rod worths and requirements at the most limiting condition during the cycle. The required shutdown margin is based on accident analyses presented in Section 5.0.
1487F:6 5
e -
5.0 ACCIDENT ANALYSIS
5.1 INTRODUCTION
AND SUPMARY
~
The power capability of Millstone II is evaluated considering the !
consequences of those incidents examined in the BSRI2) , using the previously accepted design basis specified in Section 1.2. It is !
concluded that the core reload will not adversely affect the ability to safely operate at 100% of rated power during Cycle 6. For the l overpower transient, the fuel centerline temperature limit of 4700 0F can be accommodated with margin in the Cycle 6 core. The burnup dependent densification model(3) was used for fuel temperature evaluations. The LOCA limit at rated power can be met by maintaining peak linear heat rates at or below 15.6 kw/ft.
I 5.2 ACCIDENT EVALUATION The effects of the reload on the design basis and postulated incidents analyzed in the BSRI2) and updated in the Cycle 4 and 5 reload safety analyses (6,7) were examined. In most cases, it was found that the effects were accommodated within the conservatism of the initial assumptions used in the previous safety analysis. For those incidents
~
dich were reanalyzed, it was determined that the applicable design bases are not exceeded, and, therefore, the conclusions presented previously are still valid. For the reanalysis, parameters were I
selected to bound a plugging level of 2500 steam generator tubes.
l l
A core reload can typically affect accident analysis input parameters )
in the following areas: core kinetic characteristics, shutdown margin, ,
CEA worths, and core peaking factors. Cycle 6 parameters in each of these areas were examined as discussed below to ascertain whether new i
accident analyses were required. l 1
l A safety evaluation was performed to assess the impact of the increased j level of steam generator tube plugging on the non-LOCA accident l analyses. The approach used was to identify the important parameters for each transient, determine which of these parameters were affected i
1487F:6 6
by the higher steam generator tube plugging levels and determine the impact on the safety analysis. In addition, a study was made of each currently applicable accident analysis to identify margins to safety limits which'could be used to offset flow reduction penalties. Based
. on this, evaluation, operation at a $5% reduction in thermal design flow and a maximum of $15% effective steam generator tube plugging level will not result in violation of the safety limits.
5.2.1 KINETICS PARAMETERS A comparison of Cycle 6 kinetics parameters with the current limits, established by the BSR and Cycle 4 and 5 reload safety analyses, is presented in Table 2. None of the parameters in Table 2 exceeded the limiting range of values established by the previous safety analyses.
However, the total trip reactivity curve as a function of position calculated for Cycle 6 was more limiting than that calculated for Cycle 5. The Cycle 6 curve was therefore used in all accident i reanalysis.
~5.2.2 SHUTDOWN MARCIN Changes in minimum shutdown margin requirements may impact the safety analyses, particularly the steamline break and boron dilution accident analyses. Cycle 6 shutdown margin requirements are the same as Cycle 5.
5.2.3 CEA WORTHS Changes in CEA worths may affect shutdown margin. Table 3 shows that the Cycle 6 shutdown margin requirements are satisfied.
5.2.4 CORE PEAKING FACTORS Peaking factor evaluations were performed for rod out of position, and steam line break accidents to ensure that the DNB design limits are not exceeded. These evaluations were performed utilizing the existing transient statepoint information from the reference cycle (BSR) and 1487F:6 7 .
- . - _ = _ _ . . _ - -
i peaking factors determined for the reload core design. In each case, it was found that the peaking factor for Cycle 6 yielded results that were within the DNB design limits.
CEA peaking factors for Cycle 6 were within the reference cycle limits.
5.3 INCIDENTS REANALYZED 5.3.1 CEA WITHDRAWAL AT POWER The CEA withdrawal at power accident was reanalyzed for Cycle 6 to assess the impact of increased steam generator tube plugging and the corresponding reduction in thermal design flow.
Table 4 gives the key parameters assumed in the analysis. Figure 2 shows the minimum DNBR as a function of reactivity insertion rate from initial full power operation for both minimum and maximum reactivity feeeack cases.
