ML20070M129
| ML20070M129 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 01/31/1983 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML13309B223 | List: |
| References | |
| CEN-201(S)-S01, CEN-201(S)-S1, TAC-62848, TAC-62849, NUDOCS 8301120308 | |
| Download: ML20070M129 (121) | |
Text
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CEN 201 (S)
SUPP. NO.1 I
N NATURAL CIRCULATION TEST PROGRAM SAN ONOFRE NUCLEAR GENERATING STATION UNIT 2 NATURAL CIRCULATION COOLDOWN Preparedfor Southern California Edison Company NUCLEAR POWER SYSTEMS DIVISION H POWER SYSTEMS COMBUSTION ENGINEERING, INC.
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W' LEGAL NOTICE THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COMBUSTION ENGINEERING, INC. NEITHER COMBUSTION ENGINEERING NOR ANY PERSON ACTING ON ITS BEHALP:
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MAKES ANY WARRAJNY OR REPRESENTATION, EXPRESS OR l
IWWED INCLUDING THE WARRANTIES OF FITNESS FOR A PARTICULAR l
PURPOSE OR MERCHANTAStuYY, WITH RESPECT TO THE ACCURACY, COMPLETENESS, OR USEPULNES3 OF THE INFORMATION CONTAINED IN THIS t
g REPORT, OR THAT THE USE OF ANY INFORMATION, APPARATUS, ANTHOD, OR PROCESS Dines riesn IN THIS REPORT MAY NOT INFRINGE PRIVATELY OWNED RIGHTS;OR p
- 5. ASSUMES AflY UA88UTIES WITH RESPECT TO THE IAE OP.OR FOR DAMAGES RESULTING PROM THE WE OF, ANY INFORMATION, APPARATW, j
METHOD OR PROCESG DICCLOSED IN THIS REPORT.
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. C'EN-201( S)
-.Supp. No..1 0
NATURAL CIRCULATION TEST PROGRAM u,
o SAN ONOFRE NUCLEAR GENERATING STATION UNIT 2
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NATURAL CIRCULATION C00LDOWN e.
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Nuclear Power Systems Division Janua ry,1983
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Combustion Engineering, Inc.
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O' TABLE OF CONTENTS-Section Title Page
- 4' 110 INTRODUCTION AND SU MARY l
2.0 DESCRIPTION
OF TESTS AND PRETEST SIMULATIONS 7
i 3.0
. OPERATIONAL AND TEST TERMINATION CRITERIA 51 7
4.0 IMPACT ON TECHNICAL SPECIFICATIONS 57 6.0 COMPLI ANCE WITH ANALYSIS AND TESTING IN SUPPORT 61
- 0F BTP RSB 5-1
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6.0 REFERENCES
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1.0 INTRODUCTION
AND SU MARY 7
By letter dated 15 April 1982, Southern California Edison (SCE) submitted to the Nuclear Regulatory Commission (NRC) the details of a proposed Natural Circulation Test Program including test procedures and an in depth safety evaluat!on. (See Reference 1.) Following an initial review of the proposed test program, SCE met with various e
members of the NRC Staff on 20 May 1982 in Bethesda to discuss the
. natural circulation tests. At that meeting the Staff provided SCE o
with a copy of guidelines covering various analyses and tests in support of Branch Technical Position (BTP) RSB 5-1.
In view of the uncertainties involved in a plant cooldown and depressurization under natural circulation conditions, these guidelines require that tests with supporting analysis be conducted to confirm that adequate mixing of borated water added prior to or during cooldown can be achieved under natural circulation conditions, to permit estimation of the time required to achieve such mixing, and to confirm that the cooldown under natural circulation conditions can be achieved within the limits specified in the emergency operating procedures. Consequently,'
considerable discu'ssion was generated at the May 20th meeting as to whether or not the SCE Natural Circulation Test Program and in particular Test B3 of that program, see Reference 1, would demonstrate that the plant meets the requirements of RTP RSB 5-1.
At the conclusion of the meeting the NRC Staff agreed to complete their review of the low power or "A" series tests of the Natural Circulation -
y Test Program and SCE agreed to provide the Staff with additional
?nformation regarding their proposed program with respect to the requirements of the branch technical position.
On 8 September 1982, following the successful completion of the "A" series or Low Power Natural Circulation Tests, SCE met with the NRC Staf f at the San Onofre plant site to discuss natural circulation boron mixing and cooldown. At that' meeting detailed information regarding the SCE Natural Circulation Test Program and BTP RSB 5-1 was presented to the Staff. A review of this material has been completed; formal comments by the Staff and requests for additional information 1
are contained in Reference 2.
Two months after the first meeting, a second presentation on 4 November 198?. was made in Bethesda to address-these coments and requests.
This supplement, Supplement No. I to CEN-201(S) (Reference 1) contains the formal response to the comments and requests for additional information made by the NRC Staff. Specifically, Reference 2 contains ten items or concerns with respect to the SCE Natural Circulation Test Program and the analyses and testing in support of BTP RSB 5-1.
A e
summary response to each of the ten items is presented below along with the appropriate section or sections of this supplement where detailed information can be found.
1.
The analysis and testing in support of BTP RSB 5-1 will be met through the performance of essentially five discrete tests.
(Note that a demonstration of the operation of the shutdown cooling system consistent' with the recommer.dations of BTP RSB 5:1 will be performed on San Onofrd's Unit 3 prior to startup
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following the first refueling outage on that unit, see Item 2 below.) Section 5.1 (p. 61) lists these tests along with the data and information that have been obtained or that will be obtained and used in demonstrating compliance with the branch technical position. Adequate provisions have been made in order s
to properly account for differences in the actual test procedure's and the strict BTP RSB 5-1 scenario. All a
such differences, i.e., differences in decay heat levels, fluid temperatures and pressures, etc., will be carefully analyzed and the results presented in the normal post-test report.
2.
The loss of Offsite Power Test, Test B1 & B2, is designed to demonstrate that the plant can be controlled in hot standby following an actual loss of offsite power. The test will not perform a boration 2
.a and subsequent natural circulation cooldown, i.e.,
ev throughout this test the plant will be maintained in hot standby. The Natural Circulation Test, Test B3, will perform a natural circulation boration and cooldown under conditions that simulate a loss of offsite power.
See Table 2-1 (p. 21) for a detailed a
outline of this test. Test 83 will terminate when shutdown cooling system initiation conditions are attained and will not cool the plant to cold shutdown.
o Modifications to the Unit 2 shutdown cooling system to upgrade it to be consistent with the recommendations of BTP RSB 5-1 will be completed during its first refuel-ing outage.
(See response to NRC Ouestion 212.164 of Reference 3.) To perform a cooldown to cold shutdown using this system prior to the intended modifications would not meet the intent of the branch technical.
position. Therefore, this demonstration will be performed on Unit 3, an identical plant in which the modifications to the shutdown cooling system to comply with the branch technical position are in place, at a convenient time prior to startup following the first refueling outage on that unit. After a normal cooldown has first been performed, the demonstration on Unit 3 will consist of taking the plant from shutdown cooling initiation conditions to cold shutdown consistent with s
the recommendations of BTP RSB 5-1.
3.
The main feedwater system must be used to supply water to both steam generators prior to reactor trip in Test B3.
Immediately following the reactor trip, operator action will be taken to secure main feedwater and to take manual control of the auxiliary feedwater system.
Once manual control has been taken, the auxiliary feedwater system will be used to feed the steam generators for the remainder of the test. Since main feedwater will not be used, deviations in natural 3
circulation due to temperature differences will not be a factor during Test B3.
In addition, detailed data on feedwater usage will be taken for post-test analysis as stated in Section 5.2 (p. 64).
4 Letdown will remain in operation during Test B3 in order to prevent the thermal stresses that would be placed on the reactor coolant system charging nozzles e
if it were secured.
In addition, the use of letdown
- will permit continuous on line boronometer readings.
Section 5.3 (p. 65) contains a detailed discussion of the effects of letdown on boron mixing. Specifically, the effects of letdown on charging fluid temperature, baron injection velocity, and stratification are e
discussed. As concluded in that section, there are no significant differences in baron mixing when letdown is operating and when letdown is secured.
5.
Test R3 will perform an ear'y depressurization, i.e.,
the reactor coolant system will be depressurized before the reactor vessel upper head has completely cooled, and a small controlled steam bubole will actually be formed in the reactor vessel upper head region. The purpose of this portion of the procedure is to empirically check the head heat loss predictions for a natural circulation cooldown. This information is vital to the central issue of showing compliance with a
BTP RSB 5-1.
There are no thermocouples currently installed in the upper head; as system pressure is slowly lowered to saturation conditions, however, the exact onset of void formation will be innediately apparent as mass is displaced out of the upDer head into the pressurizer. Void formation will be monitored directly by observing pressurizer level and letdown valve position. Section 2.2.3 (p. 14) contains a description of this portion of the procedure along with 4
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the computer simulation results of a steam bubble formation. Also included in that section is a detailed discussion of the guidelines that will be incorporated into the actual test procedure to ensure the steam bubble is properly controlled.
6.
During the cooldown an'd depressurization portions of Test B3, both atmospheric dump valves will be used to release steam.
In addition, these valves were used to release steam during the Low Power Natural Circulation Tests and will be used during Test 81 & B2. A passive safety-grade nitrogen supply will be used to operate the atmospheric dump valves during Test B1 & B2 and Test 83. The nitrogen supply will be carefully monitored in order to evaluate usage. Atmospheric dump valve operation via the local hand wheel will also be demonstrated.
7.
All pressurizer heaters will be used in Test B3
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immediately following reactor trip to allow for full automatic plant response. Once conditions have stabilized and natural circulation is established, only those heaters capable of being supplied from the emergency diesel generators will be used for the remainder of the test, see Table 2-1 (p. 21).
s 8.
Although not explicitly included as part of Test B3, it is anticipated that the plant will be maintained in hot standby for a period of time that will exceed four hours during which the boron mixing test will be conducted. Condensate usage will be accurately measured and core decay heat will be properly accounted for in the post-test data analysis.
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9.
Reference I contains the initial computer predictions for all of the tests of the Natural Circulation Test 5
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Program. Originally, only the first thirty minutes of Test 81 & B2 and Test B3 were simulated. Subsequent to the issuance of Reference 1, the NRC Staff requested additional simulations for Test B3. These simulations are contained in Section 2.2.3 (p. 14).
In addition, Section 5.4 (p. 76) contains a computer simulation of a strict BTP RSB 5-1 scenario for comparison.
- 10. The information requested by this item is contained in Table 5-1 (p. RI) and Table 5-2 (p. 83).
In conclusion, the combination of five tests contained in the SCE Natural Circulaton Test Program, along with a demonstration of the shutdown cooling system performance on Unit,3 which will be performed prior to startup following the first refueling outage on that unit, will show that the plant can be taken from hot standby to cold shutdown within the constraints of BTP RSB 5-1.
In addition, the tests will provide the necessary information to clearly address the Staff's concerns regarding condensate storage capacity, reactor vessel upper head cooldown rates during a natural circulation cooldown, boron mixing including stratification, the effects of letdown on the boron mixing process, and differences such as decay heat levels, etc.,
between actual test procedures and the strict BTP RSB 5-1 scenario.
