ML20213D856
| ML20213D856 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 11/05/1986 |
| From: | Allen C COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 2369K, NUDOCS 8611120291 | |
| Download: ML20213D856 (39) | |
Text
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} One First National Plaza. ChicaDo, lilinois
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Commonwealth Edlaon
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Address Reply to: Post Offce Box 767 b
Chicago, Illinois 60690 - 0767 November 5, 1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
LaSalle County Station Units 1 and 2 Proposed Technical Specification Amendment to Move MSIV Iso!ation Signal to Level 1
,NRC Docket Nos. 50-373 and 50-374 Reference (a): Letter dated October 21, 1986 from C.M. Allen to H.R. Denton.
Dear Mr. Denton:
Reference (a) transmitted Commonwealth Edison's proposal for a r
Technical Specification change to move the setpoint for Main Steam Line Isolation Valve (MSIV) closure from Level II to Level I.
That transmittal referenced a General Rlectric Service Information Letter (SIL No. 367) and a Quadrex Report (1-83-008). These items are included as attachments to this letter for use by your staff.
Please direct any questions you may have regarding this matter to this office.
Very truly yours, C. M. Allen Nuclear Licensing Administrator in Attachments cc:
Dr. A. Bournia - NRR Resident Inspector - LSCS M. C. Parker - IDNS 8611120291 861105 2369K PDR ADOCK 05000373 P
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@@$EI. INFORMATIOM LETTER h
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' CLEAR FUEL ANo SERVICES OtytssON e
SAN JOSE, CAUFORNIA 93125 December, 1981 SIL No. 367 File Tab B Category 3 RPV WATER LEVEL SETPOINT FOR MSIV CLOSURE Generator load rejection and turbine trip may result in main ste.am isolation valve (MSIV) closure due to low water level in the reactor pressure vessel (RPV). The purpose of this Service Infomation Letter is to recomend that the MSIV nominal RPV low water level trip setpoint be lowered to reduce the probability of MSIV closure (reactor isolation).
DISCUSSION In the past the MSIV nominal trip sepoint for RPV low water level has generally been at RPV Level 2.
BWR/6 reactors have a MSIV nominal trip set-point at RPV Level 1, so the necessary review and analytical work has been done for these reactors. This also requires that the pnuematic supply to MSIVs be isolated at Level 1 and not at some higher level on plants having this feature.
Reducing the probability of reactor isolation from normal operating power levels may reduce safety / relief valve operations, release of energy to the suppression pool and loss of feedwater flow for plants with steam driven feedwater pumps. Also, reducing the probability of losing the main condenser as the primary RPV heat sink will help reduce forced outage time. The BWR Owners' Group and the Mark I Owners' Group have already been informed of this recommendation and have endorsed it.
[
RECOMMENDATIONS l
General Electric recomends that BWR owner / operators consider lowering the MSIV (reactor isolation) nominal RPV low water level trip setpoint to RPV Level 1 for BWR/4 and BWR/S reactors. Additional emergency core cooling analytical work would be required to make the associated licensing change.
For additional infomation contact your local General Electric service representative.
Prepared by:
- 0. J. Foster /D. R. Heising IN Approved by: MM//dd Issued by:
D. K. Willett, Manager D. L. Allred, Manager Customer Service Customer Service Information Product
Reference:
B21 - Nuclear Boiler System GENERAL $ ELECTRIC f
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- QUAD-1-83-008 FOR INFORMATION ONLY I
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COMMONWEALTH EDISON CO.
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BYJ.:.L.W9.K.tX DATE J/~ 1_9-13 P, o, 42G7 (G684-0o)
SPEC. NO 804776 PROJ. NO.
ared for:
'T-O_90183 wnnumwtauH EDISON COMPANY Chicago, Illinois 60690 Prepared by:
l QUADREX CORPORATION 1700 Dell Avenue Campbell, California 95008 Prepared by:
Approved by:
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Revision No.
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8/24/83 H. E.
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QUAD-1-83-008 t
ABSTRACT This report provides the analysis for the effect of changing the Main Steam Isolation Valve (MSIV) closure from water level 2 to water level 1 on the LaSalle County Station Unit 2.
Loss-of-coolant accident (LOCA) analyses were performed for three break sizes as defined in the FSAR.
They are the design basis accident (DBA) large break, the small recircu-2 lation suction break (0.09 ft ), and the feedwater line break. The RETRAN-02 computer code was used.