The results show that the thermal margin low pressure trip provides core-protection over the full range of reactivity insertion rates. The minimum DNBR remains above 1.30.
5.3.2 COMPLETE LOSS OF REACTOR COOLANT FLOW The loss of flow transient is sensitive to the reduced thermal design i flow because of the higher loop resistances which lead to a more rapid flow coastdown and the lower initial flows which reduce the margin to the 1.30 DNBR limit.
The loss of flow accident was reanalyzed for Cycle 6 to determine the effect of steam generator tube plugging on the minimum DNBR reached during the incident.
1487F:6 8 .
This analysis reflects the change in the trip reactivity curve, the reduction in normal operation radial peaking factor Fr, and appropriate fuel temperature curves for Cycle 6.
Table 5 gives the time sequence of events for this accident. The reactor' coolant flow, nuclear power, heat flux, and DNB transients are shown in Figures 3, 4 and 5.
The results show that the reactor coolant pump speed sensing system provides sufficient protection against clad and fuel damage. The DNBR does not decrease below 1.30 during the transient.
5.3.3 STEAMLINE RUPTURE The steamline rupture accident was reanalyzed for Cycle 6 to incorporate a more' bounding trip reactivity curve and kinetics coefficients.
The transient which was reanalyzed is the most limiting case and assumes the steamline rupture of a pipe inside the containment at the
. outlet of the steam generator. The plant initially is at no load conditions with offsite power available.
The reanalysis was performed with the assumption that auxiliary feedwater flow is initiated automatically during this transient. It was assumed that 2800 gpm of auxiliary feedwater, 35% more than the maximum runout flow, is delivered to the affected steam generator three minutes after the beginning of the transient. This is conservative with respect to the exp'ected time of auxiliary feedwater initiation since automatic actuation of the auxiliary feedwater system would occur on a low steam generator water level trip signal. The three minute time delay for auxiliary feedwater delivery is also conservatively small.
i i
't 1487F:6 9
l The assumption was also made that the minimum capability for injection of boric acid solution (1720 ppa) corresponds to the most restrictive ,
single failure in the safety injection system. This corresponds to the flow delivered by one high pressure safety injection pump and one low pressurg safety injection pump delivering full flow to the cold lee l header. , l The steam generator pressure, RCS temperature, pressurizer pressure, reactivity, and core heat flux transients for this case are shown in Figures 6-10. Table 6 lists the time sequence of events.
As shown in Figure 9 , the core returns to critical after CEA insertion (assuming the most reactive CEA is stuck in the withdrawn position).
This is due to the high cooldown rate, in the presence of a negative moderator temperature coefficient, resulting from the steam discharge and feedwater addition. However, the addition of boron from the high pressure safety injection pump brings the core subcritical again. The peak heat flux attained during this transient is small, approximately 7 percent. By this time the faulted steam generator is essentially depressurized and the primary system cooldown is due mainly to the
~
boiling of feedwater.
At 180 seconds, auxiliary feedwater flow is delivered to the faulted steam generator, which further cools down the RCS. However, the positive reactivity insertion due to this additional cooldown is offset by the addition of boron via safety injection, and the core remains subcritical .
l Results show that the DNB margin design basis will not be violated.
l That is. DNB will not occur on at least 95 percent of the limiting fuel
! rods at a 95 percent confidrace level.
1487F:6 10 l . ._ _ _ _ . . _ - _ - - . . - - - . - - _ -
6.0 REFERENCES
~
- 1. Bordelon, F. M., et. al., " Westinghouse Reload Safety Evaluation Methodology", WCAP-9273, March,1978.
- 2. Millstone Unit 2, " Millstone Unit 2 Basic Safety Report", Docket No. 50-336, March,1980.
- 3. Miller, J. V. (Ed), " Improved Analytical Model used in Westinghosue Fuel Rod Design Computations", WCAP-8785, October,1976.
- 4. Hellman, J. M. (Ed.), " Fuel Densification Experiemental Results and Model for Reactor Operation", WCAP-8219-A, March 1975.
- 5. Letter, Counsil to Reed, Millstone Nuclear Power Station, Unit No.
