ML20033A519
| ML20033A519 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/17/1981 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML20033A517 | List: |
| References | |
| TAC-47471, TAC-47472, TAC-54199, NUDOCS 8111250526 | |
| Download: ML20033A519 (39) | |
Text
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TABLE OF CONTENTS, Section Titie Page J.0 INTRODUCTION AND St#4!!ARY l
1.1 OBJECTIVES 1
1.2 GENERAL DESCRIPTION 1
1.3 CONCLUSI0lls 2
, 2.0 11ECHNIICAL DESIGil 3
2.1 GENERAL DISCUSSI0ll
-3 3.0 THER!1AL AND HYDRAULIC DESIGil 4
4.0 flVCLEAR DESIGil 5
5.0 ACCIDENT ANALYSIS 6
5.1 INTRODUCTION
AllD Sil1 MARY 6
5.2 ACCIDENT EVALUATION 6
5.2.1 KINETICS PARNIETERS 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 BORON DILUTION 8
5.3.2 CEA EJECTION INCIDENT 9
5.3.3 CEA WITHDRAWAL FR0!4 SUBCRITICAL 9
5.3.4 CEA WITHDRAWAL AT POWER 10 5.3.5 COMPLETE LOSS OF REACTOR COOLANT FLOW 10 5.3.6 SEIZED ROTOR 11
6.0 REFERENCES
12 8111250526 811117 PDR ADOCK 05000336 P
PDR 0291t
LIST OF TABLES Table-Ti tle Page 13 1
Millstone Unit 2 Cycle 5 Core Loading 2-Millstone Unit 2 Kinetics Characteristics 14 15 3
Shutdown Requirements and fergins 4
Paraueters Used in the Analysis of the 16 CEA Ejection Accident 17 5
Sequence of Events, CEA Ejection Incident 6
Sequence of Events, CEA Withdrawal from Subcritical 18 7
Parameters Used in the CEA Withdrawal Analysis 19 20 8
Sequence of Events-Loss of Coolant Flow 21 9
Summary of Results
, Seized Rotor Transient LIST OF FIGURES Figure Title Page 22 1
Core Loading Pattern 2
Millstone 2 - Safety Analysis Rod Ejection 23 Incident - HZP/BOL Huclear Power Versus Tiue i
3 Millstone 2 - Safety Analysis Rod Ejection 24 Incident - HZP/BOL Fuel and Clad Temperature Versus Time 4
Millstone 2 - Safety Analysis Rod Ejection 25 Incident - HFP/ Nuclear Power Versus Time 5
FHilstone 2 - Safety Analysis Rod Ejection 26 Incident - HFP/ Fuel and Clad Temperature Versus Time 11 0291 t I
o LIST OF TA8tES (Cont'd.)
Table Title Page 27 6.
Millstone 2 - Safety Analysis CEA Withdrawal From Subcritical - Nuclear Power Versus Time
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28 7.
Millstone 2 - Safety Analysis CEA Withdrawal From Subcritical - Heat Flux Versus. Time 29 8.
Millstone 2 - Safety Analysis CEA Wi thdrawal From Subcritical - Fuel, Clad, and Coolant Teuperatures Versus Time 9.
Millstone 2 - Safety Analysis CEA Withdrawal At Power 30 31 10.
Millstone 2 - Safety Analysis Loss of Flow -
Reactor Coolant Flow Versus Tiue 11.
Millstone 2 - Safety Analysis Loss of Flow -
32 Nuclear Power and Heat Flux Versus Time 33 l
12.
Millstone 2 - Safety Analysis Loss of Flow -
DUB Ratio Versus Time 34 13.
Itillstone 2 - Safety Analysis Seized Rotor -
Reactor Cooiant Pressure Versus Time 14.
Millstone 2 - Safety Analysis Seized Rotor -
35 RCS Flow Versus Time 15.
Millstone 2 - Safety Analysis Seized Rotor -
36 Nuclear Power and Heat Flux Versus Time t
l iii l
0291 t
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1.0 IllTRODUCTIO;l A;1D SUM %RY l.1 OBJECTIVES This report presents an evaluation for I4illstone fluclear Power Station Unit 2, Cycle 5, which demonstrates that the core reload will not adversely affect the safety of the plant. This evaluation was accomplished utilizing the methodology described in Reference 1 Based upon the above referenced methodology, only those incidents analyzed and reported in the Basic Safety Report (BSR) which could potentially be affected by fuel reload have been reviewed for the Cycle 5 design described herein.
