LD-83-066, Forwards Reanalysis of Steam Generator Tube Rupture Event W/Loss of Offsite Power & Addl Single Failure,Per NRC 830426 Request.Info Completes SER Confirmatory Item 18
| ML20077D871 | |
| Person / Time | |
|---|---|
| Site: | 05000470 |
| Issue date: | 07/22/1983 |
| From: | Scherer A ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| LD-83-066, LD-83-66, NUDOCS 8307270352 | |
| Download: ML20077D871 (44) | |
Text
_ _ _ _ _ _ _ _ _ _ _ _
C-E Power Sy;t:ms Tel 203/688-1911 Combustion Engineering Inc Telex 99297 1000 Prospect Hill Road Windsor. Connecticut 06095 POWER H SYSTEMS Docket No.: STN 50-470-F July 22,1983 LD-83-066 Mr. Darrell G. Eisenhut, Director Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Subject:
Confirmatory Item 18, Steam Generator Tube Rupture Analysis
References:
(A) Letter, C. O. Thomas to A. E. Scherer, dated April 26, 1983 (B) Letter LD-83-043, A. E. Scherer to D. G. Eisenhut, dated May 10, 1983
Dear Mr. Eisenhut:
Reference (A) documnted an NRC Staff request for a re-analysis of the steam generator tJbe rupture event With a loss of offsite power and an additional single failure. C-E provided a response schedule and a brief summary of the analytical assumptions in Reference (B). Enclosed are twenty-five (25) copies of our re-analysis. The analysis is presented as Appendix 15D and will be incorporated into the text of CESSAR-F in an amendment. The re-analysis provided addresses the concerns noted in Reference (A) and shows calculated doses within the guidelines of 10 CFR 100. We anticipate, therefore, that this submittal should close out Confirmatory Item 18 of the CESSAR-F SER.
If you have any questioin on the submittal, please contact me or Mr. G. A.
Davis of my staff at (203) 688-1911, extension 5207.
Very truly yours, COMBUSTION ENG NEERING, INC.
ans A. E. h&er Director Nuclear Licensing AES:las 9? r cc: Gary Meyer (Project Manager / USNRC) )
8307270352 830722 PDR ADOCK 05000470 E pyg
7 g APPENDIX 150 STEAM GENERATOR TLBE RUP"URE WITH A LOSS OF 0FFSITE F0WER AND S ' NG'_E FAILURE 1 50.1 IDENTIFICATION OF EVENT AND CAUSES The steam generator tube rupture (SGTR) accident is a penetration of the barrier between the reactor coolant system (RCS) and the main steam system which results from the failure of a steam generator U-tube. Integrity of the barrier between the RCS and main steam system is significant from a radiological release standpoint. The reactivity from the leaking steam generator tube mixes with the shell-side water in the affected steam generator. Subsequent to reactor trip and turbine trip, the radioactive fluid is released through the steam generator safety or atmospheric dump valves as a result of the postulated loss of normal AC power.
A SGTR event results in a depressurization of the RCS. Prior to reactor trip, the radioactivity is transported through the turbine to the condenser where the noncondensable radioactive materials would be released via the condenser air ejectors. As a result of the reactor trip the turbine / generator trips and normal ac power may be lost. The electrical power would then be unavailable for the station auxiliaries such as the reactor coolant pumps, and the main feedwater pumps. Under such circumstances the plant would experience a loss of load, normal feedwater flow, forced reactor coolant flow, condenser vacuum, and steam generator blowdown. The loss of offsite power subsequent to the time of reactor trip and turbine / generator trip is assumed in the analysis, since it produces the most adverse effect on the radiological releases. The plant is brought to shutdown cooling entry conditions by the operator, as described in Reference 1, through the use of the steam generator atmospheric dump valves (ADVs), pressurizer backup heaters, auxiliary spray, safety injection system, and auxiliary feedwater system.
In addition to the above scenario, the most limiting single failure with respect to radiological releases is assumed to occur. The systems used to mitigate the consequences of this event are; the safety injection system (SIS), pressurizer pressure control system (PPCS), pressurizer level control system (PLCS), auxiliary feedwater system (AFWS), and the ADVs. The single failures which may impact the radiological consequences of the SGTR event are; failure of an ADV to close in the affected steam generator after the operator initially opens it, and failure of a diesel generator to start following the loss of offsite power.
The failure of an ADV to close in the affected steam generator will result in additional steam release until the operator is able to isolate the ADV by closing the associated block valve. The failure of the diesel generator to start will leave the fcilowing components inoperable; one HPSI pump, one charging pump, one auxiliary feedwater pump, and one half of the pressurizer backup heaters. The partial loss of the heat removal capabilities of the safety injection flow and auxiliary feedwater flow may require the operator to steam from the affected generator in order to maintain the RCS in a subcooled state. This steaming would be in addition to that which may be required to
. . i s prevent overfilling of the affected steam generator. The excess steaming due to the failure of the ADV to close is larger than that resulting from the ,
failure of a diesel generator. Therefore, the failure of an ADV to close in {
the affected steam generator is the most limiting single failure with respect to radiological releases.
Diagnosis of the SGTR accident is facilitated by radiation monitors which j initiate alanns and infonn the operator of abnormal activity levels and that 1 l
corrective operator action is required. These radiation monitors are located in the condenser air ejector exhaust, steam generator blowdown lines, and turbine and auxiliary building ventilation ducts and stack. Additional diag-nostic information is provided by RCS pressure and pressurizer level response j indicating a leak and by level resoonse in the affected steam generator.