2, Proposed License Amendment, Power Uprating, February 12, 1979
- 6. Letter, Counsil to Clark, Millstone Nuclear Power Station Unit No.
2, Cycle 4 Refueling - Reload Safety Analysis, June 3,1980
- 7. Letter, Counsil to Clark, Millstone Nuclear Power Station Unit No.
2, Cycle 5 Refueling - Reload Safety Analysis, November 17, 1981.
l i
1487F:6 11
TABLE 1 Millstone Unit 2 Cycle 6 Core Loading
~ '
Initial BOC**
Number of Enrichment % Theoretical Burnup Average Region Type Assemblies w/o U235 Density (MWD /MTU)
B+ CE 1 2.336 95 17450 F1 }[ 24 2.697 94.54 24020 F2 }[ 48 3.297 94.87 22560 G1 j{ 24 2.720 95.00 13940 G2 }[ 48 3.191 94.70 9370 H1 }[ 24 2.70 95* 0 H2 j( 48 3.20 95* 0
+ *
- The Region H1 and H2 densities are nominal. Average densities of 94.5 theoretical were used for Region H1 and H2 nuclear design evaluations.
- EOL Cycle 5 burnup assumed: 11,500 MWD /MTU.
1487F:6 12
TABLE 2 MILLSTONE UNIT 2 KINETICS CHARACTERISTICS
~ '
Current Limit Cycle 6 Most Positive Moderator Temperature Coefficient (ap/*F) x 10-4 +0.5 from 0 to 70% Power +0.5 from 0 to 70% Power
+0.4 from 70 to 100% Power +0.4 from 70 to 100% Power Most Negative Moderator Temperature Coefficient (ap/*F) x 10-4, ARI -3.8 -3.8 Doppler Temperature Coefficient (ao/ F x 10-5) -1.2 to -1.92 -1.2 to -1.92 Delayed Neutron Lifetime 8,ff (%) .479 to .634 .479 to .634 Prompt Neutron Fraction (usec) <32.2 <32.2 Maximum Differential Rod Worth of two CEA groups moving together at HZP (pcm/in) 36.6 36.6 1487F:6 13
TABLE 3 SHUTDOWN REQUIREMENTS AND MARGINS MILLSTONE UNIT 2 - CYCLE 6
, Control Rod Worth (W) Cycle 5 Cycle 6 l
BOC EOC BOC EOC All Rods Inserted 7.84 8.67 7.94 8.65 All Rods Inserted Less Worst Stuck Rod 6.47 6.59 6.49 6.70 (1) Less 10 Percent 5.82 5.93 5.84 6.03 Control Rod Requ'rements Reactivity Defects (Combined Doppler, T,yg, Void and Redistribution Effects) 1.94 2.64 1.93 2.69 Rod Insertion Allowance 0.36 0.36 0. 41 0.41 (2) Total Requirements 2.30 3.00 2.34 3.10 Shutdown Margin ((1) - (2)) (up) 3.52 2.93 3.50 2.93 Required Shutdown Margin (20) 2.90 2.90 2.90 2.90 i
- 1487F
- 6 I4 l -- _ - .. - __ _.-_ - . . . . - . - . . _ . - - . . - . - . - - . - - - - .-. -
TABLE 4
, PARAMETERS USED IN THE CEA WITHDRAWAL ANALYSIS
. Parameter Units Cycle 6 Initial Core Power Level MWt 102 percent of 2700 Core Inlet Coolant Temperature *F 551 Reactor Coolant System Pressure psia 2200 Moderator Temperature Coefficent 10~46p/*F -2.5 and + 0.5 Doppler Coefficient Multiplier -