The results of neu analyses are included and the justification for the applicability of previous results for the remaining incidents is provided.
1.2 GEtlERAL DESCRIPTI0il The Millstone II reactor core is comprised of 217 fuel assemblies arranged in the configuration shoun in Figure 1.
Each fuel assembly has a skeletal structure consisting of five (5) zircaloy guide thimble tubes, nine (9) Inconel grids, a stainless steel botton 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 cf slightly enriched uranium dioxide ceramic pellets contained in Zircaloy-4 tubing which is plugged and seal welded at the ends to l
encapsulate the fuel.
Ilominal core design parameters utilized for Cycle 5 are as follows:
[
Core Power (ihit) 2700 System Pressure (psia) 2250 Reactor Coolant flou (GPM) 370,000*
Core Inlet Temperature ( F) 549 l
J Average Linear Power Density-(ku/ft) 6.065 (based on best estimate hot, densified core average stack height of 134.4 inches) i l
- llinimum guaranteed safety analysis flou 1
l 0291t l
The core loading pattern for Cycle 5 is shown in figure 1.
Twenty-four (24) interior feed assemblies containing 2.7 w/o U235 and forty-eig..
(48) peripheral feed assemblies containing 3.2 w/o U235 are replacing The seventy-two (72) Combustion Engineering (CE) batch D assemblies.
batch B assembly in Cycle 4 is replaced by a batch B assembly that was discharged at the end of Cycle 1.
A summary of the Cycle 3 fuel inventory is given in Table 1
1.3 CONCLUSION
S From the evaluation presented in this report, it is concluded that the Cycle 5 design does not result in the previously acceptable safety limits for any incident to be exceeded. This conclusion is based on the following:
9 1.
Cycle 4 expected burnup of 11,050[40 MWD /MTU.
There is adherance to plant operating limitations as given 2.
in the Technical Specifications.
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0291t
2.0 MECHANICAL DESIGN 2.1 GENERAL DISCUSSION The mechanical design of the Cycle 5 fuel assemblies is essentially fdentical to that of the Cycle 4 assemblies. The Westinghouse fuel assemblies are designed to be fully compatible with all resident Millstone 2 fuel assemblies and core components-(e.g adequate clearances for insertion of CEA's, plugging devices, etc.). Table 1 summarizes pertinent design parameters of the various fuel regions.
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3 0291t L
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3.0 THERMAL AND HYDRAULIC DESIGN A description of the thermal and hydraulic design of the Westinghouse Millstone Il reload fuel assembly to be utilized in Cycle 5 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 11 fuei assemblies.
No significant variations in thermal margins will result from the Cycle
. 5 reload. The Cycle 5 analysis takes a partial credit of 3.0% of the net conservatism which exists between convoluting and summing the uncertainties of various measured plant parameters in power space.
This partial credit was previously applied in Cycle 3 and Cycle 4 and is discussed in more detail in the Cycle 4 Reload Safety Evaluation Report (Reference 6).
-o 0291t 4
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i 4.0 NUCLEAR DESIGN j
The Westinghou e nuclear design procedures, computer programs,'and calculation models utilized in the Millstone II, Cycle 5 reload oesign
}-
are presented in the BSR. With one exception, no changes or modifica-tivas ware required for Cycle 5 evaluation. The exception relates.to the local' power density trip setpoint. methodology.
Cycle 5 accident
)
simulations take credit for the variable high power trip by_ terminating accidents 5% above the variable high power trip. Also P values (see L
]
BSR Section 6.0) are computed only if the maximum allowed power censity
, of 21 kw/ft is exceeded.
3 The Cycle 5 core loading results in a maximum linear heat rate-of less I'
f-than 15.6 kw/ft at all fuel heights at rated poier.. Table 2 provides a l
summary of changes in the Cycle 5 kinetics-characteristics compared with the current limit based on the reference safety analysis ( ).
It.
can be seen from~the table that all of the Cycle 5 values fall within current-limits, except as discussed in Section 5.3.
Table 3 provides the contol 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.
0291t 5
5.0 ACCIDENT ANALYSIS 5.1. INTRODUCTION AND
SUMMARY
-The power capability of Millstone II is evaluated considering.the consequences of those incidents examined in the BSR(2)
, using the
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previously accepted design basis.