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2.0 DESCRIPTION
OF TESTS AND PRETEST SIMULATIONS
-2.1 Basic Description of Tests 2.1.1 Loss of Offsite Power Test (Test B1 & B2) (Test B1 and Test B2 of
. Table I.G.1-1 of Reference 3 were combined into one test; this single test is designated Test B1 & B2 throughout this supplement.)
a Objective - To demonstrate the ability to shutdown and maintain the reactor in hot standby using only emergency power following a loss of offsite power. To demonstrate system operation under conditions that simulate a total loss of AC power.
Method - The reactor is critical at 20 + 1% of rated thermal power with all four reactor coolant pumps operating. Previous power history is such that adequate decay heat is available for natural circula-tion. All plant safety-related and auxiliary loads are powered from their normal sources. At time zero, the reactor is tripped and offsite power is simultaneously interrupted. Emergency diesel operation is demonstrated, natural circulation is established, and plant conditions are stabilized. Selected safety-related loads are securJd to simulate a loss of both offsite and onsite AC power.
(Total loss of AC power is simulated to an extent that will not risk plant damage.)
2.1.2 Natural Circulation Test (Test B3)
Objective - To measure the plant response to a total loss of forced reactor coolant flow from 80% power. To verify natural circulation following the trip from 80% power. To demonstrate that adequate boron mixing can be achieved under natural circulation conditions. To demonstrate the ability to perform a natural circulation cooldown to shutdown cooling system initiation conditions. To evaluate natural circulation flow conditions.
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s Method - The plant is stable at 80 + 0.57. of ra,ted thermal power.
Previous operating history is such that adequate decay heat is available for natural circulation. The steam bypass control system and both feedwater control systems are in automatic. Pressurizer
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pressure and level control systems are in automatic. All plant safety-related and auxiliary loads are being powered from external sources, the reserve auxiliary transformers, to assure a continuous power supply. To initiate the test, the reactor is tripped, the reactor coolant pumps are tripped, and natural circulation is established.
Conditions are stabilized and natural circulation is verified. Boron i
concentration in the reactor coolant system is increased and proper boron mixing is demonstrated. A natural circulation cooldown is then performed followed by a plant depressurization until shutdown cooling system entry conditions are attained. Note that the procedure will attempt an early depressurization, i.e., the reactor coolant system will be depressurized before the reactor vessel upper head _ has cooled, in order to check the accuracy of the upper head' steam bubble predictions.
2.2 Pretest Simulations Reference 1 provided detailed computer simulations of each of the five tests in the SCE Natural Circulation Test Program using the Combustion Engineering Long Term Cooling (LTC) digital computer code. For the Loss of Offsite Power Test (Test B1 & B2) and the Natural Circulation Test (Test B1) the first thirty minutes only of each test were simulated. Additional simulations of Test B3 were requested by the NRC Staff and are presented in Sections 2.2.3 and 5.4 below. For each of the simulations performed by the LTC computer code, certain assumptions concerning plant behavior and operator actions were made.
These assumptions are listed below along with the corresponding simulation results. Except where noted, all assumptions were chosen to model actual plant conditions as accurately as possible.
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2.2.1 Comparison of Pretest Computer Simulations for the Low Power Natural Circulation Tests (Tests A1, A2, and A3) with Actual Test Results The low power portion of the Natural Circulation Test Program, i.e.,
Tests A1, A2, and A3, was successfully completed on Unit 2 of the San Onofre Nuclear Generating Station during the week of 5 September 1982. The test program was prepared consistent with the requirements a
of NUREG 0737'for the purpose of operator training and providing meaningful technical information beyond that obtained in the normal startup test program of Regulatory Guide 1.68 The tests were conducted in three parts using a critical reactor maintained between 1% and 3% of rated thermal power to simulate decay heat as follows:
1.
Natural Circulation Verification (A1).
2.
Natural Circulation at Reduced Pressure (A2).
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Natural Circulation with one Steam Generator isolated ( A3).
Although the primary purpose of the "A" series tests was to provide operator training pursuant to Item 1.G.1 of NUREG 0737, certain data was obtained that will be presented in subsequent sections of this report to show compliance with the testing and analyses required in support of BTP RSB 5-1.
Consequently, a brief summary of the actual "A" series tests is presented below for familiarization. Also included in this summary, where appropriate, will be a comparison of the LTC computer simulation predictions from Reference 1 with actual test data.
Initial conditions for the first sucessful low power natural circulation test run were established in accordance with procedure at approximately 12:30 a.m. on the seventh of September, 1982.
Representatives from the Reactor Systems Branch, Test Review Group, and the regional office of the NRC were present to observe the l
evolution. Conditions at the time of the test run were as follows:
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1.
Reactor power stable at about 2.5% of rated thermal power.
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All four reactor coolant pumps operating.
3.
Group 6 control element assemblies > 60
-o inches withdrawn. All other control element. assemblies full out.
4.
Cold leg temperatures approximately 545'F.
5.
Pressurizer level control system in automatic with pressurizer level approximately-33%.
6.-
Pressurizer pressure control system in automatic with pressurizer pressure approximately 2250 ps,,ia.
7.
Steam generator levels approximately 69% via manual control of motor driven auxiliary feedwater pumps.
8.
Both steam generator pressures approximately 1000 psia via manual control of the e
atmospheric dump valves.
9.
All four plant protective system channels in operation, 10.
In order to maintained the reactor critical with reactor coolant pumps secured and to comply with the safety analysis section of Reference 1, the following reactor protective system instrument line-up was in effect:
(1) High linear power trip reset to 9.25%,
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(2) High log power trip reset to 100%,
(3) 10 #% bistable reset to > 10%, (4) All four' core protection calculators in bypass, and (5) Low reactor coolant system flow trip-setpoint adjusted to allow for. critical oper-ation under zero flow conditions.
e The first portion of the test program, Test A1, was initiated by simultaneously tripping all four reactor coolant pumps. The reactor remained critical to simulate decay heat and all plant parameters-behaved as predicted by tha LTC computer code. Pump coastdown was observed as the plant went into natural circulation. Hot leg temperatures (Thot) increased, cold leg temperatures (Tcold)were initially constant then allowed to decrease, pressurizer level increased, and system pressure increased slightly..At the start of the transient reactor power was approximately 2.5%.
Power began to decrease shortly after pump trip as Thot and hence average loop temperature (Tavg) increased since the plant was operating with a slight negative moderator temperature coefficient at the time of the test. No operator action was taken and power stabilized at approximately 1.5". about ten minutes into the test. This corresponded to a reactor vessel AT of approximately 27'F.
At approximately one hour into the transient, operator action was taken to increase power by withdrawing Group 6 control element assemblies. T began to hot increase and hence T increased. A slight increase in pressurizer 3yg level occurred as a result of system expansion. This increase in pressurizer level along with manual operation of the backup heaters resulted in an increase in system pressure. Operator action was taken to initiate auxiliary spray using one charging pump. When system pressure increased further, a second and a third charging pump were started. System pressure, however, continued to increase even though all three charging pumps were now in operation. This behavior occured since the main spray valves were open on a high pressure signal and charging flow, instead of going tnrough the spray nozzle into the pressurizer, was circulating back through the main spray valves into 11
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the reactor coolant. system cold legs. Both main spray valves were closed allowing charging to flow into the pressurizer and system pressure was rapidly decreased into its normal band.
After natural circulation had been maintained for about two hours, the second portion of the test program, Test A2, was initiated by securing all pressurizer heaters. The test provided data on ambient heat e
losses from the pressurizer, auxiliary spray efficiency, and subcooled margin monitor performance that was not obtained during hot functional testing. Pressurizer level was held constant via the pressurizer level control system operating in automatic and a slow system pressure reduction occurred when all pressurizer heaters were secured. The reduction was followed for about a forty minute period during which time a pressure loss of slightly less than 2 psi / minute was observed.
At the end of this time auxiliary spray flow was initiated, both main spray flow control valves were closed, and data taken on the resulting pressure decrease. The auxiliary spray flow from one charging pump produced a depressurization rate of approximately 20 psi / minute. The test was secured when sy' stem pressure reached about 2000 psia.
Heaters were then energized and system pressure was returned to nornal.
The third and final part of the test program, Test A3, involved a demonstration of natural circulation with one steam generator isolated. Reactor power was reduced to approximately 1.5". and Tcold was lowered to approximately 51R*F in order to prevent lifting the safety valves on Steam Generator 2 following its islolation.
(Stean generator pressure corresponding to a cold leg temperature of 518 F during natural circulation is approximately 800 psia.) The test was initiated by isolating Steam Generator 2, i.e., securing auxiliary feedwater flow, securing blowdown, and shutting the atmospheric dump valve. Reactor power was then slowly decreased to about 1"..
All plant parameters during the resulting transient behaved as predicted by the LTC computer code. Tcold in the operating loop remained ennstant while T in the loop with the isolated steam generator, cold Loop 2, increased as the entire system heat load slowly transferred to 12
Temperature differences at the core exit thermocouples due to flow distributions through the reactor core and incomplete temperature mixing were not observed.
(See Section 5.3 of
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this supplement for a complete discussion of this effect.)
Loop 2 AT, i.e., hot leg temperature minus cold leg temperature in Loop 2, was allowed to decrease to approximately 9 F at which time Steam Generator 2 was returned to service and full natural circulation flow reestablished. Testing and training were successfully completed at about 5 a.m.
The reactor was manually tripped and reactor coolant pumps were restarted.
Initial conditions for testing were re-established approximately twelve hours later and the natural circulation tests were repeated a second time to ensure training of all appropriate site personnel.
Table 2-1 of Reference 1 contains steady-state numerical results for various plant parameters as predicted by the LTC digital computer code.. Gomparisons of the reactor vessel A Ts and core flowrates with actual t'est data are presented in Figure 2-1 (p. 25) and in Figure 2-2 (p. 26). As can be seen from these figures, excellent agreement between the LTC predictions and test results was obtained. In addition, detailed analysis of all test data is in progress with the goal of further benchmarking and fine tuning of the LTC code. A more accurate refinement of certain measured plant parameters and as found plant conditions such as pressurizer ambient heat losses, auxiliary spray temperatures and efficiences, steam qenerator heat transfer charateristics during natural circulation,.and loop resistance to natural circulation flow have been performed and serve to further increase the accuracy of the simulations in Sections 2.2.3 and 5.4 below.
2.2.2 Loss of Offsite Power Test (Test 81 & B2)
For a detailed discussion of assumptions and computer simulation results see Section 2.2.4 of Reference 1.
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2.2.3 Natural Circulation Test (Test B3)
Section 2.2.5 of Reference 1 contains the results of the initial computer simulations of the Natural Circulation Test, Test B3.
Originally, the results for the first thirty minutes only following the reactor trip were presented. Subsequent to the issuance of Reference 1 and the performance of the Low Power Natural Circulation a
Tests at San Onofre's Unit 2, the NRC Staff requested additional and more detailed simulations of Test 83. (See Reference 2.) Specifically, it was requested that the computer predictions be expanded to include the boration, cooldown, and the depressurization portions of the Natural Circulation Test. Figures 2-3 through 2-24 (pp. 27 to 48) contain the results of the expanded pretest computer simulations.
