LD-87-053, Forwards Revised Steam Generator Tube Rupture Analysis for Fsar.Rev Incorporates Revised Analysis Using Reactor Coolant Gas Vent Sys for Depressurization Purposes
| ML20235F445 | |
| Person / Time | |
|---|---|
| Site: | 05000470 |
| Issue date: | 09/18/1987 |
| From: | Scherer A ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | Miraglia F NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation |
| References | |
| LD-87-053, LD-87-53, NUDOCS 8709290126 | |
| Download: ML20235F445 (45) | |
Text
__
E COMBUSTION ENGINEERING September 18, 1987 LD-87-053 ;
Docket No. STN 50-470F Mr. Frank J. Miraglia l i Associate Director for Projects Office of Nuclear Reactor Regulation Attention: Document Control Desk U.S. Nuclear Regulatory Commission )
Washington, DC 20555 l l
Subject:
Auxiliary Pressurizer Spray System for CESSAR-F
Reference:
(A) C-E letter LD-87-009, A. E. Scherer to F. J. Miraglia, i dated February 6,1987 j (B) NRC letter, G. W. Knighton to A. E. Scherer, dated j March 24,1987 !
Enclosure:
CESSAR-F, Appendix 15D-Steam Generator Tube Rupture With Loss of Offsite Power and Single Failure 1
Dear Mr. Miraglia:
Enclosed with this letter is a revised Steam Generator Tube Rupture (SGTR) analysis for the Combustion Engineering Standard Safety Analysis l Report - FSAR (CESSAR-F). This analysis is intended to close out the i
" auxiliary pressurir.er spray" issue for System 80 R as it pertains to our $
Final Design Approval (FDA-2). Previously, Combustion Engineering had j provided additional information [ Reference (A)] in order to complete NRC j review of the CESSAR-F confirmatory issues on natural circulation i cooldown and SGTR. As indicated in Reference ( A), the Staff's l conclusions regarding the natural circulation cooldown issue and compliance j with Branch Technical Position RSB 5-1 for Class 2 plants remains valid l when considering the proposed reliability improvements in the Auxiliary i Pressurizer Spray System ( APSS). j For the SGTR analysis, however, the NRC responded [ Reference (B)] that i insufficient information had been provided to determine wlaether the Auxiliary Pressurizer Spray Systern as modified, and its associated water supply, are safety grade. Specifically, the Staff commented that if the APSS is being credited in the SGTR analysis, then it should be designed to be safety grade.
In order to comply with the Staff's requirements, therefore, a proposed j revision to CESSAR-F Appendix 15D is forwarded for your review. This i revision incorporates a revised SGTR analysis which uses the Reactor I l
Power Systems 1000 Prospect Hdl Road (203) 688-1911 Coinbustion Engineering. Inc. Post Office Box 500 Telex: 99297 8709290126 070919 W;ndsor, Connecticut 06095-0500 0 PDR ADDCK 05000470 A PDR lg l
Mr. Frank J. Miraglia LD-87-053 September 18,.1987 Page 2 Coolant Gas Vent System for depressurization purposes. The system is safety grade and fully complies with single failure criteria requirements.
The enclosed analysis shows that although use -of the Reactor Coolant Gas Vent System rather than the Auxiliary Pressurizer Spray System results in a slightly higher dose, it is still well within the 10 CFR 100 guidelines.
We trust that you will find the material in the enclosure sufficient to closeout the auxiliary pressurizer spray issue for CESSAR-F and FDA-2.
If you have any questions on the enclosure, please feel free to contact me or Mr. T. L. Cameron of my staff at (203) 285-5217.
Very truly yours, COMBUSTION ENGINEERING, INC.
. m.
A. . Scherer Director Nuclear Licensing AES:ss Enclosure
s APPENDIX 15D STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF 0FFSITE POWER AND SINGLE FAILURE 15D.1 IDENTIFICATION OF EVENT AND CAUSES The steam generator tube rupture (SGTR) accident is a penetration of the I 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 noncondensible 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 in dump system,valves (ADVs),
and auxiliary pressurizer feedwater system.backup heaters,
% pressauxiliary wi w ga cspray, ya, safety $jectio 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 following 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 stearaing would be in addition to that which may be required to
)
15D-1 hrntat No.1 Esbruary 27, W84 1
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 die'sel _ 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 initiate alarms and inform the operator of abnormal activity levels and that 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 indicating a leak and by level response in the affected steam generator.