1 1 50. 2 SEOUENCE OF EVENTS AND SYSTEMS OPERATION
! Table 15D-1 presents a chronological list of events which occur during the steam generator tube rupture event with a loss of offsite power and stuck open ADV, from the time of the double-ended rupture of a steam generator U-tube to the attainment of shutdown cooling entry conditions. The sequence presented demonstrates that the operator can cool the plant down to shutdown cooling entry conditions during the event. All actions required to stabilize the plant and perform the required repairs are not described here.
Table 150-2 contains a matrix which describes the extent to which normally operating plant systems are assumed to function during the course of the event.
Table 150-3 contains a matrix that summarizes the utilization of safety systems ,
as they appear in the transient analyses.
The operator actions assumed in this analysis are consistent with the C-E Emergency Procedure Guidelines (EPGs) documented in Reference 1. The major operator actions assumed in the analysis are summarized below and listed in Figure 15D-16.
- 1) The operator opens one ADV in each steam generator in order to cool the RCS to 550 F (temperature of hottest hot leg) at a cooldown rate of 100* F/ hr. The hot leg temperature value is derived frc= the EPO bracketed value of 565*F. The initial cooldown of the RCS is aimed at preventing reopening of the MSSVs on the affected steam generator by cooling down the RCS to 10 F below the saturation temperature corresponding to the
' MSSV opening pressure setpoint. An additional 5'F is employed to account for instrument uncertainties. The technical specification cooldown rate of 100 F/hr is used in lieu of a plant procedure specific cooldown rate.
This rapid cooldown rate requires a larger valve opening area and results in more steam flow out the stuck open ADV. The 100 F/hr cooldown rate translates into a 12'.' opening for one ADV of each steam generator.
- 2) The time delay for the first operator action identified above is consistent
' with the guidelines of Reference 2 which recomends a 5 minute operator action delay time for a steam generator tube rupture, and an additional two minutes for conipletion of a discrete operator manipulation, Thus, a 7 minute operator action time is assumed for completion of the first action, namely, opening of the ADVs to cool down the RCS to 550 F.
, 3) The operator attempts to isolate the affected steam generator when the RCS temperature is below 550*F. At this time it is assumed that the ADV in the affected steam generator sticks open. The operator will be alerted that the ADV has not closed by the following signals:
(a) continued alarming from the stack radiation monitors.
(b) continued indication of steam flow through the flow measuring venturiis on the steam generator, and (c) decreasing steam generator level in spite of the attempted isolation which should have caused the level to increase.
- 4) The operator closes the block valve associated with the stuck open ADV on the affected generator 30 minutes after the attempted isolation. This delay is consistent with the criteria for operator actions outside of the control room as stated in Reference 2. The time delay includes the time it takes to get to the location of the block valve, and the time required to completely close the v31ve.
- 5) The operator initiates auxiliary spray flow in order to regain pressurizer level two minutes after the block valve is closed. The timing of this action is consistent with the guidelines of Reference 2. The operator will use the HPSI system, prescurizer backup heaters, and auxiliary sprays to control RCS inventory and subcooling.
- 6) The operator continues to cool down the RCS using the unaffected generator at 20 F/hr. This reduction in cooldown rate from 100*F/hr to 20 F/hr maximizes the radiological release during the long term cooldown by; a) delaying entry into shutdown cooling until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after event initiation, thereby, maximizing the primary heat to be renoved through the ADVs within the 0-8 hour time period.
b) maximizing the primary to secondary leak, thereby, increasing the operators use of the operable ADV in the affected steam generator to prevent its overfilling, and I c) maintaining the primary to secondary leakage at a higher enthalpy, thereby, maximizing the flashing fraction of the leakage at the secondary side.
- 7) The operator maintains approximately a 20*F subcooling margin as per Reference 1.
- 8) The operator will use the unisolated ADV on the affected SG in order to prevent its overfilling due to the primary-to-secondary leak.
The success paths followed to mitigate the consequences of this evenc are as follows:
o \
Raactivity Control:
A pressurizer pressure decrease could result in the generation of a number of Core Protection Calculator (CPC) trips such as RCS saturation trip, low DNBR trip, or low RCS pressure boundary trip. In this analysis the pressurizer pressure decrease results in the generation of a CPC RCS saturation trip.
Subsequently the CEAs drop into the core. The RCS pressure starts to decrease more rapidly and a Safety Injection Actuation Signal (SIAS) is generated on a low pressurizer pressure signal. As a result, additional negative reactivity will be added to the system, in the form of borated water from the refueling water tank (RWT). Once the plant parameters have been stabilized, the operator adjusts the boron concentration to insure that a proper negative reactivity shutdown margin is achieved prior to cooldown. The baron concentration is adjusted by manually throttling the HPSI discharge valves to replace RCS volume shrinkage.
Reactor Heat Removal:
During the pre-trip part of the transient, reactor heat removal is accomplished in the nont.al manner. After turbine trip and consequential loss of offsite power, the reactor heat is removed by natural circulation. Additional cooling is available through the injection of relatively low enthalpy RWT water.
Primary System Integrity:
After initiating cooldown procedures, the coerator must reestablish the pressur-izer water level. During the cooldown phase, the pressurizer heaters, auxiliary spray, and HPSI pumps are used by the operator to control the RCS pressure and level.
When the RCS pressure has been reduced to approximately 650 psia, the operator will vent or drain the SITS to reduce their pressure and will then isolate them.
Secondary System Integrity:
Following the loss of offsite power the main feedwater flow is terminated and the Steam Bypass Control System (SBCS) is left inoperable. The Main Steam Safety Valves (MSSVs) open when secondary pressure increases and provide a path for removal of generated and/or stored core heat. As the secondary water level decreases the Auxiliary Feedwater System (AFWS) is actuated and restores the level .