For reference purposes, Table 2-1 (p. 21) contins a detailed outline of the Natural Circulation Test. The initial assumptions concerning system responses and operator actions used for this computer run are (Note that all ad'itional assumptions concerning operator as follows:
d actions during the boration, cooldown, and the system depressurization will be detailed in the discussion below.)
1.
The plant is initially stable at 80% of rated thermal power. Post-trip decay heat values are obtained by multiplying the standard decay heat curve by the initial power level in percent.
(See Figure 2-3.)
2.
Initial steam genera' tor pressure is approximately 910 psia. The steam bypass control system is in automatic.
3.
Emergency feedwater flow at 400 gpm per steam generator and 70*F is initiated when steam generator narrow range level reaches 21%.
4 Main feedwater flow is reduced to 5". per steam generator immediately following the reactor trip then manually secured ten minutes later.
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5.
' Steam generator _ levels are maintained via. manual "0N/0FF" type control of the auxiliary feedwater system. (See Figures 2-23 and 2-24.)
6.
Pressurizer. pressure and level controls ar.e in automatic.
7.
Full power to all plant safety-related and auxiliary loads is available throughout the test except where intentionally secured to simulate the strict BTP RSR 5-1 scenario.-
A.
The transient is initiated at time zero by simultaneously tripping the reactor and all four reactor coolant pumps.
9.
A turbine trip occurs n.5 seconds after the reactor trip.
Following the initial trip, reactor coolant system mass flow decreases rapidly as reactor coolant pump speed decreases. Pump speed will reach zero at about 250 seconds after trip, and full natural circulation flow, see Figure 2 4, will be established approximately ten minutes into the transient. Once full natural circulation flow is established, the reactor coolant system mass flow will steadily decrease as the core thermal output or decay heat level steadily decreases. The small perturbatinns to the mass flow that are seen in L
Figure 2-4 during the boration, cooldown, and depressurization l
portions of the simulation are the result of the transient effects of increases in charging flow. System pressure, initially 2250 psia, will decrease rapidly to about 2075 psia immediately following the reactor coolant pump trip as pressurizer level decreases in response i
to reactor coolant contraction. As natural circulation is l-established, the average coolant temperature will increase as hot leg s
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temperature increases and reactor coolant expansion will occur. The resulting rise in pressurizer level coupled with the automatic action of the pressurizer pressure control system will act to increase system pressure to approximately 2300 psia. All pressurier heaters will be available for use during the actual test for the initial transient to allow for full automatic system response. Once natural circulation is established and plant conditions have stabilized, all pressurizer heaters with the exception of those capable of being supplied from the emergency diesel generators, i.e., the IE pressurizer heaters, will be secured for the remainder of the test. For the computer simulation all non-1E pressurizer heaters are secured one hour after reactor trip just prior to boration, see Figures 2-6 and 2-7.
Pressurizer level decreases rapidly following reactor trip in response to reactor coolant contraction. Level recovery occurs as natural circulation is established and reactor coolant expansion takes place. Mass is also added to the reactor coolant system during this period by the automatic operation of the charging pumps, see Figure 2-9.
Note that since variable letdown flow will be maintained throughout Test B3, a minimum of one charging pump will be in. operation at all ' times and flow will be maintained to the loop charging nozzles except when auxiliary spray is in use. If the variable letdown flow were to be secured, "0N/0FF" type operation of the positive displacement charging pumps would be required to maintain pressurizer level. This type of operation coupled with the absence of charging fluid preheating would result in abnormally high thermal stresses across the loop charging nozzles.
In addition to preheating charging fluid and thus limiting thermal stresses, the use of letdown will greatly enhance data acquisition by permitting continuous on line boron concentration measurements via the boronometer.
Once natural circulatien is established and plant conditions have stabilized, the reactor coolant system will be borated in order to obtain a proper shutdown margin for a plant cooldown. Concentrated boric acid solution is added to the reactor coolant system by charging using the chemical and volume control system. Section 9.3.4 of Reference 3 contains as detailed description of the San Onofre 16
chemical and volume control. system (CVCS); Figure 2-25 (p. 49) contains as simplified CVCS diagram showing the line-up used for boration. Positive displacement type charging pumps each with a capacity of 44 gpm take suction on a concentrated borated water source, the boric acid makeup tanks, via boric acid makeup pumps or a gravity drain line. This solution is then introduced directly into the reactor coolant system through the charging nozzles. The boron a
concentration in the pressurizer is increased as necessary by switching the output of the charging pumps from the reactor coolant i
system' loops to the pressurizer via the auxiliary spray line, see Figure 2-25, with simultaneous operation of the pressurizer heaters to limit depressurization. During Test B3, boron samples at the locations stated in Table 2-1 will be taken in order to verify adequate mixing of the added boron and to permit estimation of the time required to achieve such mixing. As stated previously, letdown will be in operation in order to prevent the large thermal stresses that would be placed on the charging nozzles if it were secured and to enhance data taking during boration.
(See Section 5.3 of this supplement for a de' tailed discussion of the effects of letdown on boron mixing.) For the computer simulation of the Natural Circulation Test, boration is initiated one hour after the reactor trip using two charging pumps. Fifteen minutes after the initir. tion of boration, one charging pump is secured and flow is diverted from the locp charging nozzles to the auxiliary spray line in order to borate the pressurizer.
(See Figure 2-11.) A depressurization rate of slightly less than 20 psi / minute as seen in Figure 2-5 is obtained. Charging flow was then diverted back to the loop charging nozzles and boration is completed ' fifteen minutes later again using two charging pumps.
During the actual boration in Test B3, sevaral cycles of auxiliary spray will probably be required to properly borate the pressurizer.
Only one auxiliary spray cycle is performed for the simulatinn, however, since subsequent cycles would be virtually identical to the first and new Information would not be gained.
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Following boration and verification of proper mixing, a natural
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circulation coolewn must be performed to lower loop temperatures below the desiga temperature of the shutdown cooling system. For this evolution, the atmospheric dump valves will be opened, the steam bypass control system valves will be closed, and a plant cooldown at '
the maximum administrative cooldown rate of 75'F per hour will be initiated.
(Reactor coolant system loop temperatures are shown in e
Figures 2-12 and 2-13.) For Test B3, pressurizer level will be maintained during the cooldown by charging with the charging water boron concentration automatically adjusted by the reactor coolant auto makeup system to prevent boron dilution. Cooling the plant-at the highest rate possible, i.e., the maximum administrative cooldown rate, will create the largest temperature differential between the reactor vessel upper head and the remainder of the reactor coolant system and thus maxmize upper head cooling. During this period, system pressure will be maintained relatively high in order to prevent reaching saturation pressure in the reactor vessel upper head and thus preventing a premature steam bubble formation. Note that during the cooldown and depressurization portions of Tes't B3, the atmospheric dump valves will be used to draw steam from'the steam generators.
In addition, nitrogen usage by these valves will be monitored so that a verification of the capacity of the safety-grade nitrogen supply bottles can be made.
Once the cooldown is complete, a depressurization will be performed to lower plant pressure below the design pressure of the shutdown cooling system. During this part of the test, an early depressurization will be performed, i.e., the reactor coolant system will be depressurized with the reactor vessel upper head temperature still relatively high in order to check the accuracy of the cooldown curve shown in Figure 2-14 The early depressurization will result in a small controlled steam bubble being formed when system pressure reaches the saturation pressure of the reactor vessel upper head region. By knowing the pressure at which the steam bubble initially forms, the temperature of the upper head can be determined, and from this information a cooldown rate can be calculated. Note that during Test B3, the control element 18
drive mechanism (CEDM) blower will be operating to prevent damap to the CEDM windings. Since the CEDM blower is not safety-grade and since its operation will enhance reactor vessel upper head cooling, an appropriate allowance will be made when verifying the upper head cooldown predictions. For the computer simulation, system.
depressurizaticn is initiated thirty minutes after loop cold leg temperature reaches 350*F using auxiliary spray and one charging pump as shown in Figure 2-11.
The depressurization will be rapid as shown In Figure 2-5 becau:e of the relatively large temperature differential between auxiliary spray and the pressurizer steam space. System pressure drops from 2250 psia to approximately 560 psia, the predicted saturation pressure for the reactor vessel upper head, in about 50 minutes at which time steam bubble formation connences. Since the San Onofre Unit 2 reactor vessel currently does not have thermocouples i
installed in the upper head, the exact pressure at which the bubble will begin to form will not be known. As system pressure is slowly lowered to saturation conditions for that region, however, the onset of steam bubble formation wiki be immediately apparent as mass is displaced out of the upper head into the pressurizer. This type of behavior is well known and is documented in Reference 10.
Pressurizer level increases rapidly as mass is displaced out of the reactor vessel upper head as shown in Figure 2-8.
Variable letdown flow, see Figure 2-10, will automatically increase to its maximun rate as the pres-surizer level control system compensates for the level rise. When pressurizer level increases to just above 40%, the following operator a
actions are assumed to take place: 1) Auxiliary spray is secured to
'stop the depressurization, 2) Loop charging 1-s initiated using one charging pump, and 3) The IE pressurizer heaters are energized. With these assumed operator actions, the size of the steam bubble, see Figure 2-15, is limited to approximately 100 ft3 Based upon these results, appropriate test termination and operational criteria have been added to Section 3.0 of this supplement to ensure that the size of the steam bubble formed in the reactor vessel upper head will be 3
less than 300 ft. For comparison, Figure 2-26 (p. 50) shows a side 3
view of the reactor vessel with an 800 ft steam bubble. By follow-ing the criteria specified in Section 3.0 for controlling tne upper 19
3 head' steam bubble, its size will be limited to less than 300 ft which in turn will confine it to well above the reactor vessel upper guide structure.
The final portion of Test B3, once a steam bubble has formed, is to increase system pressure by 100 psi, wait one hour to allow the reactor vessel upper head temperature to decrease further, then e
continue depressurization.
(See Table 2-1. )
If a second steam bubble is formed, depressurization will be stopped, system pressure will be increased by 100 psi, and depressurization will be recommended one hour later. This cycle will be repeated as required until shutdown cooling system entry pressure is obtained. For the computer simulation only one of the above cycles was performed since subsequent cycles would be virtually identical to the first ard no new information would be gained.
9 9
4 e
I 20
Table 2-1 Procedure Outline for the Natural Circulation Test Test B3 Procedure Outline Discussion I.
Initial conditions.
A.
Plant stable at 80 + 5%
A.
Adequate decay heat is of rated thermal power.
available for natural circulation.
B.
All four plant protective system channels are in operation.
C.
Steam bypass control system and both feedwater control systems are in automatic.
D.
pressurizer pressure and level control systems are in automatic.
E.
Plant safety-related and E.
All plant auxiliary and safety-auxiliary loads are related loads will be energized transferred to the reserve throughout the initial auxiliary transformers.
transient. Loss of offsite power will be simulated only during this test. The Loss of Offsite Power Test, Test B1 &
B2, is designed to test the various transfer functions and demonstrate plant behavior under an actual loss of offsite power as required by Regulatory Guide 1.68, Rev. O.
II. Procedure.
A.
Reactor coolant pump A.
During the initial transient trip / natural circulation, and the subsequent boration, steaming will be accomplished via the steam bypass control system. The main feedwater system will be used to supply steam generators pre-trip; the auxiliary feedwater system will be used to supply steam generators post-trip.