RETRAN-02 is a state-of-the-art computer code, widely used in the nuclear industry for transient thermal-hydraulic analysis of light-water reactor systems.
In this analysis the verification of the RETRAN-02 model was established by comparing RETRAN-02 results with results from LaSalle FSAR for the DBA case.
Based on this study, it is concluded that the change in MSIV closure caused by lowering the MSIV isolation setpoint to water level 1 has insignificant impact to loss of coolent events (FSAR chapters 6 and 15) in terms of reactor vessel pressure, core water inventory, suppression pool temperature, and fission product release to the environment.
4 9
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QUAD-1-83-008 TABLE OF CONTENTS P_ age ABSTRACT i
LIST OF TABLES iv LIST OF FIGURES v
- 1. 0 INTRODUCTION 1-1 1.1 Approach Used in this Study 1-1 1.2 Conclusions 1-1
2.0 DESCRIPTION
OF ANALYSIS 2-1 2.1' Selection of Break Sizes 2-1
- 2. 2 Analysis Method 2-1 2.2.1 Computer Code 2-2 2.2.2 Model Description 2-3 2.2.2.1 Large/Small Break 2-3 2.2.2.2 Feedwater Line Break 2-3 2.2.3 Model Verification 2-3 3.0 RESULTS OF ANALYSIS 3-1 3.1 DBA, large Break 3-1 3.1.1 Vessel Pressure 3-1 3.1.2 Core Inventory 3-1 3.1.3 Suppression Pool Temperature 3-2 3.1.4 Fission Product Release 3-2 3.2 Small Break 3-2 3.2.1 Vessel Pressure 3-2 3.2.2 Core Inventory 3-3 3.2.3 Suppression Pool Temperature 3-3 3.2.4 Fission Product Release 3-3
-ii-
QUAD-1-83-008 TABLE OF CONTENTS (Continued)
Pa e 3.3 -Feedwater Line Break 3.3.1 Vessel Pressure 3-5 3.3.2 Core Inventory 3-5 3.3.3 Suppression Pool Temprature 3-6 3.3.4 Fission Product Release 3-6
4.0 REFERENCES
4-1 i
-iii-s-
QUAD-1-83-000 LIST OF TABLES Table Pace 2-1 Initial Conditions Used for RETRnN-02 Input 2-5 2-2 Sequence of Events for DBA Large Break Analysis 2-6 2-3 Sequence of Events for SBA, Recirculation Suction Break Analysis 2-7 2-4 Sequence of Events for Feedwater Line Break Analysis 2-8 3-1 Radiological Effects Due to small Break 3-8 3-2 Radiological Effects Due to Feedwater Line Break 3-8
-iv-
QUAD-1-83-008 i
l LIST OF FIGURES Figure M
2-1 RETRAN-02 Model for Large/Small Break LOCA 2-9 2-2 RETRAN-02 Model for Feedwater Line Break 2-10 2-3 Normalized Power vs. Time After Break DBA Large Recirculation Suction Break 2-11 2-4 Core Average Pressure vs. Time After Break DBA large Recirculation Suction Break 2-12 2-5 Core Inlet Enthalpy vs. Time After Break DBA large Recirculation Suction Break 2-13 2-6 Water Level Inside Shroud vs. Time After Break DBA Large Recirculation Suction Break 2-14 3-1 Normalized Vessel Dome Pressure vs. Time After Break DBA Large Recirculation Suction Break 3-9 3-2 Water Level Inside Shroud vs. Time After Break DBA Large Recirculation Suction Break 3-10 3-3 Normalized Vessel Dome Pressure vs. Time After Break Small Recirculation Suction Break, Break Area = 0.09 sq. ft.
3-11 3-4 Water Level Inside Shroud vs. Time After Break Small Recirculation Suction Break, Break Area =
0.09 sq. ft.
3-12 3-5 Normalized Vessel Dome Pressure vs. Time After Break Feedwater Line Break, Outside Containment 3-13 3-6 Water Level Inside Shroud vs. Time After Break Feedwater Line Break Outside Containment 3-14 I
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QUAD-1-83-008
- 1. 0 INTRODUCTION In this study LOCA transient analysis was performed for three selected line breaks for Commonwealth Edison Company (Ceco) LaSalle Station unit 2.