The operator will take control of the plant and open one ADV in each steam generator in order to cool the RCS at the Technical Specification limit of 100 F/hr. The technical specification cooldown rate of 100 F/hr is used in lieu of a plant procedure specific cooldown rate. This rapid cooldown rate requires a larger valve opening area in comparison to that for less rapid cooldown rates and results in more steam flow through the stuck open ADV.
Once the indicated RCS hot leg temperature is below 550 F the operator will attempt to isolate the affected steam generator. The operator will use the l secondary steam activity alarms in order to identify the affected steam genera tor. 1
, Due to the assumed failure of one ADV to reclose, the operator will need to close the ADV block valve to isolate the affected steam generator. The decreas-ing SG level in the affected steam generator, continued indication of steam flow from the affected steam generator, and continued alanning from the stack radiation monitors will alert the coerator to the stuck open ADV. Once the affected steam generator is isolated the operator will steam from the intact steam generator in order to bring the RCS to shutdown cooling entry conditions.
The operator will steam from the affected SG in order to prevent overfilling.
Radioactive Effluent Control:
A Containment Isolation Actuation Signal (CIAS) is generated subsequent to the SIAS. CIAS isolates various systems to reduce or terminate radioactive releases.
CIAS actuates primary and containment isolation equipment. Other actions may be initiated by B0P systems. See Applicant's FSAR for details.
1 50.3 ANALYSIS OF EFFECTS AND CONSEOUENCES 150.3.1 CORE AND SYSTEM PERFORMANCE A. Mathematical Model The thermal hydraulic response of the Nuclear Steam Supply System (NSSS) to the steam generator tube rupture with a loss of offsite power and stuck open ADV was simulated using the CESEC-III computer program up to the time the operator takes contr:1 of the plant and a CESEC-III based cooldown algorithm thereafter. The CESEC-III computer program is described in Reference 3. The thennal margin on DNBR in the reactor core was evaluated using the TORC computer program (Reference 4) as described in Section 15.0.3 with the C l critical heat flux correlation described in CENPD-162 (Reference 5).
B. Input Parameters and Initial Conditions The initial conditions and input parameters employed in the analyses of the system response to a steam generator tube rupture with a concurrent loss of offsite power and stuck open ADV are listed in Table 150-4.
Additional discussion on the input parameters and the initial conditions are provided in Section 15.0. Conditions were chosen to maximize the radiological releases.
The initial reactor operating conditions were varied over the operating space given in Table 15.0-5 to determine the set of conditions which would produce the most adverse consequences following a steam generator tube rupture with a loss of offsite power and stuck open ADV. Various combinations of initial operating conditions were concidered in order to determine the reactor trip time which would result in the most adverse radiological releases. The parametric studies indicated that the maximum offsite mass release is obtained when the transient is initiated with the minimum allowed RCS pressure, minimum initial pressurizer liquid volume, maximum initial steam generator liquid volume. maximum core power, minimum
I O
a core coolant flow, and maximum core coolant inlet temperature. This combination of initial conditions results in an early generation of a reactor trip signal due to exceeding the CPC hot leg saturation tempera-ture range limit.
C. Results The dynamic behavior of important NSSS parameters following a steam generator tube rupture is presented in Figures 150-1 to 150-15.
For a double-ended rupture, the primary to secondary leak rate exceeds the capacity of the charging pumps. As a result, the pressurizer pressure gradually decreases from an initial value of 2100 psia. The primary to secondary leak rate and drop in pressurizer water level causes the second and third CVCS charging pumps to turn on. Even with all three CVCS charging pumps on line the pressurizer pressure and level continue to drop. At 47 seconds a reactor trip signal is generated due to exceeding the CPC hot leg saturation temperature range limit. The pressurizer empties at approximately 546 seconds (Figure 15D-5). At 570 seconds a safety injection actuation signal is generated, and by 620 seconds the safety injection flow is initiated. After the pressurizer empties, the reactor vessel upper head begins to behave like a pressurizer, and controls the reactor coolant system pressure until the pressurizer begins to refill at approximately 4020 seconds. Due to flashing caused by the depressurization, and the boil off due to the metal structure to coolant heat transfer, the reactor vessel upper head begins to void at about 77 seconds (Figure 150-6). Consequently, the RCS pressure (Figure 150-2) begins to decrease at a lower rate at this time.
Following reactor trip and with turbine bypass unavailable, the main steam system pressure increases until the MSSVs open at 52 seconds to control the main steam system pressure. A maximum main steam system pressure of 1330 psia occurs at 56 seconds. Subsequent to this peak in the pressure, the main steam system pressure decreases, resulting in the closure of the main steam safety valves at 95 seconds. The MSSVs cycle twice more in this manner until the operator takes control of the plant.
Prior to reactor trip, the main feedwater control system is assumed to be in the automatic mode and supplies feedwater to the steam generators such that steam generator water levels are maintained. Following reactor trip, the main feedwater flow is terminated due to the loss of offsite power. As the level in the steam generators decrease an Emergucy Feed-water Actuation Signal (EFAS) is generated resulting in auxiliary fesdwater ficw which acts to restore the SG level.