21
Table 2-1 (cont.)
procedure Outline Discussion
- 1. Trip reactor.
- 2. Establish natural circulation.
- 3. Stabilize plant and take data.
1
- 4. Verification of natural circulation.
- 5. Measure power-to-flow ratio.
- 6. Secure all but 1E 6.
Only pressurizer heaters that pressurizer heaters.
can be powered from the diesel generators will be used for the remainder of the test to simulate a loss of offsite power.
B.
Baron mixing demonstration.
- 1. Determine required 1.
Amount added will be the boron increase.
greater of 100 ppm or that required for a proper shutdown margin at approximately 350'F.
- 2. Initiate boron sampling.
2.
Sample locations and frequencies are designed to demonstrate adequate mixing and permit estimation of times required to achieve such mixing.
- a. Hot leg #2 - 30 min.
- b. Pressurizer - 60 min,
- c. Volume control tank -
60 min.
- d. Boronometer - 10 min.
- 3. Initiate boration.
- a. Charge to loops using a.
Maximize mixing in the chemical a minimum of two and volume control system and charging pumps.
enhance boronometer readings.
22
Table 2-1 (cont.)
arocedure Outline Discussion
- b. Periodically operate b.
Boron mixing in' pressurizer.
heaters and auxiliary spray simultaneously.
- c. Maintain pressurizer level 33 + 3?..
- d. Acceptance criteria for adequate boron mixing - three successive hot leg samples within + 10 ppm, pressurize 7 and volune control tank within + 10 ppm of reactor coolant system.
average.*
C.
Cool d own.
- 1. Ensure adequate feedwater 1.
Actual requirements to be available.
determined via careful analysis of post-test feedwater inventory data.
- 2. Initiate Hot Leg #2 boron samples and boronometer readings.
- 3. Transfer steam loads to 3
Nitrogen usage to be evaluated atmospheric dump valves.
via careful roonitoring of the Air to be supplied to nitrogen supply bottles, these valves via the safety-grade nitrogen supply bottles.
- 4. Using both atmospheric 4.
Charge as necessary due to dump valves, initiate coolant contraction to cooldown.
maintain pressurizer level.
The concentration of the charging fluid to be maintained via the RCS auto makeup system.
l
- a. < 75*F/hr.
- b. Feed steam generators via the auxiliary feedwater system.
l 23 i
I
Table 2-1 -(cont.)
procedure Outline Discussion
- c. Stabilize plant after c.
Cooldown to a cold leg cooldown complete for temperature of about 350*F.
approximately 30 min.
D.
Depressurization.
D.
The proc 6 dure will perform an early depressurization to check the accuracy of upper head steam bubble formation pre-dictions.
- 1. Maintain pressurizer level control system in automatic if possible.
- 2. Initiate auxiliary spray.
- a. Monitor for reactor vessel upper head steam bubble for-mation.
- b. When steam bubble b.
System
- pressure at the time of formation is steam bubble formation will be indicated, stop used to determine upper hbad depressurization prior temperature and thus cooldown to exceeding a pres-rate.
surizer level of 40%,
increase system pressure by 100 psi, and wait one hour before continuing.
- c. Repeat Step II.D.2 until the plant is depressurized to shut-s down cooling system entry pressure.
E.
System restoration.
E.
Restart one reactor coolant pump and continue to take boron concentration data in order to check for possible boron stratification, e
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3.0 OPERATIONAL AND TEST' TERMINATION CRITERIA-Detailed operational and test termination criteria for the Loss of Offsite Power Test (Test'B1 & B2) and the Natural Circulation Test (Test B3) were presented in Section 4.0 of Reference 1.
For completeness, these criteria are repeated below. As a result of certain procedure changes since Reference 1 was issued, several additional operational and test termination criteria for the "B" series tests have also been specified in this section.
Test termination criteria for the "B" series tests are listed in 3.1 bel ow.
(All numbers represent indicated values.) These criteria represent specific points where manual operator action will be taken to terminate testing in order to ensure the-safe operation of the plant. Opera ~tional criteria for the "B" series tests are listed in 3.2 below. (All numbers represent indicated values.)
The various operational criteria form an envelope within which the operator should try to maintain the plant. While operation outside the envelope defined by the operational criteria is not specifically prohitrited, appropriate action should be taken to regain this envelope as necessary. Compliance with the operational and test termination criteria for Test 81 & B2 and Test B3 will ensure the following minimum conditions for safe operation are met: 1) Sufficient sub-cooled margin exists in the reactor coolant system to ensure adequate core cooling and prevent loop void formation, 2) Sufficient water level in each steam generator exists to ensure an adequate heat removal capability, 3) Sufficient pressurizer level exists to ensure adequate pressure control can be /naintained, and 4) An adequate margin l
l to critical heat flux exists in the core at all times during natural circulation.
3.1 Test Termination Criteria 3.1.1 During Test 81 & B2, terminate testing per procedure if any of the following conditions occur:
51
a.
Any abnormalities occur beyond the scope of the current test.
b.
Proper steam generator levels cannot be maintained during the-simulated loss of all AC power.
c.
RCS pressure cannot be adequately controlled during the simulated loss of all AC power.
d.
Adequate natural circulation cannot be maintained.
3.1.2 During Test B3, terminate testing per procedure if any of the following conditions occur:
a.
Tcold (except during cooldown)
< 500'F.
b.
T
) 600*F.
hot c.
< 50*F.
subcooled margin d.
Adequate natural circulation cannot be maintained.
e e.
Reactor coolant system pressure cannot be adequately controlled, i.e., uncontrolled pressure increases or decreases.
f.
Pressurizer level recovery, i.e.,
pressurizer level decreasing toward its normal band, has not 52
begun by 30 minutes following reactor vessel' upper hrad steam
. bubble formation.
3.2 Operational Criteria 3.2.1 The following set of operational criteria should be met during the performance of Test 81 & B2:
a.
Ensure both emergency diesels start and operate properly, b.
Maintain normal steam generator water levels.
c.
Monitor for void formation.
d.
Ensure auxiliary feedwater. pumps start and operate properly.
e.
Do not restart reactor coolant pumps if a condensible void is known to be present in the reactor vessel upper head.
f.
Boron concentration of any makeup water added to the plant must be the same or greater than reactor coolant system boron concentration to prevent dilution.
3.2.2 The following set of operational criteria should be met during the performance of Test B3:
a.
> 50*F.
subcooled margin s
s, 53 s
b.
Core exit temperature--
1 600*F.
(Highest Thot) c.-
If Tcold f alls below 510*F, except during cooldown, adjust feedwater flow and/or steam flow as necessary to control the temperature decrease.
d.
Boron concentration of any makeup water added to the plant must be the same or greater than reactor coolant system boron concentration to prevent dilution.
e.
Do not restart reactor coolant pumps if a condensible void Js known to be present in the
.the reactor vessel upper head.
f.
Maintain normal pressurizer level prior to depressurization.
g.
Only 1E grade pressurizer heaters should be used for the boration, cooldown, and depressurization portion of the test procedure.
In the event that more heater i
power is required, additional l
heaters should be energized and their usage recorded.
h.
Prior to depressurization, maintain system pressure at approximately 2250 psia or as 54
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jh (,, ~ .i ,,.s x-g '. j. During cooldown, use the safety ~ r g ' " ":o grade nitrogen accumulators with the atmospheric dump valves as ./ M o.s A. necessary to measure nitrogen usage. 3j 3 ^ se t a., Y k. Replenish condensate inventory ..1 as required. t, p l. Do not exceed a reactor coolant s ..s. system cooldown rate of 100*F/hr (30'F if below 280*F). f 9 m. Monitor for reactor coolant system ~ ) ~ ~ void formation. n. Do not exceed a pressurizer cooldown rate of 200*F/hr. I o. In order to limit the size of the reactor vessel upper nead steam bubble during system depressurization, perform the following prior to pres-surizer level exceeding 40%: Secure auxiliary spray and initiate loop charging using one charging pump; l energize IE presurizer heaters. ~, ' i s 55 i 4 w .-r I
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4.0 IMPACT ON TECHNICAL SPECIFICATIONS Based upon a careful review of Reference 4, it was determined that three technical specifications will require exception to allow for the performance of the "B" series tests. Table 4-1 (p. 60) lists the technical specifications that require excepcion along with the tests for which these exceptions will be needed. The following paragraphs note the reasons for the exceptions and the basis for allowing the tests to be performed. 4.1 Reactor Coolant Loops and Coolant Circulation - Hot Standby '(Tech. Spec. 3.4.1.2) The technical specification covering reactor coolant loops and coolant circulation in hot standby requires at least one reactor coolant loop and its associated steam generator with at least one associated reactor coolant pump be in operation. The operation of one reactor coolant pump will provide adequate flow to ensure quick mixing of added bornn, prevent possible temperature stratification, prevent possible boron stratification, and produce gradual reactivity changes during boron concentration reductions in the reactor coolant system. Consequently with no reactor coolant loop in operation while in hot standby, all activities involving a reduction in baron concentration should be suspended and immediate corrective action to return the required loop to operation should be initiated. During the performan,ce of the "B" series tests, operation of both reactor coolant loops will be interrupted by intentionally securing power to all four reactor coolant pumps. In Test 81 & R? the plant will be maintained in hot standby with offsite AC power secured, and in Test 83 the plant will be taken from hot standby to temperatures and pressures that permit the initiation of the shutdown cooling system under conditions that simulate a loss of offsite power. In both tests, however, no operations will be permitted that could cause a dilution in the reactor coolant system boron concentration and an adequate loop subcooled margin will be maintained at all times. l 57
4.2 Reactor Coolant Loops and Coolant Circulation - Hot Shutdown (Tech.' Spec. 3.4.1.3) The technical specification covering reactor coolant loops and coolant circulation in hot shutdown requires at least one reactor coolant loop and its associated steam generator with at least one associated reactor coolant pump or shutdown cooling Train A or shutdown cooling Train B be in operation. The operation of at least one of the above systems will provide adequate flow to ensure quick mixing of added boron, prevent possible temperature stratification, prevent possible boron stratification, and produce gradual reactivity changes during boron concentration' reductions in the reactor coolant system. Consequently with no reactor coolant loop or shutdown cooling train in operation while in hot shutdown, all activities involving a reduction in boron concentration should be suspended and immediato corrective action to return the required loop or train to operation should be initiated. During' Test R3, the plant will be taken from hot standby to temperatures and pressures that permit the initiation of the shutdown cooling system under conditions that simulate a loss of offsite power, i.e., all four reactor coolant pumps secured. The shutdown cooling system will not be placed in operation. In this test, however, no operation will be permitted that could cause a dilution in the reactor coolant system boron concentration and an adequate loop subcooled nargin will be maintained at all times. 4.3 Condensate Storage Tank (Tech. Spec. 3.7.1.3) The technical specification for the condensate storage tanks requires that both tanks be operable with a contained water volume of at least 123,000 gallons in T-121 and 280,000 gallons in T-120 (Note _that 123,000 gallons is the' required volume for tank T-121 corresponding to a maximum achieved power of 80%.) Operability is demonstrated by verifying the contained water volume is within this limit once every twelve hour period. The operability of the condensate storage tanks 58
with the minimum water volume ensures that sufficient water is available to maintain the reactor coolant system at hot standby conditions for 24 hours with steam discharge to atmosphere concurrent with a total loss of offsite power. During the performance of the "B" series tests, the auxiliary feed-water system will be used to feed both steam generators. As a result, the contained water volume in tank T-121 could fall below the required 123,000 gallon minimum. The decay heat level, however, that was assumed when calculating this minimum volume is based upon a reactor trip following approximately two years of sustained power operation at 80% of rated thermal power. The actual decay heat level that will exist when Test 81 & B2 and when Test B3 are performed will be considerably less than this assumed value. Therefore, operation with tank T-121 volume below 121,000 gallons for short periods of time during the "B" series tests will not impact plant performance since the margin of safety as defined in the basis for the condensate storage tank technical specification will not be reduced. O e e G 59
l i Table 4-1 Technical Specification Impact Tech. Spec. Test Number Title R1 8 R2 83 3.4.1.2 Reactor Coolant Loops and Coolant X X Circulation - Hot Standby 3.4.1.3 Reactor Coolant Loops and Coolant X Circulation - Hot Shutdown 3. 7.1.'3 Condensate Storage Tank X X t e G 60