The purpose of the analysis is to establish the differences in
' Nuclear Steam Supply System (NSSS) response to the lowering of the MSIV closure from water level 2 to level 1.
The significance of the differences between MSIV closure at level 1 and at level 2 was assessed in terms of vessel pressure, core inventory, suppression pool temperature and the potential for release of radioactive fission products.
1.1 Approach Used in This Study In this study, state-of-the-art,' best estimate computer code is used to analyze LOCA transient for three selected break sizes. This is to ensure that the delay in MSIV closure, due to lowering of MSIV isolation setpoint, does not impact any of the FSAR chapter 6 or 15. events that assumes MSIV closure at low water level 2.
The three different break sizes considered were selected according to the break spectrum in the FSAR chapter 6 analysis and fission product release considerations in chapter 15 anal-ysis.
The results considered for each break size ine.luded the following items:
1 o
Peak vessel pressure o
Core inventory o
Peak suppression pool temperature o
Fission product release.
[
The results of the MSIV level 1 analysis will be compared to FSAR accepted values to establish the fact that the effect of the MSIV level change is insignificant in terms of reactor safety.
- 1. 2 Conclusions Based on the analysis performed in this study, it is concluded that the delay in MSIV closure caused by lowering of MSIV isolation setpoints to 1
1-1
QUAD-1-83-008 level I has insignificant impact to FSAR chapter 6 and 15 events in terms of vessel pressure, core inventory, suppression pool temperature, and fission product release. This conclusion, therefore, infers that the MSIV level change will have no significant effect with respect to reactor safety.
I l-2
QUAD-1-83-008
- 2. 0 DESCRIPTION OF ANALYSIS 2.1 Selection of Break Sizes Three break sizes are selected for analysis. They are:
the DBA large break, small break, and feedwater line break.
The Design Basis Accident (DBA) case of large recirculation suction break is considered because of its severence in core performance and because of the rapid and large blowdown. The limits on the containment and radiological effects are of great concern. Also, this case, with such fast transient flow conditions, sets up a basis for benchmarking our model, when comparing the results with the FSAR.
Among the break spectrum of the FSAR for small and intermediate breaks (up to 1.0 ft ), the break size of 0.09 ft2 2
on the recirculation suction line gives the highest peak cladding temperature (PCT) of 1736*F.
The slow blowdown rate causes the water level to drop slowly. As a result of this, the differential time for the MSIV closure from water level L2 to L1 is a significant period of 40 seconds.
Thus, it is important to investigate the possible impact on core performance and radiological consequence due to this larger differential time from MSIV closure level change.
The other smaller break from the FSAR spectrum is 0.02 ft2 recirculation suction break. Though it may give an even larger differential time of MSIV closure, the PCT is only 1250*F.
The cases with which the immediate closure of the MSIV is more desirable in reducing the radiological effects to the environment are the main steam line break and the feedwater line break, both being outside the containment. The main steam line break is not analyzed because the MSIV is closed by fast steam line flow rather than by water level. Also, a 2-1
QUAD-1-83-008 feedwater line break releases the largest amount of liquid ~outside the containment and thus provides the envelope evaluation' relative to this type of occurrence.
i 2.2 Analysis Method 2.2.1 Computer Code The computer code used for this analysis was the RETRAN-02 MOD 2 computer code.
RETRAN-02 M002 is a state-of-the-art, best-estmate computer code, widely used in the nuclear industry for transient thermal-hydraulic analysis of light-water reactor systems. The RETRAN computer program is the result of an extensive code develop-ment effort sponsored by Electric Power Research Institute (EPRI) since 1975. The RETRAN-02 computer program is an extension of the RETRAN-01 program designed to provide add-on. analysis capabilities for:
Small break loss-of-coolant accidents o
Balance-of plant modeling o
Anticipated transients without scram (ATWS).
Some significant features of RETRAN-02 that allow the analyses of most BWR transients are one-dimensional kinetics, dynamic slip, vector momentum, separator, and the auxiliary neutron void fraction models.
The LaSalle plant model, based on the RETRAN-02 MOD 2, is described in section 2.2.2.
A detailed description of the RETRAN-02 computer code can be found in volume 1 of reference 1.
The RETRAN-02 MOD 2 computer code verification and quality assurance is described in detail in reference 4.
2-2
QUAD-1-83-008 2.2.2 Model Description In setting up the RETRAN-02 computer model, the main computer modeling references used are the CECO computer model for RETRAN-02 for the LaSalle Plant, and generic GE computer models for LOCA analysis.