150.2 SEQUENCE OF EVENTS AND SYSTEMS OPERATION Table 150-1 presents a chronological list of events which occur during the steem 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 snutdown 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 15D-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 150-16.
- 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 from the EPG 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 i 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 l 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. I q
- 2) The time delay for the first operator action identified above is consistent I with the guidelines of Reference 2 which recommends a 5 minute cperator )
action delay time for a steam generator tube rupture, and an additional i two minutes for completion 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.
1SD-2 MC&* MO A SW Febra ry 27. 1981 l l
L.
f l
1 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 -
L that the ADV has not closed by the following signals:
l-
-(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 y,sclose the valve, W %Y gresswsur
- 5) The operator Litie,.es aux'i2ry spr y #5 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 4
will u V4,d spray:se to the HPSI control RCSsystem, pressurizer inventory backup heaters, and aux ' f ryywatrps.r and subcooling. V
- 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 removed 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 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 Refercnce 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 event are as follows:
150-3 "^" ~ "^ o February ?7, ' coa l
Reactivity Control:
l 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 ,
watertank(RWT). Once the plant parameters have been stabilized, the operator 1 adjusts the boron concentration to insure that a proper negative reactivity ;
shutdown margin is achieved prior to cooldown. The boron concentration is l adjusted by manually throttling the HPSI discharge valves to replace RCS volume shrinkage.
Reactor Heat Removal: j During the pre-trip part of the transient, reactor heat removal is accomplished in the normal 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.
Prinary System Integrity: G res w N ec M After initiating cooldown procedures, the operator must reestablish the pressur-izer w During the cooldown phase, the pressurizer heaters, d 'i ry J ~ l vads.paer,ater and level HPSI .pumps are used by the operator to control the RCS pressure and I level. i 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 I the level.
The operator will take control of the plant und 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 i requires a larger valve opening area in comparison to that for less rapid j 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 q secondary steam activity alarms in order to identify the affected steam i gene rator, l
150 4 Men h t No. 4 Febrary 27 1984 I
-- _ _ _ _ _ _ _ _ _ _ _ _ _ . I
.- 4 9
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 alarming from the stack i
radiation monitors will alert the operator 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. i Radioactive Effluent Control:
A Containment Isolation Actuation Signal (CIAS) is generatec' 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 80P systems. See Applicant's FSAR for details. l 150.3 ANALYSIS OF EFFECTS AND CONSEQUENCES i
15D.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 control of the olant and a CESEC-III based cooldown algorithm thereafter. The CESEC-III computer program is described t in Reference 3. The thermal margin on DNBR in the reactor core was evaluated using the TORC computer program (Reference 4) as described in q Section 15.0.3 with the CE-1 critical heat flux correlation described in l CENPD-162 (Reference 5). i i
B. Input Parameters and Initial Conditions 1 l
The initial conditions and input parameters employed in the analyses of l 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 I are provided in Section 15.0. Conditions were chosen to maximize the 1 radiological releases. !
The initial reactor operating conditions were varied over the cperating 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 considered in order to deterTnine 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 f)
-,u.
[
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 temperature range limit.
C. Results The dynamic behavior of important NSSS parameters following a steam generator tube rupture is presented in Figures 150-1 to 15D-15. 9 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 150-5). At 570 seconds a safety injection actuation signal Cs generated, and the safety 10 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 I approximately 4020 seconds. Due to flashing caused by the depressuriza-tion .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 15D-6). Consequently, the RCS prer ;ure (Figure 15D-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. Subseouent 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. g 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 Emergency Feed- l water Actuation Signal (EFAS) is generated resulting in, auxiliary feedwater flow which acts to restore the SG level.
l At 460 seconds the operator takes control of tne plant and opens one ADV on each SG to cool down the plant. This is consistent with the EPGs. At 2100 seconds the RCS has been cooled to 550*F. The operator isolates the L auxiliary feedwater to the affected generator, closes the main steam i isolation valves of both steam generators, and attempts to close the ADV of the affected generator. The operator recognizes that the ADV did not 1 close and has the appropriate block valve closed within 30 minutes. The operator then initiates an orderly cooldown by means of the atmospheric I
150-6 t 10 l
v ..