At 460 seconds the operator takes control of the plant and opens one ADV on each SG to cool down the plant. This is consistent with the EPGs. At l 2100 seconds the RCS has been cooled to 550*F. The operator isolates the auxiliary feedwater to the affected generator, closes the main steam isolation valves of both steam generators, and attempts to close the ADV of the affected generator. The operator recognizes that the ADV did not close and has the appropriate block valve closed within 30 minuto . The operator then initiates an orderly cooldown by means of the atmospheric
dump valves and the auxiliary feedwater flow to the unaffected steam generator. Thereafter, the operator will steam the affacted steam generator in order to prevent overfilling due to the leak flow. After the pressure and temperature are reduced to 400 psia and 350*F, respec-tively, the operator activates the shutdown cooling system and isolates the unaffected steam generator.
The maximum RCS and secondary pressures do not exceed 110% of design pressure following a steam generator tube rupture event with a loss of offsite power and stuck open ADV, thus, assuring the integrity of the RCS and the main steam system.
Figure 150-14 gives the main steam safety valve integrated flow rates versus time for the steam generator tube rupture event with a loss of offsite power and a stuck open ADV. At 460 seconds, when operator action is assumed, no more than 41500 lbm of steam from the damaged steam generator and 41470 lbm from the intact steam generator are discharged via the main steam safety valves. Also, during the same time period approximately 17560 lbm of primary system mass is leaked to the damaged steam generator.
Subsequently, the operator begins a plant cooldown at the technical specification cooldown rate (100 c/hr) using both steam generators, the atmospheric dump valves, and the emergency feedwater system. Once the affected steam generator is isolated, it is assumed that the operator reduces the cool down rate to 20 F/hr. For the first two hours following the initiation of the event, 484000 lbms of steam are released to the environment through the atmospheric dumo valves. For the two to eight hour cooldown period an additional 973000 lbms of steam are released via the atmospheric dump valves.
15D.3.2 RADIOLOGICAL CONSEQUENCES A. Physical Model The evaluation of the radiological consequences of a postulated steam generator tube rupture assumes a complete severance of a single steam generator tube while the reactor is operating at full rated power, a loss of offsite power three seconds after turbine / generator trip and a stuck open ADV. Occurrence of the accident leads to an increase in contamination of the secondary system due to reactor coolant leakage through the tube break. A reactor trip occurs automatically as a result of approaching saturation conditions in the hot leg at approximately 47 seconds after the event initiation. The reactor trip automatically trips the turbine /
genera tor.
The steam gen!rator pressure will increase rapidly, resulting in steam discharge as well as activity release through the main steam safety valves. Venting frcm the affected ste&m generator, i.e., the steam generator which experiences the tube rupture, continues until the secondary l system pressure is below the main steam safety valve setpoint. After 460 l seconds, the operator initiates a plant cooldown at the technical specifi- '
- cation cooldown rate (100 F/hr) using the steam generators, atmospheric l dump valves, and the emergency feedwater system. The technical specification J
i ,
cooldown rate of 100*F/hr is used in lieu of a plant procedure specific cooldown rate. This rapid cooldown rate requires a larger valve opening area and results in more steam flow through the stuck open ADV. Upon
- isolation of the affected generator the cooldown continues at 20*F/hr using the unaffected generator. The operator may steam the affected steam generator to prevent its overfilling.
i The analysis of the radiological consequences of a steam generator tube rupture considers the most severe release of secondary activity as well as primary system activity leaked from the tube break. The inventory of f iodine and noble gas fission product activity available for release to i the environment is a function of the primary-to-secondary coolant leakage rate, the assumed increase in fission product concentration, and the mass of steam discharged to the environment. Conservative assumptions are made for all these parameters.
B. Assumptions and Conditions The assumptions and parameters employed for the evaluation of radiological releases are: ,
- 1. Accident doses are calculated for two different assumptions: (a)an event generated iodine spike (GIS) coincident with the initiation of the event and (b) a pre-accident iodine spike (PIS).
! 2. Technical specification limits are employed in the dose calculations i i for the primary system (1.0 uCi/gm) and secondary system (0.1 uCi/gm) activity concentrations.
- 3. Following the accident, no additional steam and radioactivity are released to the environment when the shutdown cooling system is placed in operation.
- 4. A spiking factor of 500 is employed for the GIS.
- 5. For the PIS condition, the technical specification limit (60 uCi/gm) for the primary system activity concentration is employed.
- 6. Technical specification limit (1 gpm) for the tube leakage in the unaffected steam generator is assumed for the duration of the transient.
- 7. The tube leakage which flashes to steam is assumed to be released to the atmosphere with a decontamination factor (DF) of 1.0.
- 8. A 0F of 100 is assumed between the steam generator water and steam phases.
- 9. The 0-2 hour and 2-8 hour primary-to-secondary leakage through the rupture is 272000 lbm and 376000 lbm, respectively.
10.
The x 10-ajmosphejic sec/m dis;;ersion factors employed in the analyses are:for the exc for the low population zone.
<-e, n m - - - - - - , . - , - ~ - - , - --.2-n--~ -, --v---,,-+.w,-.c ---r -w rww-- -
e .-. w+ e- - < r - , , - - ,
- 11. Dilution of primary and secondary systems due to HPSI flow and
, auxiliary feedwater flow is accounted for in the dose calculation.
C. Mathematical Model The mathenatical model enployed in the evaluation of the radiological consequences during the course of the transient is described in Section 15.0.4.
D. Results The two-hour exclusion area boundary (EAB) and the eight-hour low popula-tion zone (LPZ) boundary inhalation doses for both the GIS and the PIS are presented in Table 150-5. The calculated EAB and LPZ doses are well within the acceptance criteria.