In preparing the computer input, the main source refer-ences used are the LaSalle County Station FSAR (reference 2).
GE design specifications, plant drawings, diagrams, and data sheets are supplied by CECO, and Sargent & Lundy. A list of the initial conditions used for this study is given in table 2-1.
The conditions and assumptions used in the computer code modeling follow as closely as possible to those stated in the FSAR.
2.2.2.1 Large/Small Break The RETRAN-02 computer model used for the large break analysis and small break analysis is the same and is shown in figure 2-1.
A total of 28 control volumes, including three core volumes, are used.
There are 47 junctions in the model. The sequence of events for the large break LOCA and small break LOCA analyses are listed in tables 2-2 and 2-3 respectively.
2.2.2.2 Feedwater Line Break The RETRAN-02 computer model used for feedwater line break is shown in figure 2-2.
It is almost the same as figure 2-1 except that there is no break on the recirculation suction.
The turbine and the condenser are represented by a negative fill junction and a conservative constant blowdown rate is assumed for the dose release analysis. The sequence of events is listed in table 2-4.
-2.2.3 Model Verification The verification of RETRAN-02 model used for LOCA analysis in this study is established by comparing RETRAN-02 results with results from FSAR for the DBA large recirculation suction break with MSIV 2-3
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QUAD-1-83-008 closure at the water level L2.
The parameters to be compared are normalized power, core average pressure, lower plenum enthalpy, and water level inside the core shroud.
These parameters are all plotted against time after the break.
Figure 2-3 shows the comparison for normalized power vs. time between RETRAN-02 model and FSAR results. The RETRAN-02 results are almost identical to FSAR results.
Figure 2-4 shows the comparison for core average pressure vs. time between RETRAN-02 and FSAR results.
In both cases, the initial pressure started at about 1070 psia. After 20 seconds, the FSAR model depressurized to about 650 psia and the RETRAN-02 model depressurized to about 740 psia.
This results in a maximum pressure difference of (740 - 650 x 100) % = 13.8% between the two models.
Figure 2-5 shows the comparison for lower plenum enthalpy.
In this case, the maximum percentage difference is (540 - 35.9 x 100)% =
- 0. 76L Figure 2-6 shows the comparison for water level inside the shroud.
There is a difference of about 3 feet in the initial condition of the two cases. However, after about five seconds, the two water levels match pretty well with each other as shown in the figure.
Based on the above comparisons, it is concluded that the results from the RETRAN-02 model agree reasonably well with those from the FSAR for the DBA large break LOCA analysis for the time period considered in this study. Thus, the basis for the model on RETRAN used for the MSIV close level change is substantiated.
2-4
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QUAD-1-83-008 p(2.6.3 TABLE 2-1.
Initial Conditions Used for RETRAN-02 Input 1.
Thermal power level, Hwt 3489 (105% NBR) '
2.
Steam flow, iba per hr 14.87 x 108 (105% NBR) /
3.
Core flow, iba per hr 108.36 x 10e 4.
Feedwater flow rate, iba per sec 4130 5.
Vessel dome pressure, psia 1055.0 /
6.
Turbine bypass ca'pacity, % NBR 25
- 3. 0,
7.
Core coolant inlet enthalpy, BTU per lbe 531.4 8.
Turbine inlet pressure, psia 976.7 9.
Fuel lattice 8x8/
10.
Core bypass flow, %
12.0 7 11.
Scram reactivity FSAR figure 15.0-2 12.
Control rod drive speed, position versus time FSAR figure 15.0-2
- 13. Jet pump ratio, M 2.28 14.
Analytical setpoints for S/R valves Relief function, psig 1076, 1086, 1096, /
1106, 1116 15.
S/R valves capacity, % NBR 110.0 16.
Number of valve groupings simulated Relief function, no.
5 17.
Relief valves function delay, seconds 0.4 18.
High dome pressure ARI scram setpoint, psig 1135.0
Vessel level trips, inches from steam skirt bottom (instrument zero)
Level 8 - (L8) 55.5 Level 2 - (L2)
-50.0 21.
RPT delay, seconds 0.190 /
- 22. Turbine stop valves closing time, seconds 0.15
- 23. MSIV closing time, seconds 4.0
- 24. Turbine bypass valves opening time, seconds 0.27 l
25.