5.0 COMPLIANCE WITH ANALYSIS AND TESTING IN SUPPORT OF BTP RSB 5-1 5.1 Introduction The analysis and testing in support of Branch Technical Position (BTP) RSB 5-1 will be met by Southern California Edison (SCE) through the performance of essentially five discrete tests. Each of these tests have been assembled into a single program, the SCE Natural Circulation Test Program. Three of the five tests, i.e., the low power natural circulation tests or "A" series tests, were performed during the week of i September 1982. Section 2.2.1 above contains a brief summary of the results of the testing conducted during that week. The remaining two tests, the "B" series tests, will be performed in early 1983 as part of the Power Ascension Test Program. Sections 2.1.1 and 2.1.2 above contain brief descriptions of the "B" series tests and Sections 2.2.2 and 2.2.3 contain the detailed results of pretest computer simulations. Various data and information from each of the five tests of the Natural Circulation Test Program that have been obtained or that will be obtained and used in demonstrating compliance with BTP RSB 5-l'are as follows: A. Natural Circulation Demonstration (Test A1) - This test was used to demonstrate the basic natural circula-tion capability of the plant. Detailed analysis of test data has provided empirical values for mass flowrate vs core thermal output, reactor vessel AT vs core thermal output, loop resistances or k factors, etc. Section 2.2.1 above contains the results of certain portions of this data analysis. Also, as noted in that section, all appropriate analysis results available to data have been factored into the simulations in Sections 2.2.3 and 5.4 of this supplement to further enhance the accuracy of the computer predictions. l l 61
B. Demonstration of Natural Circulation at Reduced Pressures (Test A2) - This test was used to demonstrate that natural circulation could be maintained at reduced system pressures and that proper loop subcooling could be maintained through the use of the chemical and volume control system and pressurizer heaters. Detail-ed analysis of test data has provided information on ambient losses from the pressurizer and auxiliary spray efficiency. C. Demonstration of Natural Circulation with Reduced Heat Removal Capacity (Test A3) - This test was used to demonstrate that natural circulation could be main-tained with one steam generator isolated. Detailed analysis of test data has provided information on flow conditions in the reactor coolant loops, reactor vessel inlet plenum and downcomer regions, and the reactor ~ core. At the conclusion of the actual test and prior to reactor coolant pump restart, a concentrated slug of boron was added to the reactor coolant system and mix-ing was followed on the plant reactivity meter. The results of this boron addition experiment will be presented in detail in Section 5.3 belew and will be used as one argument to support the overall boron mixing process in the plant. D. ' Loss of Offsite Power (Test 81 & B2) - This test will be used to demonstrate that natural circulation can be established and the plant can be maintained in hot standby during an actual loss of offsite power. A baration and subsequent cooldown will not be performed; the test will clearly demonstrate, however, that the plant can be controlled in hot standby using safety-grade equipment concurrent with a loss of offsite power. Table 5-1 (p. 81) contains a list of systems and components that will be used during this test to show 62
compliance with BTP RSB 5-1. Also included in that table is the safety qualification of the system or component, the location from which it is operated during the test, whether or not it meets the functional requirements of the branch technical position, and an appropriate jusification for reliance on the system or compoment if it does not meet BTP RSB 5-1. E. Natural Circulation Test (Test 83) - This test will actually oerfonn a natural circulation boration, cool-down, and depressurization from hot standby to shutdown cooling system initiation conditions. The test will terminate when shutdown cooling system entry conditions are attained and will not initiate shutdown cooling. A demonstration of the performance of the shutdown cooling system within the context of BTP RSB 5-1 will be completed on San Onofre's Unit 3 at a convenient time prior to the first refueling outage on that unit. '- Table 5-2 (p. 83) contains a list of systems and components that will be used during this test to show compliance with BTP RSB 5-1. Also included in that table is the safety qualification of the system or i component, the location from which it is operated during the test, whether or not it meets the functional requirements of the branch technical position, and an appropriate jusification for reliance on the system or compoment if it does not meet BTP RSB 5-1. Three additional topics will be addressed in the remainder of this section. In 5.2 below the method of evaluating feedwater usage will be considered, in 5.3 the overall boron mixing process in the plant will be discussed, and in 5.4 a computer simulation of a strict BTP ~ RSB 5-1 scenario will be presented. Where appropriate, differences l between the actual conditions that will exist in the plant during test performance and the strict BTP RSB 5-1 scenario will be detailed and 1 properly justified, i l 63
5.2 Evaluation of Feedwater Usage Reference 5 states that "the seismic Category I' water supply for the ~ auxiliary feedwater system for a pressurized water reactor should have sufficient inventory to permit operation at hot standby conditions for at least four hours followed by cooldown to the conditions permitting operation of the residual heat removal system." As a further requirement, the " inventory needed for cooldown should be based on the longest cooldown time needed with either only onsite or only offsite power available with an assumed single failure." Implicit in this requirement is a decay heat level assuming a reactor trip following sustained power operations from 100". of rated thermal power. The San Onofre Unit 2 design provides a 150,000 gallon seismic Category I condensate storage tank, T-121, as the primary source of water for tne auxiliary feedwater system. While the capacity of T-121 alone is not sufficient to meet the storage requirements stated in Reference 5, the San Onofre Unit 2 design also provides a 500,000 gallon seismic Category !! condensate storage tank which has been enclosed in seismic Category I reinforced concrete retaining walls. Proper seismic Category I piping and connections exist between T-121 and the 500,000 gallon water source such that sufficient water would be retained and transferred for use by the auxiliary feedwater system to meet the requirements of Reference 4 in the event of a safe shutdown earthquake. (See Sections 5.4.3 and 9.2.4 of Reference 6 for a detailed discussion of the San Onofre condensate storage tanks.) As required in Section 5.4.3 of Reference 6 and as necessary to satisfy the testing and analysis in support of BTP RSB 5-1, detailed data on feedwater usage will be taken during Test B3. Main feedwater must be used to supply both steam generators pre-trip. Immediately following reactor trip, main feedwater will be manually secured and auxiliary feedwater will be used to supply both steam generators for the remainder of the test. All feedwater data will then be analyzed in order to demonstrate that suf ficient inventory exists to meet the above requirements. This analysis will be performed in essentially three steps. Fi,rst,' actual core thermal output or decay heat levels 64
during Test B3 will be determined by carefully measuring the steady-- state reactor vessel differential temperatures and comparing these values with the empirical results obtained during the Low Power Natural Circulation Tests, see Figure 2-1 (p. 25). Second,-the dif ferences will be quantified between the actual decay heat levels that exist during Test 83 and the decay heat levels that would exist in the plant following a trip from sustained operations at 100% of rated thermal power. Finally, feedwater usage determined from test data will then be extrapolated to demonstrate that sufficient storage capacity exists to meet the requirements of Reference 5, i.e., sufficient seismic Catagory I storage capacity exists to permit four hours of operation at hot standby followed by cooldown to conditions permitting use of the residual heat removal system concurrent with a loss of offsite power and an assumed single failure. 5.3 Roron Mixing Process a The basic design of the San Onofre Nuclear Generating Station has . included all required systems, procedures, and components necessary to reach and maintain hot standby conditions in the reactor coolant system under a variety of postulated accident conditions and scenarios. (Hot standby conditions are those in which the plant is maintained at normal operating temperatures and pressures with the reactor shutdown or subcritical.) With the issuance of Reference 5, it has become necessary to consider the various methods by which hot shutdown and subsequent long term cooling can be achieved. The achievement of long term cooling in the present context means the successful initiation of the residual heat removal system.
- Further, hot shutdown and subsequent long term cooling must be accomplished using only safety-related systems and components concurrent with a loss of offsite power and an assumed single failure. Three basic functions or processes must be successfully accomplished prior to the initiation of the residual heat removal system.
(For San Onofre this system is the shutdown cooling system.) The three basic processes are
- 1) baration, 2) cooldown, and 3) depressurization.
65
The first of these three basic processes, boration, should be divided into two discrete phases. The first phase is the delivery phase. During delivery concentrated borated water must be introduced into the reactor coolant system. For San Onofre this is accomplished by charging using the chemical and volume control system. (See Section 9.3.4 of Reference 3 for a detailed description of the San Onof re chemical and volume control system.) A discussion of the delivery - phase is presented in Sections 2.2.3 and 5.4 of this report. The second phase of boration is the mixing phase. During mixing added boron must be evenly distributed and dispersed througaout the reactor coolant system. This dispersion is essential if proper reactivity control is to be maintained when plant temperatures are reduced during Cooldown. The second of the three basic processes, cooldown, involves heat removal during the transition from normal operating temperatures and pressures to conditions that permit operation of the shutdown coolind system. Rejection of sensible heat and decay heat generated in the rhactor core is accomplished at the steam generators through the use of the atmospheric dump valves. Flow of coolant through the reactor core and in the reactor coolant system is maintained via natural circulation. The third of the three basic processes, depressurization, will involve the use of the pressurizer auxiliary spray systen since main spray is not available. For San Onofre a controlled depressurizatior. using auxiliary spray is accomplished by charging from the chemical and volume control systen with flow diverted away from the reactor coolant loops to the pressurizer spray nozzle. Detailed discussions of the cooldown and depressurization processes are presented in Sections 2.2.3 and 5.4 of this report. The discussion in the remainder of this section will concentrate on the mixing phase of the boration process. Refarence 5 requires adequate tests with supporting analyses to confirm that sufficient nixing of borated water added prior to or 66
during cooldown can be achieved under natural circulation conditions and to permit estimation of the times required to achieve such mixing. On a microscopic level the process of boron mixing, i.e., the even distribution and dispersion of added boron throughout the reactor coolant system is quite complicated. Several computer codes, e.g., LOFTRAN (Reference 7) and BITRAN (Reference 8), have been written that apparently have the capability to model some of-the microscopic details of the mixing process. Care must be exercised when using the results of these codes, however, since significant differences in boron distribution predictions have been obtained using LOFTRAN and BITRAN which are not readily explainable, the capability to directly analyze temperature stratification and boron stratification in LOFTRAN and BITRAN does not exist, and the mixing predictions produced by these codes have not been thoroughly benchmarked against actual plant
- data, if the problem of confirming adequate boron mixing and of estimating the times required to achieve such mixing is approached on a macroscopic level then the tests and experiments designed to meet these objectives and the subsequent analyses should be straight forward in nature. The basic procedure would involve adding a predetermined amount of borated water to the reactor coolant system under specified test conditions. Appropriate samples at key sample locations would then be taken at specified time intervals until adequate mixing was observed. By knowing the initial boron concentration and the amount of boron added at the beginning of the experiment, the expected final concentration can be determined. Using this number coupled with the data obtained from the boron sampling analyses, an estimation of mixing times can be calculated and the presence of stratification can be detected. A final complete solution would require showing that the conditions under which the test was conducted envelop any set of conditions which might be imposed or which could be postulated to occur in the plant.