I dump valves and the auxiliary feedwater flow to the unaffected steam
. generator.. Thereafter, the operator will steam the affected 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 l 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 15D-14 gives the main steam safety valve integrated flow rates versu:; 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 F/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 20jF/hr. For the first two hours following the initiation of the event, 48f000 lbms of steam are released to the environment through the atmospheric dump valves. For the two to eight hour cooldown period an additional ^73000-lbms of steam are released via the atmospheric dump valves. NM 150.3.2 RADIOLOGICAL C0!! SEQUENCES 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 /
generator.
The steam generator pressure will increase rapidly, resulting in steam discharge as well as activity release through the main steam safety valves. Venting from the affected steam generator, i.e. , the steam generator which experiences the tuce rupture, continues until the secondary syetem pressure is below the main steam safety valve setpoint. After 460 seconds, the operator initiates a plant cooldown at the technical specifi-l cation cooldown rate (100*F/hr) using the steam generators, atmospheric dump valves, and the emergency feedwater system. The technical specification 15D-7 h end s 2 .~9 I Is mary 27, 100?
i e ..
. 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.
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
, iodine and noble gas fission product activity available for release to 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 for the primary system (1.0 uC1/gm) and secondary system (0.1 pC1/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 opetation.
- 4. A spiking factor of 500 is employed for the GIS.
- 5. For the PIS condition, the technical specification limit (60 pCi/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 stea;n is assumed to be released to the atmosphere with a decontamination factor (DF) of 1.0.
- 8. A DF 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-N000& lbm and BM006 lbm, respectively.
330fte $331%
- 10. The agmosphegic dispersion factors employed in the analyses are: 3 x 10- sec/m for the exclusion area boundary and 1.5 x 10~4 sec/m for the low population zone.
150-8 le+ent 5 9 Fiur uai j 27,- 1SIL4, .
1
- 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 mathematical.model employed 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 15D-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 from the rest of the RCS.
The upper head region liquid level remains above the top of the hot leg throughout the transient. Natural circulation cooldown is not impaired during the transient.
15D-9 Amendment No. 9 February 27, 1984
150.5 = REFERENCES 1). " Combustion Engineering Emergency Procedure Guidelines", CEN-152, l, ' Revision Ol', 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.
l
- 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.
i i
l l
l 15D-10 Amendment No. 9 {
February 27, 1984
L TABLE 150-1 (Page 1 of 2)
SEQUENCE OF EVENTS FOR A STEAM GENERATOR TUBE I
RUPTURE WITH A LOSS OF OFF5ITE POWER AND STUCK OPEN ADV
(
Time Setpoint (Sec) Event or Value Success Path 0.0 Tube Rupture Occurs ---
9 40 Third Charging Pump karted, 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 47.15 Trip Breakers Open --- Reactivity Control 48 Turbine / Generator Trip . ---
Secondary System Integrity 10 Reactivity Control 51 Loss of Offsite Power ---
52 LH Main Steam Safety Valves open, 1265 Secondary System Integrity psia 9 52 RH Main Steam Safety Valves open, 1265 Secondary System Integrity psia 56 Maximum Steam Generator Pressures 1330 Both Steam Generator, psia 97 Main Steam Safety Valves Closed, 1190 Secondary System Integrity 11 i psia J 121.0 Steam Generator Water Level 25 Secondary System Integrity Reaches Emergency Feedwater !