Suppression pool temperature limit, F 200.0 I
l 2-5
QUAD-1-83-008 TABLE 2-2.
Sequence of Events for DBA, Large Break Analysis TIME (sec)
EVENTS 0
Design-basis loss of-coolant accident assumed to start; normal auxiliary power assumed to be lost.
s0 Drywell high pressure and reactor low water level reached.
All diesel generators signaled to start; scram; HPCS, LPCS, LPCI signaled to start on high drywell pressure.
3.15 Reactor low-low level reached. HPCS receives second signal to start.
4.15 Main steam isolation valves close.
4.55 Reactor low-low-low water level reached.
Second signal to start LPCI and LPCS; auto-depresurization sequence begins.
<13 Divisions 1 and 2 diesel generators ready to load; start to close containment isolation valves.
~27 HPCS injection valve open and pump at design flow, which completes HPCS startup; LPCS and LPCI (RHR "C") pumps at i
rated speed.
<32 LPCI (RHR "A" and "B") pumps at rated speed.
<40 LPCI "C" and LPCS pumps at rated flow, LPCI "C" and LPCS injection valves open, which completes the LPCI "C" and LPCS startups.
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2-6
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4 QUAD-1-83-008 TABLE 2-3.
Sequence of Events for SBA, Recirculation Suction Break Analysis TIME (sec)
EVENTS 0
A break of size 0.09 square feet assumed to start.
$1 Feedwater assumed ramped to zero and recirculation pumps tripped.
- 10 Low water level reached and reactor signaled to scram.
s24 Low-low water level reached, MSIV signaled to start closing, reactor scram.
s67 Low-low-low water level reached, LPCI and LPCS signaled to start, ADS initiated.
s86 High drywell pressure reached, LPCI, LPCS received second signal to start.
s90 S/R valves opened to relief vessel pressure.
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s QUAD-1-83-008 TABLE 2-4.
Sequence of Events for Feedwater Line Break Analysis TIME (sec)
EVENTS
$0 Feedwater line outside containment assumed to break, recirculation pumps tripped due to loss of feedwater.
<30 Low-low water level reached, MSIV signaled to start closing, HPCS initiated, reactor scram.
- 120 S/R valves open and close to maintain the reactor vessel pressure at $1100 psia.
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FSAR RETRAIN-02 FIGURE 2-3 NORMALIZED POWER VS. TIME AFTER BREAK DBA LARGE RECIRCULATION SUCTION BREAK l
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QUAD-1-83-008 4
3.0 RESULTS OF ANALYSIS The purpose of the current ECCS analysis is to ensure that the lowering of MSIV level set point from L2 to L1 does not impact any of the chapter 6 or 15 events that assume MSIV closure at level 2.
Current analysis studies 3 cases of LOCA: DBA, small break, and feedwater line break and the results are presented in the following sections.
Based on this analysis, it is ccncluded that the delay in isolation has negligible effect on key parameters, i.e., vessel pressure, core inventory, suppres-sion pool temperature and fission product release. Hence, lowering the MSIV set point will have no influence on the reactor safety performance.
3.1 _DBA, large Break 3.1.1 Vessel Pressure In general, when the MSIV Oloses, the vessel ~ pressure starts to increase to a peak a few seconds later and then afterwards the fast depressurization due to the large break area dominates the pressure response of the transient.
Figure 3-1 shows the plots for vessel dome pressure versus time for both MSIV closure at water level L2 and MISV closure at water level L1.
For the first case, it is shown that when the MSIV starts to close at 4.15 seconds after the break, the dome pressure starts to rise again and then depressuriza-tion dominates a few seconds later.
For the second case, when the MSIV starts to close at 8.0 seconds, i.e., 3.75 seconds later, again the dome pressure increases and then falls off due to depressurization.
3.1.2 Ccre Inventory The water level inside the shroud for the two cases are plotted in figure 3-2 and it is seen that there is insignificant effect up'on the water level of the vessel.
3-1 i
. QUAD-1-83-008 3.1. 3 Suppression Pool Temperature Engineering calculations based on equations ~ supplied in reference 3 )
~
were used'to obtain the maximum suppression pool temperature T.
f The value of T was determined.as 138*F for MSIV closure at L2, and f
T as 137*F for MSIV closure at L1. Thus, the change in MSIV f
closure water level has no significant effect upon the blowdown from the break. The suppression pool temperature continues to rise due to energy sources from decay heat and fuel relaxation and sensible heat of the core structure.