l The postulated set of conditions under which adequate boron mixing must be demonstrated to occur are stated in the functional requirements of BTP RSB 5-1. These conditions require that only safety-grade systems and components be employed concurrent with a loss { 67
of offsite power and an assumed single failure. For San Onofre the assumed single failure is the failure of one emergency diesel generator to start. (See response to NRC Question 212.139 of-Reference 3.) The intention now is to consider tne boron mixing process on a macroscopic level. In order to properly do this, a hypothetical slug of borated water will be introduced into the reactor coolant system. The status of equipment and conditions assumed to exist in the plant are those that will exist when Test B3 is actually performed. The progress of the hypothetical slug of boron will be followed as it transits the reactor coolant system. Component-configuration and features conducive to mixing will be elaborated upon; actual experimental and test data will be presented in support of the overall mixing process as appropriate. To begin the discussion of the mixing process, the hypothetical slug of concentrated borated water must first be introduced into the reactor coolant system. This is accomplished by charging via the chemical and volume control system. (See Section 9.3.4 of Reference 3 - for a detailed description of this system.) Boron from a highly concentrated borated water source, the boric acid makeup tanks, is injected directly into tho reactor coolant system using positive displacement type charging pumps at either one or both of two locations as shown in Figure 5-1 (p. 90). Figure 5-2 (p. 91) shows a cross section of a cold leg pipe through one of the charging inlet nozzles. Note that charging flow penetrates the cold leg pipe 90' from its top. Two features of this mode of boron delivery are important to the mixing process. The first feature of importance is the amount of borated water that must be added to obtain the required shutdown margin for a plant cooldown and the time frame over which this boron is added. The second feature of impnrtance is the velocity of the charging fluid at the inlet nozzle as it penetrates into the reactor coolant. A third feature, the temperature of the charging fluid, will also be considered. As will be shown, however, because of this mode of boron delivery and the reactor coolant system configura-tion the effects of charging fluid temperature on the mixing process are negligible. 68
Section 3.1.2.7 of Reference 4 requires a minimum boron concentration in the boric acid makeup tanks of 7.5 wt. % (weight percent). This concentration corresponds to 13,125 ppm (parts per million) boron. At-end of core life with the most reactive control rod fully withdrawn, 3367 gallons of a 7.5 wt. % solution of boric acid are required to be added to the reactor coolant system to ensure the reactor remains subcritical throughout a plant cooldown. End of life conditions are specified since the nagnitude of the moderator temperature coefficient is larger than at the beginning of core life and therefore more boron must be added. Charging using two charging pumps under these condi-tions would require approximately 38 minutes to add the necessary amount of solution. At all other times in core life with conditions less severe, i.e., all rods fully inserted and a moderator temperature coefficient of less magnitude, a smaller amount of boron and therefore a correspondingly shorter charging time will be required to obtain the proper shutdown margin. During Test B3, for example, approximately ten minutes of charging time will be needed to properly borate. Although the amount 0f boron that must be introduced into the reactor coolant system will vary considerably over core life, the rate of addition will be essentially constant since positive displacement type charging pumps are utilized. The delivery phase of the mixing process will be smooth and continuous with all required boron added in forty minutes or less depending on plant conditions. Given the constant rate of addition, the velocity at the charging nozzles can be readily determined. Table 5-3 (p. 87) shows steady-state fluid velocities at the charging nozzles as a function of the number of charging pumps with letdown in operation and with letdown secured. Note that the effect of letdown is to increase both fluid velocities and temperatures at the charging nozzle, i.e., bnth the fluid velocity and temperature with letdown in operation using a given number of charging pumps is greater than the corresponding situation without letdown. For comparison with the numbers in Table 5-3, the velocity of the coolant in the cold legs as determined following the Low Power Natural Circulation Tests was 1.35 ft/sec. Core thermal output was 0.89% of rated thermal power for this determination, and 69
l steady-state symmetric natural circulation flow conditions existed. Considering the limiting charging situation, i.e., one charging pump with no letdown and flow to both inlet nozzles, the velocity of the a charging fluid will be 2 -1/2 times greater than the velocity of the coolant in the cold legs. This difference in. velocity will ensure that the charging fluid, and hence boron, will penetrate substantially into the reactor coolant and that conditions whereby complete mixing can begin to take place will be established. Again, as shown in Table 5-3, letdown has two important effects on charging fluid. First, the use of letdown increases the velocity at the charging inlet nozzle, and second, the use of letdown increases charging fluid temperatures. During Test 83 letdown will not be isolated and hence charging velocities and temperatures will be relatively high. (Note that letdown will remain in operation during this test in order to prevent the thermal stresses that would be a placed on the charging nozzles if it were secured.) If letdown were isolated and hence charging velocities and temperature were relatively r low the probability of influencing the mixing process through boron and temperature stratification would be increase. Because of the mode 4 of boron delivery and the configuration of the plant, however, stratification will not occur. Refering to Figure 5-2 again, the concentrated boron solution will penetrate 90 from the top of the cold leg pipe. Charging velocity at the inlet nozzle is high relative j to the velocity of the coolant in the cold leg during natural circulation such that the boron solution will penetrate substantially into the reactor coolant. In. addition, Reynolds number calculations - clearly indicate that flow through the cold leg will be turbulent. These two factors, penetration of charging fluid into the reactor coolant system and turbulent flow in the cold leg, will create the mechanism for substantial mixing of added baron as it flows from the charging inlet nozzle down the cold leg toward the reactor vessel inlet plenum, and prevent establishing the conditions necessary for boron and temperature stratification. e 70 i.
e Reference 9 contains experimental data which can be used to support the conclusion that substantial mixing of an injected baron solution will in fact take place as it is carried toward the reactor vessel inlet plenum by the fluid in the reactor coolant system cold legs. The tests in Reference 9 were designed to verify the mixing models used by Combustion Engineering in evaluating pressurized thermal shock events and, as will be shown, certain of the results are applicable to the baron mixing process. Figure 5-3 (p. 92) shows the data from one of these experiments Test 42. The work was funded by the Electric Power Research Institute and performed by Creare, Inc., on a 1/5 scale mixing facility. Each of the points in Figure 5-3 are thermocouples with steady-state temperatures as indicated. The test was designed to simulate the initiation of safety injection flow during natural circulation and the results show that mixing was approximately 90", complete at the end of the cold leg. The actual test conditions employed and the plant conditions to which they correspond are given in Table 5-4 (p. 88). Although, as stated in Reference 9, data from a larger test facility are' not yet available to confirm the scaling rationale used to obtain the corresponding plant conditions in Table 5-4, several conclusions relevant to the boron mixing process can be drawn from the results in Figure 5-3. Using the highest nozzle velocity, for conservatism, in the absence of letdown from Table 5-3 (13.6 ft/sec) and actual results from the Low Power Natural Circulation Tests, a table similar to Table 5-4 can be compiled for the situation in which charging is used to add fluid to the reactor coolant system vice safety injection. This information is presented in Table 5-5 (p. 89). Since the maximum j charging flowrate (WCHr,) is lower than the safety injection flowrate (W3g) and since the ratio of loop flowrate to charging flowrate is greater than the ratio of lonp flowrate to safety injection flowrate, it can be concluded that mixing the results in Figure 5-3 for safety injection envelop the situation where colder borated water is introduced into the cold leg piping via the charging system. This conclusion is further supported by the configuration of Cold Legs 1A and 2A, i.e., the relative position of the charging inlet nozzles to l 71
the safety injection nozzles. As seen from Figure 5-4 (p. 93) the charging inlet nozzles are actually located further from the reactor vessel inlet plenum than the safety injection nozzles thus allowing for increased interaction with the coolant in the cold leg and hence even more complete mixing of the charging fluid. At this point in the discussion of the boron mixing process, the hypothetical slug of borated water that was added to the reactor coolant system at a loop charging nozzle has traveled down the cold leg to the reactor vessel inlet plenum. As was shown, thermal and hence boron mixing at tne inlet plenum will be greater than 90%
- complete, i.e., the added boron will be 90% mixed with the coolant in the cold leg at the inlet to the reactor vessel. This mixing will preve nt the establishment of a definite colder temperature layer and hence prevent stratification in the cold legs. As the fluid from the cold leg enters the inlet plenum it will gain a counter clockwise rotation as it flows into the downcomer region of the reactor vessel.