Actuation Signal (EFAS) Analysis
- l Setpoint in the Unaffected Generator, percent wide range 10 122.0 EFAS Generated ---
131.0 Steam Generator Water Level 25 Secondary System Integrity Reaches EFAS Analysis Setpoint in the Affected Generator, percent wide range Amendment No. 11 August 30, 1985
TABLE 15D-1 (Page 2 of 2) 9 Time- Setpoint (Sec) Event or Value Success Path a
132.0 'EFAS Generated ---
Emergency Feedwater Initiated to Secondary System Integrity 10 167.0 ---
Unaffected Steam Generator 177.0 Emergency Feedwater Initiated to --- Secondary System Integrity affected Steam Generator 460 Operator Initiates Plant Cooldown ---
Reactor Heat Removal by Opening One ADV on each SG g 546 Pressurizer Empties ---
570 Pressurizer Pressure Reaches 1578 Reactivity Control Safety Injection Actuation Signal (SIAS) Analysis Setpoint, psia 10 570 Safety Injection Actuation Signal Reactivity Control Generated
.570 Safety Injection Flow Initiated --- Reactivity Control 2100 Operator Attempts to Isolate the 550 Secondary System Integrity Damaged Generator, RCS Tem., 'F 3900 Operator Closes the ADV Block --- Secondary System Integrity Valve ~
9 C
4020 Operator t f" =ens
+ ~ f'ress
^ - " 4 = ry Primary System Inventory Spr:y Fle" 'A d'/ Ved
[4500 Operatorhontrols dumbemy. 20 Prinary System Integrity T- _ "rg Backup Pressurizer Heater Output and HPSI Flow to Reduce RCS Pressure and Control subcooling, F 28,800 Shutdown Cooling Entry Conditions 400/350 Reactor Heat Removal Reached, RCS Pressure, psia /
Temperature, 'F CAeaes fMS*' art /p y o' va, %
w=t ~ x 46uq
TABLE 15 D-2 (Shset 1 of'2)
DISPOSITION OF NORMALLY OPERATING SYSTEMS.
FOR' THE STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF 0FFSITE POWER AND STUCK OPEN ADV f ' &f e
'\]\\e'$Qj\ , ,
SYSTEM b
- 1. Main Fee'dwater Control System /
- 2. flain Feedwater Pump Turbine Control System * /
- 3. Turbine-Generator Control System * /
.5. Pressurizer Pressure Control-System /
- 6. Pressurizer Level Control System /
- 7. Control Element Drive Mechanism Control System /
- 8. Reactor Regulating System /
- 9. . Core Operating Limit Supervisory System /
- 10. Rcactor Coolant Pumps /
- 11. Chemical and Volum Control System /
- 12. Secondary Chemistry Control System * / 1
- 13. Condenser Evacuation System * / j
- 14. Turbine Gland Sealing System * / l
- 15. Nuclear Cooling Water System * /
- 16. Turbine Cooling Water System * /
- 17. Plant Cooling Water System * / l
- 18. Condensate Storage Facilities * /
- 19. Circulating Water System * /
- 20. Spent Fuel Pool Cooling and Clean-Up System * /
- 21. Non-Class lE (Non-ESF) A.C. Power * /
22.. Class lE (ESF) A.C. Power * / j
- Balance-of-Plant Systems - Ammondment No. 9 February 27, 1984 I ,
l E_ _ _ _
/ <
TABLE 150-2 (CONTINUED) (Sheet 2 of 2)'
DISPOSITION OF NORMALLY OPERATING SYSTE!45
- FOR I THE STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF 0FFSITE POWER AND ST'JCK OPEN ADV -
Yo 4,, '$ $2 *g$,
.;\,;'sisp,gj'3 , ,h k $, 'c SYSTEM ,
Fo#
1- 23. Non-Class lE D.C.' Power * /-
- 24. . Class 1E D.C. Power *' /
NOTES:
- 1. ' Portions of this system are isolated, eith'er
' automatically (see applicant's SAR) or manually
.by the. operator.once he determines which stean generator contains.the ruptured tube.
1 i
-1 1
Amnen dnent No. 9 February 27, 1984 I I
. . TABLE 15 0-3 UTILIZATION OF SAFETY:SYSTEl15 FOR THE STEA!1 GENERATOR TUBE RUPTURE UITH A LOSS OF OFFSITE POWER AND STUCK OPEN ADV 9( 'o Y $ #q 4 Y 0
so /s
, ts,l%, $+>,,
& Q #^
SYSTEM # *
- 1. Reactor. Protection System /
- 2. DNBR/LPD Calculator /
- 3. Engineered Safety Features Actuation Systems /-
- 4. Supplementary Protection System
- 5. Reactor Trip Switch Gear /
- 6. .tiain. Steam Safety Valves * /
- 7. Primary Safety Valves
- 8. Main Steam Isolation System * / "
1
- 9. Emergency Feedwater System * /
- 10. Safety Injection System / 'l
- 11. Shutdown Cooling System / 3
- 12. Atmospheric Dump Valve System * /
- 13. Containment Isolation System *
- 14. Containment Spray System *
.15. Iodine Removal System *
- 16. Containment Combustible Gas Control System *
- 17. Diesel Generators and Support Systems * /
- 18. Component (Essential) Cooling Water System * /
- 19. Station Service Water System * /
HOTES:
- 1. The operator manually isolates the affected steam generator.