1
50.4 CONCLUSION
S The radiological releases calculated for the SGTR event with a loss of offsite power and a stuck open ADV are well within the 10CFR100 guidelines. The RCS and secondary system pressures are well below the 110% of the design pressure limits, thus, assuring the integrity of these systems. Additionally, no violation of the fuel thermal limits occurs, since the minimum DNBR remains above the 1.19 value throughout the duration of the event.
Voids form in the reactor vessel upper head region during the transient, due to the thermal hydraulic decoupling of this region frca the rest of the RCS.
The upper head region liquid level renains above the top of the hot leg throughout the transient. Natural circulation cooldown is not impaired during the transient.
l 0
1 50. 5 REFERENCES
- 1) " Combustion Engineering Emergency Procedure Guidelines", CEN-152,
- Revision 01, November,1982.
- 2) " Time Response Design Criteria for Safety-Related Operator Actions",
American National Standard, ANSI N660, Draft.
- 3) LD-82-001 (dated 1/6/82), "CESEC Digital Simulation of a Combustion Engineering Nuclear Steam Supply System", Enclosure 1-P to letter from A. E. Scherer to D. G. Eisenhut, December, 1981.
- 4) " TORC Code - A Computer Code for Determining the Thermal Margin of a
, Reactor Core", CENPD-161-P, July,1975, Proprietary Information.
- 5) "C-E Critical Heat Flux - Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Space Grids", CENPD-162-P, April, 1975, Proprietary Information.
1 P
i i
I 8
1
TABLE 150-1 SEQUENCE OF EVENTS FOR A STEAM GENERATOR TUBE l
RUPTURE WITH A LOSS OF OFFSITE POWER i AND STUCK OPEN ADV '
Time Setpoint (Sec) Event or Value Success Path 0.0 Tube Rupture Occurs ---
40 Third Charging Pump Started, feet -0.75 Primary System Integrity below program level 40 Letdown Control Valve Throttled -0.75 Primary System Integrity Back to Minimum Flow, feet below program level 47 CPC Hot Leg Saturation Trip Signal ---
Reactivity Control 48 Turbine / Generator Trip: Stop ---
Secondary System Integrity Valves Start to Close CEAs Begin to Drop ---
Reactivity Control 51 Turbine stop Valves Closed ---
Secondary System Integrity Loss of Offsite Power ---
52 LH Main Steam Safety Valves open, 1265 Secondary System Integrity psia 52 RH Main Steam Safety Valves open, 1265 Secondary System Integrity psia 56 Maximum Steam Generator Pressures 1330 Both Steam Generator, psia 95 Main Steam Safety Valves Closed, 1218 Secondary System Integrity psic 167 Auxiliary Feedwater Actuation on 19.76 Secondary System Integrity Low Steam Generator Level Trip Signal, Intact Steam Generator, feet above tube sheet 177 Auxiliary Feedwater Actuation on 19.76 Secondary System Integrity Low Steam Generator Level Trip Signal, Ruptured Steam Generator, feet above tube sheet
l l
l TABLE 15D-1 (Cont'd. )
1 Time Setpoint (Sec) Event or Value Success Path 460 Operator Initiates Plant Cooldown ---
Reactor Heat Removal by Opening One ADV on each SG 546 Pressurizer Empties ---
570 Safety Injection Actuation Signal, 1578 Reactivity Control psia 620 Safety Injection Flow Initiated --- Reactivity Control 2100 Operator Attempt to Isolate the 550 Secondary Systen Integrity Damaged Generator, RCS Tem. , F 3900 Operator Closes the ADV Block ---
Secondary System Integrity Valve 4020 Operator Initiates Auxiliary Primary System Inventory Spray Flow 4500 Operator Controls Auxiliary 20 Primary System Integrity Spray Flow, Backup Pressurizer Heater Output, and .9 PSI Flow to Reduce RCS Pressure and Control Subcooling, *F 28,800 Shutdown Cooling Entry Conditions 400/350 Reactor Heat Removal Reached, RCS Pressure, psia /
Teroperature, OF
TABLE 15 0-2' (Sheat 1 of 2)
DISPOSITION OF NORMALLY OPERATING SYSTEMS FOR THE STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF 0FFSITE POWER AND STUCK OPEN ADV
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- 4. Steam Eypa:s Control Systr. /
- 5. Pressurizer Pressure Control System /
- 6. Pressuri:er Level Control System v 6
- 7. Control Element Drive Mechanism Con rel System v {
- 8. Reactor Regulating System /
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- 11. Chemical and Volume control System v i v
- 12. Secondary Chemistry Centrol System
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- 14. Turbine Gland Sealing System * / f;
- 15. Nuclear Cocling '.latar System * / !
- 16. Turbine Cooling Water Systcm* v' !
- 17. Plant Cooling.'la:er System * / l
- 18. Condensate Storage Facilities + / l
- 19. Circulating Water System * /
Spent Fuel Fcci Cocling anc Clean-Up System * / !
- 20. ,
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- 1. Reac:or Protec:icn Sys:em /
- 2. DNSR/t.P0 Calculator
- 3. Engineered Safety Features Actuatien Systics /
- 4. Supplementary ?rotectica System I
Reactor Trip Ssitch Gear '/ '
5.