This rotating motion results from the cold leg piping configuration as. shown in Figure 5-5 (p. 94) a'd was actually observed during pre-core n testing of the safety injection system by plant personnel standing above the uncovered reactor vessel. Fluid rotation in the downcomer thus results in the mixing of the added boron with coolant entering the vessel from all four cold legs. As the coolant leaves the downcomer region it flows past a flow skirt, as shown in Figure 5-6 (p. 95), and into the bottom of the reactor vessel. From there it flows up past the lower support structure and into the core. From the core, flow proceeds up to the upper guide structure and out into the hot legs. As can be seen from the side view in Figure 5-6, two regions of potential stratification or non-mixing of boron exist. These two regions are the upper head and the vessel bottom. In the reactor vessel bottom region, there are two factors that will act to prevent stratification. First, coolant leaving the downcomer region will be substantially mixed due to mixing in the cold legs and due to downcomer fluid rotation. This mixing, both thermal and boron, means 72
that dense highly concentrated. pockets or slugs of borated water will be eliminated prior to reaching the bottom of the reactor vessel. Second, the reactor vessel bottom is smooth and completely free of any Instrument penetrations thus allowing complete and unhindered interaction of all fluid entering the region. In the reactor vessel upper head, however, temperature stratification and delayed mixing of boron will be present since a relatively rapid mechanism for introducing boron into that region does not exist. At present, it appears as if the only viable mechanism for getting boron into the upper head is diffusion. Although diffusion is a relatively slow process and therefore the boron concentration in the upper head will very probably be less than in the actively flowing portions of the reactor coolant system during cooldown, this will not present a reactivity control problem since enough boron will always be added to the reactor coolant system to properly borate its entire volume. Therefore as diffusion or other possible mechanisms for mixing occur and the stagnant upper head volume becomes borated, a proper shutdown margin will always be maintained. As stated above, boron entering the reactor vessel will substantially mix with the coolant from all four cold legs due to downcomer fluid rotation. As a result, complete or almost complete mixing of added boron should be accomplished by the time the reactor coolant has traveled past the flow skirt and up through the core to the outlet plenum. Experimental data from Test A3 can be used to support the idea of complete mixing of all fluid from the cold legs by the time flow leaves the core. (See Section 2.0 of Reference 1 for a det.siled l description of Test A3.) Figure 5-7 (p. 96) shows this test data. l Specifically, Figure 5 7 contains the core exit thermocouple readings l and the loop resistance temperature detector readings during the asymmetric natural circulation test. Note that the fluid entering from Cold Leg 2 is approximately 21*F warmer than the fluid entering from Cold Leg 1 and that complete thermal mixing occured in the vessel since only random deviations in core exit thermocouple readings from the average core exit temperature of 548.4*F were observed. Con-sequently, downcomer fluid rotation and interaction in the bottom of 73
w-the reactor vessel appear to be viable and substantial mechanisms to ensure boron mixing in the reactor coolant system during natural circulation. One final area exists in the reactor coolant system that will i contribute to the overall boron mixing process. As the hypothetical ] slug leaves the core completely mixed with the coolant that entered the reactor vessel from all four cold legs, it will split with half flowing down Cold Leg 1 toward Steam Generator 1 and half flowing down Cold Leg 2 toward Steam Generator 2. Once in the steam generator, the boron slug will pass up through the tube sheet and into the U-tubes. As shown in Figure 5-8 (p. 97) these U-tubes do not have identical lengths. Instead, tubes at the center of the bundle are much shorter than tubes on the outside of the bundle. Since all U-tubes have the same diameter, the velocity of the fluid in each tube will be equal. The final result will therefore be that an increment of coolant entering one of.the shorter U-tubes will exit that tube and pass into the cold legs sooner than an increment of coolant entering one of the longer tubes. Hence a mechanism exists for elongating or spreading out of added boron in the actively flowing regions of the reactor coolant system. Once the elongated hypothetical slug of boron passes out the steam generator U-tubes, its first transit around the primary system will be complete as it again splits, see Figure 5-1, flowing into the two reactor coolant pump suction legs and eventually back into the cold legs. In summary, the process of boron mixing was discussed by considering a ~ hypothetical slug of boron added via the charging system and following the slug as it made one transit around the reactor coolant system. By considering component configurations and using actual experimental data where appropriate, three significant mechanism were found to be present that would ensure complete mixing of added boron and prevent stratification in all actively flowing pnrtions of the reactor coolant system, i.e., all regions except the reactor vessel upper head. These mechanism were the substantial mixing of added boron with fluid in the cold leg, fluid rotation in the reactor vessel downcommer, and 74 l l l
I elongation by the steam generator U-tubes. In addition, it was shown o. that the presence or absence of letdown would have a negligable effect on boron mixing and that delayed mixing in the reactor vessel upper head would not effect reactivity control. One final piece of experimental data exists and will be presented in support of the conclusion that complete mixing of all added boron will take place in the actively flowing regions of the reactor coolant system. At the conclusion of the Low Power Natural Circulation Tests and prior to reactor trip, a concentrated ten gallon slug of boron was actually added to the plant and its progress as it transited the reactor coolant system was followed on the reactivity computer. Conditions at the time of the test were as follows: 1. Reactor critical at approximately 1.2% of rated thermal power. 2. All four reactor coolant pumps secured. 3. Reactor coolant system pressure approximately 2250 psia. 4 Cold leg temperature approximately 515*F. 5. Hot leg temperature approximately 548'F. 6. Steam generator pressure approximately 771 psia. 7. Group 6 control element assemblies > 60 inches withdrawn. All other control element assemblies full out. Figure 5-9 (p. 48) shows the behavior of the plant reactivity computer following the addition of the ten gallon slug of borated water. For reference, Figure 5-10 (p. 99) shows the location of the detectors t used by this computer. Note that the signal from both detectors is l averaged and that this average signal is then used to plot reactivity. l 75 l I l
Several importantEpieces of:information can-be obtained by 1ooking at the general reactivity trends in Figure 5-9. First, with each pass of-the boron' slug through the core a negative reactivity spike resulted; from these spikes a loop transit. time of 4.8 minutes could be directly measured. Second, with each pass through'the core the slug of boron became more and more diluted, i.e., the slug'of boron became more and i more elongated and mixed.. This conclusion is' reached by' noting that-each successive negative reactivity spike is smaller and broader than the one before. Finally, the boron' slug was almost completely mixed by the time it made its fourth. pass through the core approximately 17 minutes after it was first added. i-5.4. Computer Simulation of a Strict BTP RSB 5-1 Scenario Section 2.2.3 of this supplement contains the expanded computer simulation of Test B3 as requested by the NRC Staff. The results presented in that section simulate the actual test procedure only'and as such represent an aid to assist or guide the operators during actual test performance. For comparison with the results 16 Section 2.2.3, the results of a computer simulation of the strict BTP RSB 5-1 scenario are presented below. Note that' a complete simulation from hot standby to cold shutdown is not performed, i.e., a 'boration is performed followed by a cooldown to 350'F only. The reason for presenting a partial simulation only of the strict BTP RSB 5-1 scenario is the uncertainties associated with the reactor vessel upper head cooldown rates. Only after Test R3 is actually performed and the 4 upper head cooldown predictions of Figure 2-14 (p. 38) are empirically j verified can a complete simulation of a strict BTP RSB 5-1 scenario be l performed. Even though only a partial cooldown to cold shutdown is performed below, the results are meaningful when compared to the results in Section 2.2.3 since they show that piant response to Test i B3 and plant response to the BTP RSB 5-1 scenario are very similar. Figures 5-11 through 5-26 (pp. 100 to 115) contain the results of the partial simulation of a strict BTP RSB 5-1 scenario. The simulation was performed using only safety-grade equipment concurrent with a loss 76
of offsite power and an assumed single failure. (Note that the IE pressurizer heaters, although not safety-grade, were used.) For San Onofre the assumed single failure is the failure of one emergency diesel generator to start. (See response to NRC Question 212.139 of Reference 3.) The initial assumptions concerning system responses and operator actions used for this computer run are as follows: (Note that additional assumptions concerning operator actions during boration and cooldown will be detailed in the discussion below.) 1. The plant is initially stable at 100% of rated thermal power. Post-trip decay heat values are obtained by multiplying the standard decay heat curve by the initial power level in percent. (See Figure 5-11.) 2. Initial steam generator pressure is 900 psia. Following the loss of offsite power, both atmospheric dump valves quick open when steam generator pressure reaches 1000 psia. 3. Emergency feedwater flow at 400 gpm per steam generator and 70'F is initiated when steam generator narrow range level reaches 23*.. 4 Main feedwater flow is lost immediately upon loss of offsite power. 5. Steam generator levels are maintained via manual "0N/0FF" type control of the auxiliary feedwater system. (See Figures 5-25 and 5-26.) 6. Letdown is lost immediately upon loss of offsite power, 7. Pressure control is maintained by manual operation of the IE pressurizer heaters. Note that with the failure of one emergency diesel generator only 200 kW of heater capacity is available for use. 77
8. A pressurizer ambient heat loss of 130 KW is assumed throughout the simulation. ~ 9. The operable emergency diesel generator is assumed to - be available five minutes following the loss of offsite power.
- 10. The transient is initiated at time zdro by tripping offsite power.
- 11. A turbine trip occurs 0.5 seconds after the reactor trip.
Following the initial trip, reactor coolant system mass flow decreases rapidly as reactor coolant pump speed decreases. Pump speed with reach zero at about 250 seconds after trip, and full natural circulation flow, see Figure 5-12, will be established approximately ten minutes into the transient. Once full natural circulation flow is established, the reactor coolant system mass flow will steadily decrease as the core thermal output or decay heat level steadily decreases. The small perturbations to the mass flow that are seen in Figure 5-12 during the boration and cooldown portions of the simulation are the result of the transient effects of increases in charging flow. System pressure, initially 2250 psia, will decrease rapidly to about 2000 psia immediately following the initial trip as pressurizer level decreases in response to reac. tor coolant contraction. As natural circulation is established, the average coolant temperature will increase as hot leg temperature increases and reactor coolant expansion will occur. The resulting rise in pressurizer level coupled with the manual operation of the pressurizer heaters once they become available act to increase system pressure. The available pressurizer heaters are secured when system pressure reaches 2275 psia and charging is secured when pressure reaches 2300 psia. Pressurizer level decreases rapidly following reactor trip in response to reactor coolant contraction. Level recovery occurs as natural circulation is established and reactor coolant expansion takes 78
e place.. Mass is.also added to the reactor cool' ant system during this period by automatic action of the charging system, see Figure 5-16. Note that for five minutes following the loss of offsite power, until the operable emergency ' diesel generator is running, no charging will ~ be available. It is also' assumed that the swing charging pump is i aligned to the failed diesel generator and will not be available for .i use until one hour into.the transient. Once natural circulation is established an'd plant conditions have stabilized, a simultaneous boration and natural' circulation cooldown will'be commenced. The boration and cooldown must proceed concur-t rently in the BTP RSR 5-1 scenario in order to_ maintain proper loop subcooling (Figure 5-20) with the limited available heater capacity. Boration is performed using two positive displacement type charging pumps each with a capacity of A4 gpm and each taking suction on a 4 concentrated borated water source, the boric acid makeup tanks, via boric acid makeup pumps or a gravity drain line. This solution is then introduced directly into the reactor coolant system through the l charging nozzies. The boron concentration in the pressuri*zer is
- increased as necessary by switching the output of the charging pumps from the reactor coolant system loops to the pressurizer via the auxiliary spray line with simultaneous operation of the available pressurizer heaters to limit depressurization. A simulation of boron mixing in the pressurizer in this manner for the computer run in this 4
section was not performed. The si.3ulation was performed in Section 2.2.1, however, and the results can be seen in Figure 2-5 (p. 29). As j an alternative, boration of the pressurizer can be delayed until plant i depressurization after the cooldown is complete. At that time, a con-I centrated boric acid solution can be introduced into the pressurizer during spray down. Concurrent with boration, a cooldown is initiated at the maximum administrative cooldown rate of 75'F per hour. i l (Reactor coolart system loop temperatures are shown in Figures 5-17 L .~ and 5-18.) Cooling the plant at the highest rate possible, i.e., the l maximum administrative cooldown rate, will create the largest temperature differential between the reactor vessel upper head and the remainder of the reactor coolant system and thus maximize upper head 79 a n.-- --.~,.