Because the change in MSIV closure water level has no effect upon the blowdoyn rate, then it
- can be concluded that it also has no significant effect upon the other mentioned energy sources after therblowdown, and so the final suppression pool temperature ir 200*F for the mode with one Low Pressure Core Injection (LPCI) working (FSAR table 6.2-5).
)
3.1. 4 Fission Product Release From the RETRAN results, the amount of blowdown from the two cases t
i of MSIV closure are about the same (within one percent for either liquid, steam, or total mass).
As shown in figure 3-2, the change of the MSIV closure water level has no significant impact upon the water level inside the core shroud.
It will not cause any fuel failure with an operational ECCS.
It also will not affect the conservative 100 percent, fuel failure LOCA analysis referenced in FSAR section 15-6, because the time for complete closure of the MSIV, at the latest is 12.0 seconds after the brea'k for water level L1. At that time the fuel cladding temperature is 540*F, just slightly lower than that of expected normal operating conditions.
3.2 Small Break 3.2.1 Vessel Pressure For small sized breaks (area < 0.09 ft ), the depressurization from 2
blowdown is relatively slow compared with the DBA case.
As shown 3-2
QUAD-1-83-008 in figure 3-3, the dome pressure starts to decrease for the first 20 seconds after the break, then it begins to rise again when the MSIV starts to close at 24 seconds with water level L2.
For the second case of MSIV closure at water level L1, the MSIV starts to close 40 seconds later at 64 seconds after the break. From 20 to 60 seconds, the dome pressure maintains at about 920 psia until the MSIV starts to close, then the pressure begins to increase. The dome pressure will rise and then level off at about 1100 psia when the S/R valve opens and then eventually it will drop off with faster depressurization at about 200 seconds after the break.
So the net effect of the MSIV closure level change is a delay of at most 40 seconds for the dome pressure to rise and then level off.
3.2.2 Core Inventory As shown in figure 3-4, the change of the MSIV closure level has no
,significant effect upon the water level insid'e the core shroud.
Up to 80 seconds after the break there is a slight difference due to the earlier relief of dome pressure for MSIV closure at L2, but the water level is still dominated by the blowdown which is only slightly affected by the change in MSIV closure level.
3.2.3 Suppression Pool feoperature, Since the heat exchanger starts at 600 seconds after the suppression pool reaches 110*F, the slow blowdown from the reactor vessel, compared with that of_the DBA case, allows the sensible and decay heat from the core to be removed for a much longer time, and thus, the maximum suppression pool temperature remains below the value of 200*F. The change of MSIV closure level also has insignificant effect upon the maximum suppression pool temperature.
3.2.4 Fission Product Release Because the change of MSIV closure level has no significant effect upon the water level inside the shroud and the blowdown rate, there 3-3
QUAD-1-83-008 is no release of fission products during LOCA. With the break size of 0.09 square feet, the peak cladding temperature of 1736*F occurs at about 450 seconds (FSAR figure 6.3-58 and table 6.3-8).
After the break with the ECCS in operation, even if the ECCS fails, the MSIV has been closed long enough to preclude fuel clad failure.
Assuming the most conservative case, that the outflow from the MSIV equals the steady state flowrate of 4130 lbm/sec., then the amount of steam out of the MSIV due to 40 seconds delay = 40 x 4130 = 1.65x105 lbm.
From the design-basis analysis for the steam line break (FSAR 15-6.4.5), with a total discharge of 100,000 lbm of coolant, and assuming all activity becomes airborne, the activity released to the environmen't would be:
o Total halogens:
6.95 x 101 curies.
o Total noble pses:
7.42 curies.
Thus, assuming direct proportion with the coolant amount, the addi-tional total activity due to the delay of 40 seconds would be:
o Total halogens:
1.14 x 102 curies, o
Total noble gases:
1.22 x 101 curies.
By the same way, the radiological effects are calculated and tabulated in the table 3-1, where it is shown that the extra activity due to the 40 seconds delay in MSIV closure is actually a small amount.
It represents only about 15 percent increase in dose for NRC DBA or 0.25 percent increase of the 10 CFR 100 limits for the most adverse case.
3.3 Feedwater Line Break (Outside Containment)
The steam quality of the coolant from the break and the extra activity was estimated based on the feedwater line break analysis of FSAR, sec-3-4
~
QUAD-1-83-008 tion 15-6.6.