- 6. Main Steam Safety Valves' /
- 7. Primary Safety Valves
- 8. Main Steam Isclation System * / 1
- 9. Emergency Feedwater System * /
- 10. Safety Injecticn System /
- 11. Shutdown Cooling System /
- 12. Atmospheric Cucp Valve System * /
- 13. Containment Isolation System *
- 14. Containment spray System *
- 15. Iodine Removal System *
- 16. Containment Ccebustible Sas Centrol Systcm* ,
- 17. Diesel Generaters and Su::or: Systems' ; v/ -
i
- 18. Component (Osantial) C:clinS Water System
- i /
- 19. Statien Service Water System * / 3 i
NOTES:
- 1. Tha 0; erat:r ar.ually iselites the affe::2d steam generator. fl
- t i
- 0alance-of-Plant Systcms - i l
=wa %
TABLE 150-4 ASSUMPTIONS AND INITIAL CONDITIONS FOR THE STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFFSITE POWER ~
AND STUCK OPEN ADV .
Assumed Parameter Value .,
, . . x Core Power Level, MWt 3876 . -
[
Core Inlet Coolant Temperature, *F
~
57 0 s ], _! J Reactor Coolant System Pressure, psia 2100 j -
Core Mass Flow Rate,106 lbm/hr - 155 ; .
s
,, ; ,e One Pin Integrated Radial Peaking Factor,i ,
1;.-
i with Uncertainty l.53 ,.
~
Steam Generr. tor Pressure, psia , 1126 7 ,.s,1
~~ M.i.
Moderator Temperature Coefficient,10 A ao/*F '
-3.5' -
t ..
' ~ s Doppler Coefficient Multiplier s 1/15 N,. 5 s s
^
~ , %%
CEA Worth at Trip, f. ao (most reactive CEA fully ,. ,? <
~
withdrawn) <
(10.0 s *
- s. .,
.. r n ,
s y /
' - + g e o w
- W R \' %
4 ,
- ' 4 q*-
t e %.
w- , ,.
N h .
4 3 3 q % ,!
% k
- ,, ,\
- ~~ .,
b
+
1 g 'i
? g
.h k a
% _s~, .;
w:
.
- _m_____________ _ _ . _
.. . . . . - . . . . . = _ - - _ - - . . __ ._ -.
/
, 'q- ww TABLE 150-5 s.
,,' - RAD;0 LOGICAL. CONSEQUENCES OF THE STEAM GENERATOR
,'- - - ;f g BE RUPTURE WITH A LOSS OF OFFSITE POWER s
AND STUCX OPEN ADV Location Offsite Doses, Rems GIS PIS
. , ~1 ,
1.*- Eichsioi Are Boundary 60 190 0-2 hr. 'ihyf0id
- 2. ' Low PopuIation Zone Outer 15 22
-(
Boundary 0-8 hr. Thyroid w
? * '
't
,r .
, ~
! 46 , , ,e
$ , ** w \
\ -
g p,
% v p %
9 I
t'
\
56 y g. g $
- 3 en v %q +
% +* .
s
.g 4
,i; w
'*l ,.
.- g ,
f 4 ,
Y <
,* f' 1
i'i \;r
.l bh s
, , qL F
! P 'k, d \
g, . " . ,
I i ._.
l l
l ll . ,
i
~
S
, .- .-r, ; ,n , ., , e . n , - - - - - -
120 , , i 100 80 -
i-5 S
t 5
a.
$ 40 20 i i i 0
0 100 200 300 400 TIME, SECONDS C-E STEAM GENER ATOR TUBE RUPTURE WITH LOSS sgur.
OF 0FFSITE POWER AND A STUCK OPEN ADV 150-1A CORE POWER vs TIME l
120 i i i i i 100 -
80 5
M 60 -
5 c.
U -
g 40 20 -
L . , , , .
0 0 5000 10000 15000 20000 25000 30000 TIME, SECONDS l
C-E STEAM GENER ATOR TUBE RU PTURE WITH LOSS Rsure i l
g 0F 0FFSITE POWER AND A STUCK OPEN ADV CORE POWER vs TIME 15D-1B l
l l
- i 2200 , , i 2100 2000 j S
uf 8 1900 m
!O E
h1800 1700 -
I ' '
1600 O 100 200 300 400 TIME, SECONDS 1 \
l C-E STEAM GENER ATOR TUBE RU PTURE WITH LOSS Figure 0F OFFSITE POWER AND A STUCK OPEN ADV g RCS PRESSUREys TIME 15D-2A
1 l l l l
OPERATOR TAKES CONTROL 0F PLANT-OPENS ONE ADV
/INHPSI EACH SG TO RCS FLOW INITIATED 2000 -
' ISOLATED, AUXILIARY FEEDWATER FLOW TO AFFECTED SG OPERATOR ATTEMPTS TO CLOSE ADV OPERATOR CLOSES BLOCK VALVE ON STUCK OPEN ADV 5 /-~
-0PERATOR INITIATES AUXILIARY SPRAY -
T1500 d OPERATOR CONTROLS AUXILIARY SPRAYFLOW, US
,
- BACKUP PRESSURIZER HEATER OUTPUT -
HgSI FLOW IN OR DER TO KEEP g1000 F SUBC00 LED RCS REACHES 500 SHUTDOWN COOLING -
ENTRY CONDITION 5 b
0 30000 0 5000 10000 15000 20000 25000 ,
TIME, SECONOS l
l l
c-E STEAM GENERATOR TUBE RUPTURE WITH LOSS Figure 0F OFFSITE POWER AND A STUCK OPEN ADV g RCS PRESSURE vs TIME 15D-2B
l 660 i i i 640 E'
g'620 HOT LEG E
< I 5
$ 6T - -
g -.
$ AVERAGE '
h580
$$ J g ,
560 -
COLD LEG -
0 lb 200 3 400 TIME, SECONDS C-E STEAM GENER ATOR TUBE RU PTURE WITH LOSS sgur.