1.15 and 0.85 CEA Worth at Trip 10-2 ap -2.9 Reactivity Insertion Rate x10~4ap/sec 0 to 2.44 Holding Coil Delay Time sec 0.5 i
\
CEA Time to 90 Percent Insertion sec 2.75 (Including Holding Coil Delay) l I
l
)
1487F:6 15
TABLE 5 SEQUENCE OF EVENTS - LOSS OF COOLANT FLOW Four pumps in operation, all pumps coasting down Event Time (sec)
Loss of power to all pumps 0. 0 Reactor coolant pump low . 91 speed setpoint reached CEA's begin to drop 1.56 Minimum DNBR occurs 3.7 I
1487F:6 16
TABLE 6 SEQUENCE OF EVENTS, STEAMLINE RUPTURE AT NO LOAD Time Event Setpoint or Value (Sec) -
0.0 Steamline Rupture occurs --
3.5 Low Steam Generator Pressure 478 psia trip signal occurs; main steam isolation begins 4.4 CEA's begin to drop into the core --
10.4 Main steamline isolation valves --
closed 11.5 SIAS initiated on low RCS pressure 1563 psia 19.5 Pressurizer empties --
59.1 Peak Reactivity 0.2755 ao 139.1 Peak Heat Flux 7.1 percent power l 180.0 Auxiliary feedwater initiated --
1487F:6 17
FIGURE 1 CORE LOADlNG PATTERN MILLSTONE UNIT 2 - CYCLE 6 ,
A B C D E F GHJKLMNPR S T V W X Y j
- I 1 H2 H2 H2 H2 2 - - H2 H2 H2 G1 F2 G1 H2 H2 H2
'H2 h
3 - H2 F2 F2 H1 F2 H1 F2 G2 F2 4 -
H2 H1 G2 F1 G2 F2 G2 F2 G2 F1 G2 H1 H2 1
5 F2 F2 H2 l H2 F2 G2 F1 H1 F2 G2 F1 G2 H1 F1 G2 6 -
H2 G2 F1 H1 F1 G1 F2 H1 F2 G1 F1 H1 F1 G2 H2 7- H2 F2 G2 F2 Gl G2 G2 F1 G2 G2 G1 F2 G2 F2 H2 8- H2 H2 e- G1 H1 F2 G2 F2 G2 Gl G1 Gl G2 F2 G2. F2 H1 G1 10 - H2 H2 11 - F2 B+ F2 F2 G2 F1 H1 F1 G1 G1 F1 H1 F1 G2 F2 SS BP SS 12 - H2 H2 13 - G1 H1 F2 G2 F2 G2 G1 G1 Gl G2 F2 G2 F2 H1 G1 14 - H2 H2 15 - H2 F2 G2 F2 Gl G2 G2 F1 G2 G2 G1 F2 G2 F2 H2 16 H2 G2 F1 H1 F1 G1 F2 H1 F2 G1 F1 H1 F1 G2 .H2 17 H2 F2 G2 F1 H1 F2 G2 F1 G2 F2 H1 F1 G2 F2 H2 18 H2 H1 G2 F1 G2 F2 G2 F2 G2 F1 G2 H1 H2 19 H2 F2 G2 F2 H1 F2 H1 F2
$ F2 H2 20 H2 H2 H2 G1 F2 G1 H2 H2 H2 21 H2 H2 H2 H2 -
Reactor Core Initial Region Type w/o U-235 B+ CE 2.336
, North F1 W 2.697 F2 W 3.297 G1 W 2.7195 SS - Secondary Source W 3.1909 G2 BP - Burnable Poison H1 W 2.70 H2 W 3.20 18
- e , ,
s
! ~
1.8 -
BOL l.7 - -
E0L E - 1.6 - -
E I
E g 1.5 - -
i 1.4 - -
1.3 - -
1 1
1 d
9 l l l t iI t I I I I I I e if I e i e i I I i 9
.I 1.0 10.0 100.0 REACTIVITY INSERTION RATE (PCM/SEC)
FIGURE 2 MILLSTONE 2 - SAFETY ANALYSIS l CEA WITHDRAWAL AT FULL POWER REACTIVITY INSERTION RATE VS. DNBR
1.0000 : : ;
.80000 -
5' di 15 .60000 - -
5E . .
8 is v
%$ .40000 - -
-t . .'
.20000 -
0.0 -
o h O o e 6 & 4 a a S TIME 4 (SED l
i FIGURE 3: KILLSTONE 2 LOSS OF FLOW REACTOR COOLANT FLOW VERSUS TIME 20
1 I
- 1. test 1.0800 <
o.