The assumptions used for the analysis are as follows:
o The added radioactive coolant spilled outside the conta'inment is the product of the differential time in MSIV closure level change and the feedwater blowdown rate.
o Both the condenser and turbine are large energy sinks, with negligible variations in thermal-hydraulic conditions and so the steam flow is dependent upon the upstream conditions and thus the flow into the turbine can be simulated by a negative fill junction which is moni-tored by a control system.
3.3.1 Vessel Pressure One of the feedwater lines outside the containment was assumed to break instantaneously and circumferential1y. The feedwater line check valves isolate the reactor from the break (figure 3-5).
Recirculation pumps are tripped due to loss of feedwater.
Steam continues to flow out of the MSIV until it closes either by L2 or
~
L1 water level trip.
Figure 3-5 shows the vessel dome pressure for the two cases of MSIV closure.
For the MSIV closure at L2, the closure time starts at 13.8 seconds after the break; for the MSIV closure at L1, the closure time starts aoout 3 seconds later.
It is shown from the plots that the dome pressure increases correspondingly, and eventually levels off when the S/R valves open and it then decreases when the HPCS comes on.
3.3.2 Core Inventory Figure 3-6 shows the water level inside the shroud versus time after the break, and it is observed that there is again no significant effect upon the water level by the change in MSIV closure level.
3-5
QUAD-1-83-008 3.3.3 Suppression Pool Temperature With the feedwater line breaks outside the containment, the sup-pression pool temperature is of no concern.
3.3.4 Fission Product Release Because of 3 seconds time differential between the closure of MSIV at water level L2 compared with that at level L1, the radiological effects from the feedwater line break outside the containment, due to the added radioactive coolant spilled out from the break, must be accounted for.
It is estimated as follows.
a.
Assuming a conservative steady state blowdown flow rate of 4130 lbm/sec, for 3 seconds, with one line break, the amount of coolant spilled out:
4130
=3x
= 6195 lbm.
b.
From FSAR analysis of feedwater line break (section 15-6.6),
with MSIV closure at water level L2, the total integrated mass of coolant leaving the break is 788,000 lbm, of which 165,000 lbm flashes to steam.
The steam has an assumed iodine carryover of 100 percent while the unflashed liquid has one percent airborne iodine.
c.
With.the MSIV closure at water level L1, and assuming the same percentage of coolant flashes to steam as that of the FSAR analysis, the amount flashed to steam 16 '
= 6195 x
= 1297 lbm.
8 The amount of unflashed liquid = 4898 lbm.
3-6
QUAD-1-83-008 d.
Because the pathway to the environment is due to particulate filtration, through the turbine building ventilation system into the plant vent stack, the airborne iodine is of main concern (FSAR 15-6.6).
Assuming that steam has 100 percent iodine carryover, while liquid has one percent carryover, the unflashed fluid is connected to " equivalent steam mass."
.. Amount of extra flashed equivalent steam due to MSIV closure at L1
=1297+4898xh=1346lbm.
The amount of flashed equivalent steam from FSAR feedwater line break analysis
=165,000+623,000xh=171,230lbm.
Therefore, the increased percentage of radioactivity to the environment due to the MSIV closure level' change is:
x 10 3 = 0.786%
=1 0
Table 3-2 shows the radiological effects due to feedwater line break. Again it shows that with MSIV closure at level L1, the additional activity introduced constitutes only ~0.8 percent of the_FSAR analysis value.
This concludes that the MSIV level change has only minor impact on the radiological effect in the event of feedwater line break.
i I
i 3-7 i
QUAD-1-83-008 TABLE 3-1 Radiological Effects Due to Small Break NRC DBA Extra Activity from 10 CFR 100 Large Break MSIV Closure at L1 (2 HRS.)
(2 HRS.)
(2 HRS.)