It is concluded that the core-reload
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will not adversely affect the ability to safely operate at 100% of rated power during Cycle 5.
For the overpower transient, the fuel
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centerline temperature limit of 4700 F can be accommodated with
' margin in the Cycle 5 core. The burnup dependent densification model(3) was used for fuel temperature evaluations.. The LOCA limit at or below peak linear at rated power can be met by maintaining Fq heat rates at.or below 15.6 kw/ft.
5.2 ACCIDENT EVALUATION The effects of the reload on the design basis and postulated incidents J
analyzed in the BSR(2) were examined.
In most cases, it was found l
that the effects were accommodated within the conservatism of the l -
initial assumptions used in the BSR safety analysis. For those incidents which were reanalyzcd, it was determined that the applicable design bases are not exceeded, and, therefore, the conclusions presented in the BSR are still valid.
I A core reload can typically affect accident analyris input parameters core kinetic characteristics, shutdown margin, in the following areas:
CEA worths, and core peaking factors. Cycle 5 parameters in each of these areas were examined as discussed below to ascertain whether new accident analyses were required.
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6 0291t
5.2.1 KINETICS' PARAMETERS
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-A comparison of Cycle 5 kinetics parameters with the current limits, established by the BSR and Cycle 4 safety analyses,' is presented in
' Table 2.
The parameters in Table 2 which exceeded the limiting range of values established by the Cycle 4 and BSR safety analyses were the most positive moderator temperature coefficient between.70% and 100%
power. operation, the maximum delayed neutron fraction, the Doppler-temperature coefficient and the maximum differential rod worth at HZP.
In addition, the total trip reactivity curve as a function of position and the Doppler power coefficient as a function of power calculated for
, Cycle 5 were more limiting than that calculated for Cycle 4.
5.2.2 SHUTDOWN MARGIN Changes in min. mum shutdown margin requirements may impact the safety analyses, particularly the steamline break and boron dilution accident Table 3 shows the change in shutdown margin requirements for analyses.
Cycle 5.
5.2.3 CEA WORTHS Table 3~shows that Changes in CEA worths may affect shutdown margin.
the Cycle 5 shutdown margin requirements are satisfied.
5.2.4 CORE PEAKING FACTORS s
Peaking factor evaluations were performed for rod out of. position, and steam line break accidents to ensure that the DNB design limits are not l
, exceed'ed. These evaluations were performed utilizing the existing transient statepoint information from the reference' cycle-(BSR) and In each case,
-peaking factors determined for the reload core design.
it was found that the peaking factor for Cycle 5. yielded results that Nere within the DNB design limits.
Consequently, for these accidents
{
no further investigation or analysis was required.
CEA peaking factors for Cycle 5'were within the reference cycle limits.
7 0291t
x A
f 5.3 INCIDENTS REANALYZED
- 5. 3.1 - BORON DILUTION The shutdown margin requirements for Cycle 5 are more limiting than
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lhoseforCycle4. -Therefore, the amount of. operator action time r
available before a complete; loss of sliutdown margin occurs.during a boron dilution event is decreased in several cases for Cycle 5.
No changes from the Cycle 4 critical boron concentration were calculated for Cycle.5. The boron' dilution accident was reanalyzed for the startup, hot standby, and hot. shutdown operating modes, which reflected a' change in' shutdown margin..The startup analysis also. incorporated Cycle 4 (and Cycle 5) critical baron concentrations. The Cycle 4 analysis for startup conditions used very conservative, bounding-
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critical boron concentrations. The hot standby and hot shutdown analysis also assumed the RCS volume available for dilution'during shutdown cooling operation.
Shutdown Margin Requirements (%ap)
Cycle 4 Cycle 5-Power Operation Startup
-S.2
-2.9 Hot Standby
-3.2
-2.9 Hot Shutdown
-3.2
-2.9 Cold Shutdown
-2.0
-2.0 Refueling
-5.0
-5.0 For all cases, there was sufficient time for operator action to terminate the baron dilution before shutdown margin is lost.
Results Startup 64 min. to lose shutdown margin Hot Standby 24 min.
Hot Shutdown 24 min.