.s.
Ii E 8 600oo -
5s
. I $ ..cono -
E 1
I t.teso g 1.00E0 <
, ,.I.
Ws
@ 5 .sosso -
G E s< .-
5.
. 1.1000 -
g 1.como -
E i
g .scoco -
8 E .soooo .
d E E
g ..cono -
v
- =
9 d .aocoo -
=
Y l .
ia ia ia .
i 4
m nect FIGURE 4: MILLSTONE 2 LOSS OF FLOW NUCLEAR POWER AND HEAT FLUX VERSUS TIME 21
I 3.0000 ' -
2.7500- - ..
2.500 & - ..
2.2500 -
b E 2.0000-- ..
m 5
1.7500 - --
1.5000 - --
1,2500 - --
1.000C
. e d & 4 4 & &
Time (Sec)
FIGURE 5: MILLSTONE 2 LOSS OF FLOW DNBR VERSUS TIME 22
1100.O ; ; ; ;
-4 1000.00 - - 4 q -
U m
~
l 800.00 - -
1-Y
?
O 8 600.00 - --
E 4
a: -
W 400.00 - "
h U
y Aut1ED STEAM GENERATOR ,l
$ 200. 00 - - \ ~
t i
/ '
0.0 =*
5 "
o 5
o o 5
o 6
o 6 o
' o o
.o 8
8 n.,
8 m 8, 8 8 8 n w ~
TIME (SEC) l FIGURE 6: MILLSTONE 2 STEAMLINE RUPTURE STEAM GENERATOR PRESSURE VERSUS TIME l
l 23 l
/
. 600.00 ; ; 0 ~
550.00 - - "
l C 500.00 - - -
o 8 '
450.00 - - ~'
w l
E 400.00 - -
\ ~
a: ,
" \
350.00 - -
'\ FAULTED STEAM GENERATOR M
= 300.00 - - -
250.00 - ~
200.00 f f f f 4 a o o o o o o o
. o 8
8
~
8 m 8, 8 o
8 w
8
~
TIME (SEC)
FIGURE 7: MILLSTONE 2 STEAMLINE RUPTURE INLET TEMPERATURE VERSUS TIME b
24
l}@@: 0 .. ,,
2000.0 ' ..
~
- 1750.0 - ..
g 1500.0 - ..
e m
W 1250.0 -- ..
u.
E g 1000.00 - ..
ii' s
$ 750.00 - ..
W 500.00 - ..
l 250.00 - ..
l 1
100.00 0.0 + 8 8 8 8 h h g f
- o 'o o o o 6 o
o e
o o
- o o m
o o, c w i o - ru ~
f TIME (SEC) .
FIGURE 8: MILLSTONE 2 STEAMLINE RUPTURE PRESSURIZER PRESSURE VERSUS TIME i
i 25 l
l
)
.10000 ; ; ; ; ; : : 4
.07500 - -
G .05000 - -
b g
.02500 - -
b 0.O ~
P v
a:
.02500 < -
J 5 .05000 - -
R
.07500 - -
- 10000 '
6 6
6 o o o 6 6 6 o o o o c) o o 6 o o o o o
. O k k h $ O k TIME (SEC) l FIGURE 9: MILLSTONE 2 STEAMLINE RUPTURE TOTAL REACTIVITY VERSUS TIME 26
1 l
1.2000 : : :
V y
- 1.0000 - -
g .
o v
(
E_
.80000 - -
w
- s d
- .60000 - -
! W w
$ .40000 - -
=
- w w .20000 - -
g v
0.0 . . . >
g 6 o
6 o
6 o
6 o
6 o
o o
O g o o
o o
o o
ci o
e o
ci o
o _ N m e c w ~
e TIME (SEC)
I l
i -
l FIGURE 10: MILLSTONE 2 l STEAMLINE RUPTURE CORE AVERAGE HEATFLUX VERSUS TIME l
l -
i l
I
\
l l 27 L _ . _ _ _