OF OFFSITE POWER AND A STUCK OPEN ADV E CORE COOLANT TEMPERATURES vs TIME 150_3A
. \
650 i i i i i
600 550 -
O g' 10T LEG e:
@500 '
W -
p 450 -
, 5 AVERAGE 8
u
$400 o
=
! COLD LEG-350 -
l 3%
0 5000 10000 15000 20000 25000 30000 TIME, SECONDS c.E STEAM GENER ATOR TUBE RUPTURE WITH LOSS sgur.
OF OFFSITE POWER AND A STUCK OPEN ADV
~
CORE COOLANT TEMPER ATURES vs TIME 15D-3B
650 , , , i i 600 -
O g'550 -
E 5
a.
3 500 -
e -
u 5 450 -
i a
400 -
350 0 5000 10000 15000 20000 25000 30000 TIME, SECONDS C-E STEAM GENER ATOR TUBE RU PTURE WITH LOSS Figure gg 0F OFFSITE POWER AND A STUCK OPEN ADV U PPER HEAD TEMPERATURE vs TIME 15 D-4
l 600 i i i 500 -
g400 -
5 S
m -
E 3% -
5:
5 3
h200 0
E 100 -
0 0 100 200 300 400 TIME, SECONDS ,
l l
l C-E STEAM GENER ATOR TUBE RU PTURE WITH LOSS Rsure OF OFFSITE POWER AND A STUCK OPEN ADV 15D-5A I PRESSURIZER WATER VOLUME vs TIME
2000 i i i i i OPERATOR CONTROLS AUXILIARY SPRAY 1500 - BACKUP PRESSURIZER HEATERS AND HPSI _
cn FLOW TO MAINTAIN 200F SUBC0dLING t:
m I -
1000 e
[500 -
m E
5?
10 0 g U'0PERATOR INITIATES AUXILIARY SPRAY l
l
-500 * ' ' ' '
0 5000 10000 15000 20000 25000 30000 TIME, SECONDS C-E STEAM GENERATOR TUBE RUPTURE WITH LOSS sgur.
OF OFFSITE POWER AND A STUCK OPEN ADV 15 D-5B PRESSURIZER WATER VOLUME vs TIME l
l
2000 , ,
/
N g TOP OF RV g1600 -
e LU E
= -
81200 h800 d
B a
a E 4@ -
cr d
TOP 0F HOT LEG 0
O 100 200 300 400 TIME, SECONDS l
l l
C-E STEAM GENERATOR TUBE RUPTURE WITH LOSS Rsure I
/ OF OFFSITE POWER AND A STUCK OPEN ADV LIQUID VOLUME ABOVE TOP OF HOT LEGS vs TIME 15D-6A l :
O 2000, , . . . i TOP OF RV "iL
)-1600-I o
"i 8
x -
8 1200 8
g 800 -
d cr 400 -
D '
TOP OF HOT LEG 0
0 5000 10000 15000 20000 25000 30000 TIME, SECONDS l
C-E R9ure STEAM GENER ATOR TUBE RU PTURE WITH LOSS g 0F 0FFSITE POWER AND A STUCK OPEN ADV 15D-6B LIQUID VOLUME ABOVE TOP OF HOT LEGS vs TIME
~
e l i i 600000 i i i 560000 -
1 2
N 520000 E
e a
y480000 e
440000 -
400000 0 5000 10000 15000 20000 25000 30000 TIME, SECONDS I
C-E STEAM GENERATOR TUBE RU PTURE WITH LOSS Rswe g 0F 0FFSITE POWER AND A STUCK OPEN ADV RCS LIQUID MASS vs TIME 15 0-7
, 1
1400 , , ,
1350 1300 0 0 0 130 -
d a
w
[x 1200 c3 V5 1150 1100 0 100 200 300 400 TIME, SECONDS c-E STEAM GENER ATOR TUBE RU PTURE WITH LOSS Figure g 0F 0FFSITE POWER AND A STUCK OPEN ADV STEAM GENER ATOR PRESSURE vs TIME 15D-8A
1400 i i i i i 4 1200 1000 -
~
h kf AFFECTED SG g p 7 5?
Q%
o.
<3 a e 400 -
I -
f 200 -
Ui1AFFECTED SG 0
0 5000 10000 15000 20000 25000 30000 TIME, SECONDS C-E STEAM GENERATOR TUBE RU PTURE WITH LOSS sgur.
g 0F OFFSITE POWER AND A STUCK OPEN ADV STEAM GENER ATOR PRESSURE vs TIME 15D-8B
2700 i i i S -
C 230 -
5a N -
E 1800 -
5 5
o S -
5 1350 -
M 5
c.
900 -
s AFFECTED SG 450 - _
i UNAFFECTED SG s
0 0 100 200 300 400 TIME, SECONDS c-E STEAM GENER ATOR TUBE RUPTURE WITH LOSS sgur.
0F 0FFSITE POWER AND A STUCK OPEN ADV 150-9A FEEDWATER FLOW PER S.G. vs TIME
200 i i . . i 8
m f160
!3 -
5120 .