W)lR -
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5.3.2 CEA EJECTION INCIDENT The hot zero power case was re-analyzed for Cycle 5 because the Fq after CEA ejection is higher than the Fq used in the BSR analysis. The hot full power case was reanalyzed to incorporate the change in
-positive moderator coefficient betwaen 70 and 100 percent power.
Table 4 gives the key analysis parameters assumed in the analysis.
The sequence of events for this accident is given in Table 5.
The nuclear power transient and hot spot fuel and clad temperature tran-
, sients are shewn in Figures 2-5.
The results demonstrate that the limiting criteria for this accident are not exceeded. The average enthalpy of the hottest fuel pellet does not exceed the clad damage threshold of 200 cal /gm.
5.3.3 CEA WITHDRAWAL FROM SUBCRITICAL The CEA withdrawal from subcritical accident was reanalyzed for Cycle 5 due to the change in maximum differential rod worth of two CEA groups moving together at HZP and changes in the trip reactivity curve, delayed fraction, and Doppler power coefficient.
Table 6 gives the time sequence of events for this accident. The nuclear power transient, heat flux transient, and fuel, clad, and coolant temperature transients are given in Figures 7, 8, and 9.
Results of this analysis show that the peak heat flux and fuel and coolant temperatures remain well below nominal full power values.
Therefore, the DNBR is greater than the limiting value of 1.30 and no cladding damage or release of fission products to the Reactor Coolant System will result.
0291t 9
4 5.3.4 CEA WITHDRAWAL AT POWER The CEA withdrawal at power accident was reanalyzed for Cycle 5 due to the change in maximum differential rod worth of two CEA groups moving together at HZP and changes in the trip reactivity curve, delayed
' neutron fraction, and Doppler power coefficient.
Table 7 gives the key parameters assumed in the analysis.. Figure 10 shows the minimum DNBR as a function of reactivity insertion rate from initial full power operation for toth minimum and maximum reactivity feedback 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.
5.3.5 COMPLETE LOSS OF REACTOR COOLANT FLOW The loss of flow accident was reanalyzed for Cycle 5 as a result of the change in the trip reactivity curve for Cycle 5 and changes in the delayed neutron fraction and Doppler power and temperature coefficients.
Table 8 gives the time sequence of events for this accident. The reactor coolant flow, nuclear power, heat flux, and DNB transients are shown in Figures 11, 12 and 13.
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 l.30 during the transient.
[
0291t 10
5.3.6 SEIZED ROTOR The seized rotor transient was reanalyzed for Cycle 5 due to the change in the trip reactivity curve as a function of rod position and other changes in the delayed neutron fraction and Doppler power and tsmperature coefficients.
Table 9 gives a summary of results and time sequcnce of events for the seized rotor transient. The pressure, flow, nuclear power, and heat flux transients are provided in Figures 14, 15 anc 16.
The results show that the integrity of the primary coolant system is not endangered since the peak reactor coolant system pressure is much The less than that which would cause stress limits to be exceeded.
peak clad temperature is much less than 2700 F which guarantees that the core will remain intact with no loss of core cooling capability following this transient. The number of rods conservatively calculated to experience DNB is less than 2 percent.
5.3.7 Steamline Break Results for this transient will be supplied later.
11 0291t
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 0 prating, February 12, 1979 6.
Letter, Counsil to Clark, Millstone Nuclear Power Station Unit No.
2, Cycle 4 Refueling - Reload Safety Analysis, June 3, 1980 l
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l 0291t 12
TABLE 1 Millstone Unit 2 Cycle 5 Core Loading Initial 80C Number of Enrichment
% Theoretical Burnup Average Reoion Type Assemblies w/o U235 Densa (MWD /NTU)
B+
CE 1
2.336 95 17450 El CE 24 2.730 94.75 24650 E2 CE 48 3.235 94.75 22600 F1 W
24 2.697 94.54 13470 48 3.297 94.87 9650 F2 W_
G1 W
24 2.70 95*
0 G2 W
48 3.20 95*
0
- The Region G1 and G2 densities are nominal. Average densities of 94.5 theoretical were used for Region G1 and G2 nuclear design evaluations.