Anmendment lo. 9 February 27, 1984 1
- Balance-of-Plant Systems - 1
TABLE 15D-4 ASSUMPTIONS AND INITIAL CONDITIONS FOR THE STEAM GENERATOR -
TUBE RUPTURE WITH A LOS5 OF OFFSITE POWER AND STUCK OPEN ADV Assumed Parameter Value Core Power Level, MWt 3876 Core Inlet Coolant Temperature, "F 570 Reacter Coolant System Pressure, psia 2100 6
Core Mass Flow Rate, 10 lbm/hr 155 One Pin Integrated Radial Peaking Factor, with Uncertainty 1.53 J Steam Generator Pressure, psia 1126 Moderator Temperature Coefficient, 10-4 ap/ F -1.1 10 Doppler Coefficient Multiplier 1.0 CEA Worth at Trip, % ap (most reactive CEA fully withdrawn) -10.0 g 4
Amendment No. 10 ."
June 28, 1985 1
TABLE 150 RADIOLOGICAL CONSEQUENCES OF THE STEAM GENERATOR TUBE RUPTURE WITH A LOS5 OF OFF5ITE POWER AND STUCK OPEN ADV Location Offsite Doses, Rems GIS PIS
. 1. Exclusion Area Boundary 0-2 hr. Thyroid
) N
- 2. Low Population Zone Outer Boundary 0-8 hr. Thyroid h
i i
2."'OI.dasy t Nn. { .
7chYuary U, lC3L
___m_-_- _ - m.__ _ . -
l 120- i i i 100 .i 80 - -
+
E .
U.g l
_ - i 5
o.
u 40 - -
20 -
0 ' ' i 0 100 200 300 400 ilME, SECONDS l
Amendment No. 9 February 27, 19R4 c-E STEAM GENER ATOR TUBE RUPTURE WITH LOSS Figure OF 0FFSITE POWER AND A STUCK OPEN ADV 15D-1A CORE POWER vs TIME f
l L______-___
120 i i , , i !
100 -
80 -
i-E u
60 -
5 8
a.
g 40 20 - -
l L , , , , ,
0 -
0 5dOO 10000 15000 20000 25000 30000 TIME, SECONDS I
c-E STEAM GENER ATOR TUBE RU PTURE WITH LOSS Flow.
OF 0FFSITE POWER AND A STUCK OPEN ADV 15D-1B C0RE POWER vs TIME
9 2
l l l )
2100 N -
l .
- I 2000 -
5 ,
E l ul E1900 - - l w i k.
$1800 1700 - -
I O 1 2 3 400 TIME, SECONDS Amendment No. 9 February 27, 1984 )
C-E STEAM GENER ATOR TUBE RU PTURE WITH LOSS Figure OF 0FFSITE POWER AND A STUCK OPEN ADV 150-2A RCS PRESSUREvs TIME
\/ .
t 2500 , , , , ,
OPERATOR TAKES CONTROL OF PLANT { PENS ONE ADV IN EACH SG HPSI FLOW INITIATED TO RCS 2000 -
AUXILIARY FEEDWATER FLOW TO AFFECTED SG ISOLATED, OPERATOR ATTEMPTS TO CLOSE ADV OPERATOR CLOSES BLOCK VALVE ON STUCK OPEN ADV 5 -
^^
-0PERATOR INITIATES PRESSURIZERGRS ~
E1500 VENT FLOW Y OPERATOR CONTR01,5 PRES 50RIZER 3 VENT FLOW, BACKUP PRESSURIZER HEATER OUTPUT, AND HPSI FLOW IN g"1000 -