5.8x10j Whole Body Dose (rem)-
25 0.36 Thyroid Dose (rem) 300 6.06 6.0 x 10 TABLE 3-2 Radiological Effects Due to Feedwater Line Break Conservative Extra FSAR Activity 10 CFR 100 Analysis-With (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />)
MSIV at L2 MSIV at L1 Whole Body Dose (rem) 25 8.1 x 10 9 6.342x10h'
~
~
Thyroid Dose (rem) 300 7.9 x 10 6.186 x 10 (509 meters) i 3-8
l2 I
I I
I A MSIV CLOSURE AT L1 O MSIV CLOSURE AT L2 1.0 CQ 8 Og OA s a 06c _3 C
b n
0.6 O
L L
g G
U O
m O
E 0.6
~
o JU w$
0.4 BC a
ha 0.2 MSIV STARTS TO MSIV STARTS TO CLOSE CLOSE AT L2 AT L1 1
0.0 O
5 10 15 20 25 TIME AFTER BREAK, SEC.
FIGURE 3-1 NORMALIZED VESSEL DOME PRESSURE VS. TIME AFTER BREAK DBA LARGE RECIRCULATION SUCTION BREAK 3-9
60 l
l
[
[
A MSIV CLOSURE AT L1 O MSIV CLOSURE AT L2 50 A
b@
O O
8 bO 6
C C
g g
40 g
' 30 i
W"i 5
g 20 BAF 10 -
MSIV STARTS MSIV STARTS TO CLOSE TO CLOSE AT L2 AT L1 l
I l
0 l
0 5
10 15 20 25 l
TIME AFTER BREAK, SEC.
FIGURE 3-2 WATER LEVEL INSIDE SHROUD l
VS. TIME AFTER BREAK DBA LARGE RECIRCULATION SUCTION BREAK l
3-10 l
L
1.2 A MSIV CLOSURE AT L1 O MSIV CLOSURE AT L2 0
- 1.0gp O
A a
O C
C C
A A
A A
0.8 U
i 0.6 E
0.4 w
e M
3
,0.2 z
MSIV STARTS TO MSIV STARTS TO CLOSE CLOSE AT L2 AT L1 I
I I
0.0 O
20 40 60 80 100 TIME AFTER BREAK, SEC.
FIGURE 3-3 NORMALIZED VESSEL DOME PRESSURE VS.
TIME AFTER BREAK SMALL RECIRCULATION SUCTION BREAK, BREAK AREA = 0.09 SQ. FT.
3-11
l 60 l
l l
l A MSIV CLOSURE AT L1 O MSIV CLOSURE AT L2 50 8
3 0
0 C
C C
40 Oa C A
O TAF C 30 2
5 y 20 BAF 10 MSIV STARTS MSIV STARTS TO CLOSE TO CLOSE AT L2 AT L1 I
I I'
I c
O 20 40 60 80 100 TIME AFTER BREAK, SEC.
FIGURE 3-4 WATER LEVEL INSIDE SHROUD VS.
TIME AFTER BREAK.
SMALL RECIRCULATION SUCTION BREAK BREAK AREA = 0.09 SQ. FT.
3-12
1.2 l
l l
l A MSIV CLOSURE AT L1 O MSIV CLOSURE AT L2
+
h 1.04H e o
e o
2 i
0.8 w
a' O.6 a
C 0.4 en
'3 28*
0.2 MSIV STARTS TO MSIV STARTS TO CLOSE AT L2 l LOSE AT L1 O
5 10 15 20 25 TIME AFTER BREAX, SEC.
FIGURE 3-5 NORMALIZED VESSEL DOME PRESSURE VS.
TIME AFTER BREAX FEEDWATER LINE BREAX OUTSIDE CONTAINMENT 3-13
60 j
l i
A MSIV CLOSURE AT L1 O
MSIV CLOSURE AT L2 50 6
e 0
6 o
o o
o e
40 b
.1 TAF y
30 a
5 t2 2
20 BF 10 MSIV STARTS TO MSIV STARTS TO CLOSE AT L2 CLOSE AT L1 0
l l
l l
0 5
10 15 20 25 TIME AFTER BREAX, SEC.
FIGURE 3-6 WATER LEVEL INSIDE THE SHROUD VS.
TIME AFTER BREAX FEEDWATER LINE BREAX OUTSIDE CONTAINMENT 3-14
- i m
~
QUAD-1-83-008
4.0 REFERENCES
1.
"RETRAN A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems," EPRI NP-1850, May 1981.
2.
LaSalle County Station Final Safety Analysis Report, chapters 6 and 15, May 1981.
3.
LSCS - Mark II DAR, revision 8, June 1980.
4.
Analysis Report, Alternate Rod Insertion System for LaSalle County Station, unit 2, QUAD-1-83-007, revision 0, August 1983.
9 4-1
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