0291t 13
TABLE 2 MILLSTONE UNIT 2 XINETICS CHARACTERISTICS Current Limit (Cycle 4)
Cycle 5 Most Positive Ibderator Temperature Coefficient (ap /*F) x 10-4
+0.5 from 0 to 70% Power
+0.5 from 0 to 70% Power
+0.2 from 70 to 100% Power
+0.4 frou 70 to 100; Power Most lbgative Ibdcrator Teaperature Coefficient (an/*F) x 10~4, ARI
-3.8
-3.8 Doppler Temperature Coefficient (ap/ F x 10-5)
-1.2 to -1.87
-1.2 to -1.92 M
. m to. m MDE Delayed Ibutron Fraction Beff Prompt lbutron Fraction (psec)
<32.2
<32.2 Maximum Differential Rod Worth of two CEA groups moving together
<24.3 36.6 at HZP (pcm/in) t 0291t 14
TABLE 3 SHUTDOWN REQUIREMENTS AND MARGINS MILLSTONE UNIT 2 - CYCLE 5 Control Rod Worth (%Ap)
_B0 EOC All Rods Inserted 8.24 9.11 7.84 8.67
. All Rods Inserted Less Worst Stuck Rod 6.78 7.02 6.47 6.59 (1) Less 10 Percent 6.10 6.32 5.82 5.93 Control Rod Requirement _s Reactivity Defects (Combined Doppler, T, g, Void and Redistribution Effects) 1.71 2.62 1.94 2.64 Rod Insertion Allowance 0.36 0.36 0.36' O.36 (2) Total Requirements 2.07 2.98 2.30 3.00 Shutdown Margin ((l) - (2)) ( ap) 4.03 3.34 3.52 2.93 Required Shutdown Margin ( ap) 3.20 3.20 2.90 2.90 l
I 0291t 15
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TABLE 4 PARAMETERS USED IN THE ANALYSIS OF THE CEA EJECTION ACCIDENT Par'ameter HZP HFP Power Level, 0
102 Ejected ' rod worth, percent ap 0.65
.17
, Delayed neutron fraction, 0.47 0.47 Feedback reactivity weighting 2.50 1.30 Trip reactivity, percent ap 2.55 4.17
-4
-4 Moderator temperature coeff.icient, ap/ F
+0.5410
+0.4x10 F before rod ejection 2.64 q
F after rod, ejection 19.1 5.70 q
Number of operational pumps 2
4 Results Max. fuel pellet average temperature, F
3090 3930 Max. fuel center temperature, UF 3563 4983 Max. clad average temperature, F
2343 2345 Max. fuel pellet center melting, percent 0
2.3 Max. fuel stored energy, cal /gm 128 171 0291t 16
TABLE 5 SEQUENCE OF EVENTS, CEA EJECTION INCIDENT - HZP Time Event Setpoint or Value 0.0 Initiation of Transient 0.1 CEA Fully Ejected 0.28 High Power Trip Signal Generated 25 percent 0.35 Peak Nuclear Flux Reached See Fig. 3 1.18 CEA Insertion Begins 2.84 Peak Fuel Temperature Reached See Fig. 4 3.78 CEA's Reach 90 nercent Insertion SEQUENCE OF EVENTS, CEA EJECTION INCILENT - HFP Event Setpoint or Value 0.0 Initiation of Transient
.04 High Power Trip Signal Reached 112 percent 0.1 CEA Fully Ejected
.13 Peak Nuclear Flux Reached See Fig. 5
.94 CEA Insertion Begins 2.36 Peak Fuel Temperature Reached See Fig. 6 1
i 3.54 CEA's Reach 90 '>ercent Insertion l
0291t 17
TABLE 6 SEQUENCE OF EVENTS, CEA WITHDRAWAL FROM SU8 CRITICAL Setpoint or Value Time Event 0.0 Initiation of uncontrolled rod withdrawal-24.4 pcm/sec reactivity insertion rate from 10-Id of nominal power 25 percent Variable high power trip 26.2 setpoint reached 1.24 x cominal 26.5 Peak nuclear power occurs 27.2 CEA insertion begins 572 F 20.1 Feak average clad temperature occurs 45.2 of nominal 28.2 Peak heat flux occurs 698.0 F 29.0 Peak average fuel temperature occurs 18
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0291t
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TABLE 7 PARAMETERS USED IN THE CEA WITHDRAWAL ANALYSIS Units Cycle 5 Parameter Initial Core Power Level MWt 0-102 percent of 2700 Core Inlet Coolant Temperature F
532-551 Peactor Ecolant System Pressure psia 2200
-4 Moderator Temperature Coefficent 10 ap/ F
-2.5 and + 0.5 Doppler Coefficient Multiplier 1.15 and 0.85
-2 CEA Worth at Trip 10 a
-2.9
-4
- Reactivity Ir.sertion Rate x10 ap/sec 0 to 2.44 Holding Coil Delay Time sec 0.5 CEA Time to 90 Percent Insertion sec 3.1 (including Holding Coil Delay) 0291t 19
TABLE 8 SEQUENCE OF EVENTS - LOSS OF COOLANT FLOW Four pumps in operation, all pumps coasting down Time (sec)
Event 0.0 Loss of power to all pumps
.91 Reactor coolant pump low speed setpoint reached 1.56 CEA's begin to drop 3.7 Minimum DNBR occurs l
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20 0291t
F.