~
g SUBC00 LED RCS REACHES
- SHUTDOWN COOLING _
500 CNTRY CONDITIONS !
l i i i i i 0 J 0 5000 10000 15000 20000 25000 30000 TIME, SECONDS .l I
i i
l c-E STEAM GENERATOR TUBE RUPTURE WITH LOSS Figwe OF 0FFSITE POWER AND A STUCK OPEN ADV 15D-2B RCS PRESSURE vs TIME
- - - - - - - - -_ _ - - - - - e j
1 660 , i i 640 o
- -~
g620 HOT LEG s
b
$ 6% - -
W I E
5 jAVERAGE g 580 - -
N J 8
560 -
\ COLD LEG -
l l
540 0 100 200 300 400 TIME, SECONDS Amendment No. 9 February 27, 1984 C-E STEAM GENER ATOR TUBE RUPTURE WITH LOSS Figure OF 0FFSITE POWER AND A STUCK OPEN ADV 15 D-3A CORE COOLANT TEMPERATURES vs TIME
650 , , , , ,
600 -
I !
1 550 - -
o' OT LEG f
a::
@500 l m
a.
I h
g450 - _
f u
AVERAGE
$400 - -
u ;
=-
COLD LEG 350 - -
300 0 5000 10000 15000 20000 25000 30000 i TIME, SECONDS l i
C-E STEAM GENER ATOR TUBE RUPTURE WITH LOSS s our.
OF 0FFSITE POWER AND A STUCK OPEN ADV CORE COOLANT TEMPERATURES vs TIME 150-3B
'... ..r J
650 j
^
( -
l 600 - _
r p g550 - -
l
<R f 1
Bn- f :
3 500 - -
u
@450 o.
m 400 - -
I 350 0- 5000 10000 15000 20000 25000 .30000 TIME, SECONDS l
f i
l c - E. STEAM GENER ATOR TUBE RU PTURE WITH LOSS Rour.
OF 0FFSITE POWER AND A STUCK OPEN ADV UPPER HEAD TEMPERATURE vs TIME 15D-4 l l
l I
l
{
5W - - !
l 1
g400
\
3 s
a 7 3M - -
Sc 5
Di Em 2 - -
G u
a.
IM - -
0 0 100 200 300 400 TIME, SECONDS l Amendment No. 9 February 27, 1984 l
C-E STEAM GENER ATOR TUBE RU PTURE WITH LOSS Figure OF 0FFSITE POWER AND A STUCK OPEN ADV 15D-5A PRESSURIZER WATER VOLUME vs TIME
2000 1500 1 OPERATOR CONTROLS PRESSURIZER VENT FLOW i
m BACKUP PRESSURIZER HEATERS AND HPSI g FLOW TO MAINTAIN 200F SUBC0dLING 1 5000 3
s-g 500 -
h_. -
s
\
E0
~
N Gas OPER ATOR INITIATES PREssvRiz ER3VENT FLOW
' ' ' ' ' I
-500 O 5000 10000 15000 20000 25000 30000 TIME, SECONDS Figwr e c-E STEAM GENER ATOR TUBE RUPTURE WITH LOSS OF 0FFSITE POWER AND A STUCK OPEN ADV 150-5B PRESSURIZER WATER VOLUME vs TIME_.
-_.L-.L. - - - - _ - - - . _ _ . _
2000 , , ,
g TOP 0F RV g1600 - -
5 I
l 81200 C
8 g 800 _ .,
d m
E 4@ - -
5 TOP 0F HOT LEG 0 ' ' '
0 100 200 300 400 TIMF, SECONDS Amendment No. 9 February 27, 1984 c-E STEAM GENERATOR TUBE RUPTURE WITH LOSS Rsure OF 0FFSITE POWER AND A STUCK OPEN ADV 150-6A LI9UID VOLUME AB0VE TOP OF HOT LEGS vs TIME
2000- i . . . i TOP OF RV 1 , .
i d 1600 -
o "i
5 x
8 12 - -
s
= 800 - -
d cr 400 - -
D TOP OF HOT LEG h
0 I I ^ ' 1 '
0 5000 10000 15000 20000 25000 30000 TIME, SECONDS l-C-E STEAM GENER ATOR TUBE RUPTURE WITH LOSS Figure OF 0FFSITE POWER AND A STUCK OPEN ADV 150-6B LIQUID VOLUME AB0VE TOP 0F HOT LEGS vs TIME
( .