5 E-S 80 -
5 d
x W -
g 40 -
Q 0
0 5000 10000 15000 20000 25000 30000 TIME, SECONDS C-E STEAM GENER ATOR TUBE RUPTURE WITH LOSS Rsure OF 0FFSITE POWER AND A STUCK OPEN ADV 150-9B FEEDWATER FLOW TO THE INTACT SG vs TIME
l 1
~
55 , r ,
l 50 -
i
$45 -
3 i l uI f 40 4
"i s -
a 35 30 25 0 100 200 300 400 j
' TIME, SECONDS l
Figure C-E STEAM GENERATOR TUBE RUPTURE WITH LOSS g 0F OFFSITE POWER AND A STUCK OPEN ADV TUBE LEAK R ATE vs TIME 15D-10A l
N l 1 i i I 50 -
S 40 -
- y. i ko: 30 -
(
E \
E 20
(< %w
= \
10 0
0 5000 10000 15000 20000 25000 30000 TIME, SECONDS l
l l c-e STEAM GENERATOR TUBE RU PTURE WITH LOSS Rsure l OF OFFSITE POWER AND A STUCK OPEN ADV 15D-10B TUBE LEAK R ATE vs TIME l
L -
30000 i i i 25000 w
$ 20000 -
5 d
x
$ 15000
/
S ti 5 -
W 10000 -
E 5000 i i '
0-0 100 200 300 400 TIME, SECONDS l
C-E STEAM GENERATOR TUBE RUPTURE WITH LOSS sgur.
OF OFFSITE POWER AND A STUCK OPEN ADV 150_11A SEf8k INTEGR ATED TUBE LEAK vs TIME l
- a 1,800,000' , , , , ,
1,500,000 -
@1,200,000 Bi o
d 4
900,000 -
8 E
600,000 -
E 300,000 -
0 0 5000 10000 15000 20000 25000 30000 TIME, SECONDS i
C-E STEAM GENER ATOR TUBE RU PTURE WITH LOSS Figure l
! 0F OFFSITE POWER AND A STUCK OPEN ADV 150-11B
! INTEGRATED' TUBE LEAK vs TIME l
1 O
l 0.20 , , , , ,
0.16 -
8 '
Bi
$0.12 -
u Ei -
g 0.08 -
E I
EE 0.04 --
Q N (\ Y 0 5000 10000 15000 20000 25000 30000 TIME, SECONDS l
l l _
C-E STEAM GENERATOR TUBE RUPTURE WITH LOSS Figure l CF 0FFSITE POWER AND A STUCK OPEN ADV 15 D-12 l FRACTION OF LEAK FLASHED vs TIME
1 l
I 220000 i i 200000 180000 -
[ 160000 q -
) E s
ci vi 140000 -
AFFECTED SG / -
/
120000 -
UNAFFECTED SG 100000 0 100 200 300 400 TIME, SECONDS 1
C-E STEAM GENERATOR TUBE RUPTURE WITH LOSS Rsure OF OFFSITE POWER AND A STUCK OPEN ADV 150-13A j S.G. MASS vs TIME l
l i i 400000 i i i AFFECTED SG 350000-UNAFFECTED SG 300000 -
'e a
vi 250000 - }
9 w
200000 -
I -
150000 -
I i
100000 25000 3D000 0 5000 10000 15000 20000 TIME, SECONDS C-E STEAM GENERATOR TUBE RUPTURE WITH LOSS Figure OF OFFSITE POWER AND A STUCK OPEN ADV 150-13B E S.G. MASS vs TIME
.-. . _ - _ . . .- A
i
. l
~ ,
1 l
60000 , , ;
S
$ 50000 -
Bi 9
m :
e
/
$40000 5
W E
[> 3M%
s Oi20000 3
W w
(10000 s
l 0 0 100 200 300 400 TIME, SECONDS C-E STEAM GENERATOR TUBE RUPTURE WITH LOSS Rsure g 0F OFFSITE POWER AND A STUCK OPEN ADV MSSV INTEGRATED FLOW vs TIME 15D-14
1,800,000 i i i i i 1,500,000 -
1 b1,200,000 -
N d
h 900,000 -
e E
5 jo! 600,000 -
l 5
300,000 - -
1 0 i i I ' '
0 5000 10000 15000 20000 25000 30000 TIME, SECONDS C-E STEAM GENER ATOR TUBE RU PTURE WITH LOSS Figure OF OFFSITE POWER AND A STUCK OPEN ADV 15 0-15 INTEGRATED ADV FLOW vs TIME --
y_ ------ m ---- mr - r
~W* - - -
e m- y
Figura 150-16 OPERATOR ACTIO!! DURI?iG SGTR + LOP + STUCK OPE!i ADV
- RCS C00LOOWil OPERATOR U$CS ADVs TO COOL <
0 0 RCS TO 550 F AT 100 F/hR l
t ATTDtPTED ISOLATI0ft OF AFFECTED SG OPERATOR ATTE!! PTS TO ISOLATE AFFECTED SG.
THE ADV IS ASSL7CO 70 OTICK CPE" o
CLOSURE ^F ADV BLOCK VALVE OPERATOR CLOSES THE BLOCK VALVE ASSOCIATED WITH THE STUCK OPE!! ADV v
FILL PRESS'J3fIER CPERATOR LSES AUXILIARY SPDAY TO FILL PRESSURIZE?.
h C0:iTROL RCS C00L 0Wr! l 0
~ 20 F/HR l h i T00 RAPID TOO SLOW CLOSE A0Vs 0:1 OPEN ADVs C:4 UNAFFECTED SG llNAFFECTE3 SG t v CONTROL RCS SUBC00 LING Also PRESSURIZER LEVEL 0
ST 3gg = 20 F o o j
700 HIGH TOO LOW AUX SPRAY Oft AUX SPRAY OFF HEATERS OFF HEATEPS Ort DECREASE HPSI IriCREASE HPSI h h g,ROL C AFFECTED SG LEVEL 4
OPEN AFFECTED SG ADV AS NEEDED '
u
! C00LDOWil AND OEPRESSURIZE RCS TO <400 PSIA AtiD < 350 F
, . - , - - -