TABLE 9
SUMMARY
OF RESULTS FOR SEIZED ROTOR TRANSIENT Maximum Reactor Coolant System 2553 Pressure (psia)
Maximum Clad Temperature (*F) 1969 Core Hot Spot our loops in operation, one F
seized rotor Event Time (sec)
Rotor on one pump seizes 0
Low flow trip point reached
.04 Rods begin to drop 1.24 Maximum clad temperature occurs 3.6 Maximum RCS pressure occurs 3.8 0291t 21
Figure 1 Core Loading Pattern Millstone Unit 2 - Cycle 5 ~
G HJ K LM N P-R A
B C
D E
F
'l l l 1
l S
T V.
W X
Y 1
G2 G2 G2 G2 ss 2
G2 G2 G2 F?
F9 E2 G2 G2 G7 G2 F1 F2-E2 G1 El G1 E2 F2 F1 G2 3
4 G2 F1 F2 El F2 E2 F2 E2 F2 El F2 F1 G2 G2' F1 F2 E2 G1 E2 F2 F1 F2 E2 G1 E2 F2 F1 J G2
-6 G2 F2 El G1 E2 F2 El E2 El F2 E2 G1 El F2 G2 7
G2 E2 F2 E2 F2 El G1 E2 G1 El F2 E2 F2 E2 G2 G2
,8-E2 G1 E2 F2 El G1 F1 F1 F1 G1 El F2 E2 G1 E2
-9 G2 ss s_s G2
-10 F2 El F2 F1 E2 E2 Fl B+
F1 E2 E2 F1 F2 El F2 11 G2 BP G2
-12
- -_ j 3 G2 E2 G1 E2 F2 El G1 F1 F1 F1 G1 El F2 E2 G1 E2 G2
-14' 5
G2 E2 F2 E2 F2 El G1 E2 G1 El F2 E2 F2 E2 G2 1
16 G2 F2 El G1 E2 F2 El E2 El F2 E2 G1 El F2 G2 j
I G2 F1 F2 E2 G1 E2 F2 F1 F2 E2 G1 E2 F2 F1 G2'
)
18 G2 F1 F2 El F2 E2 F2 E2 F2 El F2 F1 G2 G2 El F2 E2 G1 El G1 E2 F2 F1 G2 19 ss 4
G2 G2
, G2 E2 F2 E2 G2 G2 G2 20 1
21 G2 G2 G2 G2 initial Region Type _
w/o U-235 North Reactor Core B+
CE 2.336 El CE 2.730 SS - Secondary Source CE M3 BP - Burnable Poison F2 W
3.297 22 G1 W
2.70 G2 W
3.20
.~w.~.~..~d
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ww 17 500 -
15 000 ---
=s 12.500 --
o C
5 10.000 --
w mu 5
7.5000 -
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5.0000 --
2.5000 --
,u 00 o
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l Figure 2: Millstone 2 CEA Ejection - HZP Nuclear Power Versus Time l
l 23
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5000 0
'M Q.
3 3...
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.=-
= =.
5
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rim ivo Figure 3: Millstone 2 CEA Ejection - HZP Fuel and Clad Temperatures Versus Time 24
e 6000 3 2::* 0 -
E 3 *:00 3 -
r
.i
- 3
- : -
="
E..
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rim ivo 5000 3 5 :" 0 <
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- 5. 2000 0 -
3 l.
u 1300 03 <
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h l l w
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l Figure 5: Millstone 2 CEA Ejection - HFP Fuel'and Clad Temperatures Versus Time r
26
g g.