I I I I 560000 9
[520000 y
e 480000 U
a:: .
440000 400000 0 5000 10000 15000 20000 25000 30000 TIME, SECONDS
)
I c-E STEAM GENERATOR TUBE RUPTURE WITH LOSS Rour.
OF 0FFSITE POWER AND A STUCK OPEN ADV 15 0-7 RCS LIQUID MASS vs TIME
1400 i i i 1350 -
1300 5 h e 130 - -
ul
$ t 0;
"g- 1200 - -
t.b v5 1150 -
i ' '
1100 0 100 200 300 400 TIME, SECONDS Amendment lio. 9 February 27, 1984 c-E STEAM GENER ATOR TUBE RU PTURE WITH LOSS neur.
OF 0FFSITE POWER AND A STUCK OPEN ADV 15D-8A STEAM GENERATOR PRESSURE vs TIME
i 1400 i i i i i I
i
% 1 1200
) ;
1000 - -
[m800 \s u.t \
"U g
~
AFFECTED SG
~
y
<i V5 h
400 -
4
's J
- ~ ~
U(1AFFECTED SG
, w%. . . . . .
0 0 5000 10000 15000 20000 25000 30000 TIME, SECONDS l
C-E -
STEAM GENER ATOR TUBE RU PTURE WITH LOSS s pr.
OF 0FFSITE POWER AND A STUCK OPEN ADV 15D-8B STEAM GENER ATOR PRESSURE vs TIME
l l
2700 i i i S - -
u y2250 S
8 T 1800 - -
5 !
E e
1
@1350 m
b e
AFFECTED SG 450 - _
E UNAFFECTED SG 0 ' ' '
0 100 200 300 400 TIME, SECONDS i
Amendment No. 9 February 27, 1984 C-E STEAM GENER ATOR TUBE RUPTURE WITH LOSS Fleur.
OF OFFSITE POWER AND A STUCK OPEN ADV 15D-9A FEEDWATER FLOW PER S.G. vs TIME
i
'200 v
M 160 -
t U
j 120 q, -
E E 80 _
i ;
- 1.
p< g E
W \
i i ,
O 0
5000 10000 15000-TIME, SECONDS b 25000 30000 C-E STEAM GENER ATOR TUBE RUPTURE WITH LOSS Row. )
0F 0FFSITE POWER AND A STUCK OPEN ADV 15 0-98 I FEEDWATER FLOW TO THE INTACT SG vs TIME
55 , , i 50 - -
D
$ 45 - -
3 kct: '40 - -
h4
.b a cn -
a 35 30 I
25 0 100 200 300 400 TIME, SECONDS Amendment No. 9 February 27, 108a c-E STEAM GENER ATOR TUBE RUPTURE WITH LOSS Figure OF 0FFSITE POWER AND A STUCK OPEN ADV 150-10A TUBE LEAK RATE vs TIME c_
60 i i i i i S
[/,
50 -
S 40 - -
g '
[V e W kx 30 - -
M "g 20 - -
10 0- ' ' '
0 5000 10000 15000 20000 25000 30000 TIME, SECONDS l
c-E STEAM GENERATOR TUBE RU PTURE WITH LOSS Rour.
OF OFFSITE POWER AND A STUCK OPEN ADV I 150-10B TUBE LEAK RATE vs TIME !
L_-__-__-_-
.. + . -
I i
l 30000 , , ,
25000 -
w
$ 20000 .
4 -
$ 15000 Q
W W 10000 - -
z 1
5000 ,
i 1
0/ ' ' '
0 100 200 300 400 1 TIME, SECONDS l Amendment No. 9 February 27, 1984 l'
.C-E STEAM GENERATOR TUBE RUPTURE WITH LOSS Figure 0F 0FFSITE POWER AND A. STUCK OPEN ADV INTEGRATED TUBE LEAK vs TIME 15D-11A
1,800,000 , i 1,'500,000 -
y'1,200,000 i )
3
) <
900,000 - -
i e
E {
600,000 - -
M
~
300,000 -
0 0 5000 10000 15000 20000 25000 30000 TIME, SECONDS l
c-E STEAM GENER ATOR TUBE RU PTURE WITH LOSS Rour.
OF 0FFSITE POWER AND A STUCK OPEN ADV 15D-11B
- INTEGRATED' TUBE LEAK vs TIME
- - s-_ --_- _ _
. ^..
i 0.20 0.16 - -
B Bi 50.12 - - 1 x
8 g 0.08 - -
l 5
J E k S Id 0.04 - - -
1 i A
i h,
k' ' '
AM$%b 0.00 0 5000 10000 15000 20000 25000 30000 l TIME, SECONDS l
i C-E STEAM UENERATOR TUBE RUPTURE WITHRsure LOSS !
0F 0FFSITE POWER AND A STUCK OPEN ADV 15 0 -1?' l FR ACTION OF LEAK FLASHED vs TIME I l
220000 i i i 200000 180000 -
". 160000 -
s ci vi 140000 -
AFFECTED SG 120000 UNAFFECTED SG 100000 0 1 200 300 400 TIME, SECONDS faendment No. 9 February 27, 1984 C-E STEAM GENERATOR TUBE RUPTURE WITH LOSS Rour.
OF 0FFSITE POWER AND A STUCK OPEN ADV 15D-13A S.G. MASS vs TIME
.u ,
400000 1 i i i i l
350000 - -
\ UNAFFECTED SG AFFECTED SG 300000 - - i a
vi
$ 250000 - 1 -
- E c3 V5 200000 -
-150000 - ;
t I
0 5000 10b00 15b00 20000 25000 30000 TIME, SECONDS l
c-E STEAM GENER ATOR TUBE RUPTURE WITH LOSS Roure l
0F 0FFSITE POWER AND A STUCK OPEN ADV 150-13B S.G. MASS vs TIME
.w .
60000 , , i s
9 50000 - -
i 9
m C
$40000 5
/ -
W 5
[> 30000 s
ts
$20000 - -
S 6
m (10000 S
0 0 100 200 300 400 TIME, SECONDS Amendment No. 9 February 27, 1984 c-E STEAM GENERATOR TUBE RUPTURE WITH LOSS Rgure 0F 0FFSITE POWER AND A STUCK OPEN ADV 15 0-14 MSSV INTEGRATED FLOW vs TIME
l
. 1,800,000 1,500,000 -
m
$ 1,200,000- -
i e
$ 900,000 -
$ 600,000 -
E !
300,000 -
l 0 i i i i i 0 5000 10000 15000 20000 25000 30000 TIME, SECONDS c-E STEAM GENERATOR TUBE RUPTURE WITH LOSS Roure 1 0F 0FFSITE POWER AND A STUCK OPEN ADV 15D-15 INTEGRATED ADV FLOW vs TIME i
+
RCS CDOLDGWI OPDATOR USC5 A0Vs TO COOL RCS TO SMF AT 1004/NR A
ATTtHPTID IS& Aft 0N OF AFFFM OPDATOR ATTUFT5 TO 150LAft AFFECTED $6.
THE ADV !! A55GECD 70 :TTCK OPtM EW CLOSURt 0F ADY SLOCK VALVE OPERATOR CLD515 THE SLOCK VALVI A550CIAftD WITH THE 51tCK OPtti ADV FfLL Pets 5URfra CPERATOR Usts $.-
TO FILL PRt55URIZER I
CONTROL RCS C00LDOWI
~ 20'F/HR 700 RAP!D TOO SLOW CLO5E ADVs 03 OPEN A3vs 0:e LMAFFECTED SG 08sAFFECTD %
6 L CONTROL RC1 SUSC00L!NG AIC PR[55URIZER LEVEL l
ATw,, 209 4
fRfSNRIEfk 700 p!Gn ysz tzEt m ve Fr erra m me 700 LOW m'N vin cuuo M(AT[R$ OFF MtATERS On DECREASE =Ps! INCREA5( HP5!
l 4 l
Cya0L AFFECTED 1G LfVtt l
h OPEN AFFECTED SG ADV I As xtEDED l
A {
C00LDOWN AND DEPRE550R!It RC5 I T0 a 400 P51A AND e 350'F i
- ."u,undment NO M Fo p g3ry ?g _}4f OPERATOR ACTION DURING SGTR + LOP + STUCK OPEN ADV 150-ll L__-_______________ _ . ]