3.0000 --
2.0000 --
1 m
w,, :
3 o
a.
,g x<
.30000 --
wcj
.20000 --
Oz
.'03000 -
.02000 -
.01000 O
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O O
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O O
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<u m
_e n
m o
TIME (SEC)
Figure 6: Millstone 2 CEA Withdrawal from Subcritical Nuclear Power Versus Time l
l 27 l
ll -.
1.0000
_.80000 -
x
.60000 -
3 w
H<
w*
.40000 -
.20000 -
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TIME (SEC).
Figure 7: Millstone 2 CEA Withdrawal from Subcritical Heat Flux Versus Time 28
900.00 '
850.00 -
2 800.00 -
a._
Ew 750.00 -
~~
700.00 -
el b g Teg 650.00 -
lad Inner Temp 600.00 -
ore Water Temp 550.00 -
500.00 i
O.
O O
O O
O O
O O
O O
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O O
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O d
6 6
6 d,
Oo e
O
<u m
TIME (SEC) l Figure 8: Millstone 2 CEA Withdrawal from Subcritical Fuel, Clad and Coolant Temperatures Versus Time 4
29
0 0
1 0
8 0
6 0
4 0
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L t
O a
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d s 2
l h v a
t y
w i R t
2a W B i
r N
y v
ed A D r
i nh E
a t
ot C m s
c ti u
s a
sW m
e 1
e l
i c
R lA n
e iE 8
MC N
i M
p i
6 r
9 T
e o
r N
u 4
g i
4 F
2 1
8 7
6 5
1 1
1 1
EE5lp
e 1.0000 -
.80000 -
E <'
a'E y g
.60000 -
5=
o u 8o
~
.40000 -
-o o e
< w w
Q.
M
.20000 -
4 5
5 5
o 0.0 4
o e
o o
o o
o a
o o
o o
o o
w
=
6 a
a TIME (SEC)
Figure 10: Millstone 2 Loss of Flow Reactor Coolant Flow Versus Time 31
l t
t.1000 1.0000 - -
.a0000 g
5{E 3:
.60000 -
5 g, 55.0000 E
20000 -
t' i
s.o I
i 1.1000 g 1.0000 --
l y-Avg Channel g
.80000 l
g E=
W
.60000--
E
.40000 -
l Ed
.20000 l
E l
T o.0 t.1000 y 1.0000 E
Hot Channel g
.80000
(
g E=
M.60000 1
E i
4000C <-
f E
d.20000 -
E W
8 8
g
8 g
E E
o 6
e 6
4 w
a TI4 ISEC1 l
figure 11: Millstone 2 Loss of Flow Nuclear Power and Heat Flux Versus Time
._ _.. _ _3 2 _._, _,
o 3.0000 2.7500-2.5000-2.2500--
3 4.J E 2.0000--
e
- E 1.7500--
1.5000-1.2500-1.0000 8
8 8
8 8
8 8
8 8
8 0
d J
d Time (Sec) t t
Figure 12: Millstone 2 Loss of Flow DNB Ratio Versus Time 33 L
o-
/
=2600 2500 2400 7;
/
8 2300 E
E c.
2200 2100 2000 0
2 4
6 8
10' Time (sec)
Figure 13: Millstone 2 Seized Rotor Reactor Coolant Pressure Versus Time I
I 34
(
lt.
o
+
1 o
1.0p00 l
.80000 -
m o> <a aw z
$ [
.60000 -
1-o w o o u
H E5
.40000 -
-a u =
< w w CL
=
.20000 - -
0.0 o
e o
o o
o o
e o
o e
o e
o o
o o
o o
a
.e o
<u w
m TIME (SEC) t r
Figure 14: Mil' stone 2 Seized Rotor RCS Flow Versus Time 35
1.1000 j
1.0000 --
a
.80000 --
z_ z wn X o o I
z a
.60000 --
m
= a
<W z
' o MC z o
.40000 --
4 E
.20000 --
0.0 1.1000 a
1.0000 -
<5 Ioz
.80000 -
uo z
i o
W
.60000 -
E
.40000 -
x3
.20000 --
wz 0.0 o
o o
o o
a o
a o
o o
a o
a o
S 9
9 9
e o
m a
w o
TIME (SEC)
Figure 15: Millstone 2 36 Seized Rotor Nuclear Power and Heat Flux Versus Time
-