cooling. The upper head cooldown rate predicted for this. simulation is shown in Figure 5-19. As noted above, however, this prediction must be enpirically verified before a meaningful RTP RSB 5-1 simulation can be performed. Boration is conducted for forty minutes using two charging pumps as shown in Figure 5-16. Charging with two pumps for forty minutes will ensure proper boration at the most limiting time in core life with the most reactive control rod stuck in the full-out position as discussed in Section 5.3 of this supplement. Once boration is complete, one charging pump will be secured and the cooldown will be continued. Since coolant contraction due to the cooldown will be greater than the makeup capacity of one pump, pressurizer level will begin to decrease as seen in Figure 5-15. Once normal level is regained, the second charging pump will be cycled as necessary to maintain level constant. The resulting effect on system pressure is shown in Figure 5-13. Note that given the limited available heater capacity of 200 kW, pressure did not decrease below approximately 2000 psia in this scenario and ample loop subcooling was always maintained (Figure 5-20). Once cold leg temperatures were reduced to 350*F, the simulation was terminated. Again, as stated above, a complete simulation of the BTP f 4 RSB 5-1 scenario to cold shutdown will not provide useful information until the upper head heat loss predictions for a natural circulation cooldown can be empirically verified. i l 80
e e e i Table 5-l Systems and Components Rolled upon in Test B1 & B2 to Shou Compliance with BTP R$8 5-1 Systes/ Component Safety-Grade Location Operated Meets BIP R W 5-1 Justification / Discussion (Ves/No) (Yes/No) Ausillary feedwater Yes Inside control room Yes Tlie M5 is designed to be operable from the system (MS) control room; this type of operetten was d====- Outside control room strated during the 'A' series tests. During this at the avulliary test. manual control of the bypasses around the feeduater station AFS regulating valves at the ausillary feedseter station may be assumed if fleurates became lem. 3 l It pressurizer No Inside control room No The IE pressurizer heaters will be used to ensure heaters proper subcoollag is malatained. Pressurizer pressure Yes N/A Yes Indication in control room. Indication y Pressurlier level Yes N/A Yes indication la cortrel roca. ~ indication i' .: s f r ~ Cold leg temperature Yes N/A Y6* Indication in control room. t ] Indication r / 1 Not leg temperature Yes N/A Yes Indicatten in control room. indication 4 Steam generator level Yes N/A Yes Indication in control room. Indicatics i S lf w--w---ew we v wm- -w- --n r w ---we =r- --ww-- -w-w ww =
Table 51 (cant.) System / Component Safety-Grade Location Operated fleets BIP RSS 5-1 Justification / Discussion (Yes/No) (Yes/No) s l Steam generator pressure Yes N/A Yes Indication la control room. Indication l 1 Letdown portion of No inside control room les Letdown will not be secured during this test in the chemical and order to prevent the high thermal stress that volume control would be placed upon the loop charging nozzles, system (CVCS) Atmospheric dump valves Yes inside control roon Yes Local handuheel operation of the ADVs will be (ADVs) demonstrated during this test, to Outside control room via local handuheels Nitrogen backup to ADVs Yes N/A Yes Charging portion of Yes Inside control room Yes the chemical and volume control system (CVCS) Ausillary spray system Yes inside control roon Yes As stated in Section 2.2.1 (p. 9), the main spray valves mast be closed in order to prevent circulation of availlary spray flow back to the reactor coolant system loops and times prevent depresserlaation. ]
e e r Tahle %-2 Systems and Components Reited upon in Test 81 to Show Compliance with BTP R$8 5-1 System / Component Safety-Grade Location Operated Meets BTP R58 5-1 Justification /Olscussion (Yes/No) (Yes/No) Film feedwater system No inside control roon No The ifs must be used to supply both steam (MFS) generators pre-trip. Following reactor trip, the feedwater control system will ramp down main feedwater flow to 5% of full flow at which time flow will be manually secured. To enhance plant safety, main feeduater should not be secured until proper operation of the availl'ary feedwater System s is vertfled. All additions of water to the steam o generators by the MFS during this period will be 4 carefullyaccountedforinthepost-trikfeedmater analysis descrlhed in Section 5.2 (p. 64), m W t Steam lippass control ' No Inside control room g Nd ', The steam bypass control systm will be.used post-j system,.(5BC$) t rie: to dump steam during the boren mining portion of the test. The use of ths,,$aCS during this-portion of the test isill alfine for water recovery I I L ,and therefo(e help.el}tle.lte' costs and restoration = g e . time. Operation of the' atmospheric &cp valves N, will be desenstrated dje r14 the coolda.c and.I y s s depressurization portfor:s of ;this test 'as weil as 'A \\ deelnc Test Bl's R2. i ! g i s y. ) G' l ,3 A e,, j s o. Auxftlary feedwater Yes i.. 'Isaside control room Yes The AFS is designed to be operable f rom the;-. r \\ system (AFS) control rcswl'tals type of operation was demon-s Ottside cor. trol room strated durlig the 'A series tests :Durles (bis s e .? altheauxiliary test manual con.ts01'of the bypasses arwne the-C- 1 fec-dnet er station a b Af S segulating valves at the auxillary feed, water > . /. : + I i \\ Y , station w y be asstated if flowrates'become low. - r> \\. ./, a '\\ <,'t. f, / *! ,9 lf ~ <\\ Q ' /]. \\ u 3-p +- m ,. 4 i<,f z ~ q ,~ N -\\ m 'g L 1 ' ' ' ~ ' t 1,"- _ ~,.,
, -c.
- fS.
' ~ s e e e v e J. / / p ~ n p T4hle L 7 (cont.) .d ^ ~ s j I l } y System / Component Safety-Grade Location Operated Meets BIP.pS8 5-1 i ' Justificatlost/Dik ussion j ~ (Yes/No) (Yes/Nu) 5-s 7 I men-iE pressurlier No inside control room No I 'sAll pressurlier-heaters will t;e used for.the heaters
- initial transient in this test'to allow' fer fu!!,
s autcaatic plant response.- j , /. p l IE pressurl:er No inside control room ho ' t' The IE g.ressortaer hekters nel" will be used for .. j y the boration, cooldc,wn, and dcpre:sureretion portions of this test' to ensurt! proper subcooling is maintained. Pressurlier pressure Yes N/A Yes Indication in control room. Indication g b Pressurizer level yes N/A Yes Indication in control room. till cold leg temperature Yes N/A n n on r I r m. Indication Steae generator level yes N/A indication Steam generator pressure yes N/A '"I'"* Indication e
Table 4-7 (cont.) System / Component Safety-Grade Location Operated Meets BIP RS8 5-1 Justification / Discussion (Yes/No) (Yes/No) Letdown portion of No inside control room No Letdown will be operable during this test to the chemical and prevent the high thermal stress that would be volume control placed upon the loop charging no ules if it were system (CVCS) secured. As stated in Section 5.3 (p. 65), the presence or absence of letdown will have a negligible effect on the boron mixing process. Sample system and No Outside control roon No Primary fluid samples must be taken at the boronometer locations and with frequencies stated in Table 2-1 (p. 21) in order to verify proper boron mining. Atmospheric dump valves Yes inside control room Yes Local bandwheel operation of the Anys will be (ADVs) demonstrated during this test. Outside control room via local handuheels Nitrogen backup to ADVs Yes N/A Yes Boration portion of Yes Inside control roon Yes the chealcal and volume control system (CVCS) Ausillary spray system Yes Inside control room Ves As stated in Section 2.2.1 (p. 9), the main spray valves must be closed in order to prevent circulation of auxillary spray flow bacit to the reactor coolant system lonps and thus prevent depressurization.
Table 5-2 (cont. ) l System /rm ment Safety-Grade Location Operated Meets BIP RSS 5-1 Justification / Discussion (Yes/No) (Yes/No) Reactor coolant system No inside control room No The RCS auto makeup system will be used to adjust (RCS) auto makeup system the concentration of the boric acid solution used to maintain proper RCS inventory during the coel-down and depressurlaation portions of the test in order to prevent the addition of unnecessarily large amounts of boron. cn Control element drive Po outside control roon No Although not directly relled upon to show cn mechanism (CEDM) compliance with SIP RSS 5-1 the CfDN cooling fan cooling fan will be in operation during this test to prevent equipment damage. As discussed in Section 2.2.3 (p. 14) an appropriate allowance will be made when verifing the head heat loss predictions. d
Table 5-3 Effect of Temperature on Charging Fluid Velocity Number of Temperature Velocity at charging nozzle (ft/sec)* pumps with at outlet to capacity regenerative at pump heat exchanger Single Both 1 (44 gpm)** 90 F 6.8 3.4 2 (88 gpm)** 90*F 13.6 6.8 1 (44 gpm)# 445*F 8.1 4.0 2 (88 gpm)# 405*F 15.7 7.9 3 (132 gpm)# 380*F 23.2 11.6 Velocity is given for all flow diverted through one nozzle (single) and flow split between two nozzles (both). No letdown. f Temperatures and velocities based upon steady-state test results with letdown. (Attachment C, CVCS Integrated Test, 2HB-220-01) 87
l- [L Table 5-4 Comparison of Test Conditions curing Creare Test 42 with Corresponding Plant Conditions 1 i Plant Conditions During Corresponding Plant P'a ramete r Test d? Conditions 2.7 450 WLOOP (lbm/sec) ~ W 0.3 50 3t (Ibm /sec) J 9 9 LOOP /WSI W T 155 400 LOOP (*F) T; 60 40 3 ('F) e 9 h 88
Table 5-5 Actual-Plant Conditions During a Natural Circulation Boration Without Letdown Plant Parameter Actual Plant Conditions 307 WLOOP (lbm/sec) W 31 CHG (Ibm /sec)' i0 LOOP /W HG W T 545 LOOP (*F) 90 TCHG (*F) 4 a 9 e t 89
FIGURE 5-1 REACTOR COOLANT SYSTEM 4 1 CHARGING INLET I s PUMP (0 No.1B
- o. 2A a.,
W. / ' I' l 'g," s N . m,~g', {_./ 'X STEAM REACTOR h ATO / 0 :)._. Qj NE,RATOR P= VESSEL l No.1 = i f )4_w.C... / \\ L. .2 ~,_ j b4';1 ~ r* n - -,- - t g ,s ' ') l 1 ~ O C') N I CilARGING INLET i PUMP No. 2B PUMP No. lA kg PRESSURIZER i iU i
FIGURE 5-2 CHARGING INLET N0ZZLE CONFIGURATION e e 15" CHARGING FLOW >-2.6" I.D. = 4 ~ COLD LEG PIPING CROSS j SECTION. SEE FIGURE 5-1. i 91 l
FIGURE 5-3 Fturn TEMPERATURE DISTRIBUTION FOR CREARE MIXING TEST 42 I ~ e h Il00% MIXED = lt45.8 F + p 5.62" \\ o 10 If4 14I)+5 ' 14I) hp 4 a
- 146q, tp143 46143 160 146di q,144 120 22.4" 145El 1
80 T I 1 t I 160 J '- 1464>i'146 ~ 120 .2 " A 80 1' ~ T I I t 1 0 20 40 60 80 TIME (sec) i e 92
s I FIGURE 5 11 REACTOR COOLANT SYSTEM COLD LEG CONFIGURATION l SAFETY INJECTION CHARGING INLET N0ZZLE / N0ZZLE J SIDE VIEW g.3, FROM REACTOR COOLANT PUMP g l< 158" -
- l CHARGING INLET yyn TOP VIEW FROM REACTOR REACTOR VESSEL 00LANT PUMP INLET PLENUM
/ SAFETY INJECTION N0ZZLE
FIGURE 5-5 TOP VIEW OF REACTOR COOLANT SYSTEM SHOWING DOWNCOMER FLUID ROTATION DUE TO COLD LEG ANGLE OF PENETRATION e x_ / '$1 1 / -s ~ ( l I 'w..... y u y'? ,'f -,s88 ,p -, I ' I'I ( ', esEE$Es mas'
- esEEE am a i
/' N me-,-4/ /
- H-
- d. O 94
FIGURE 5-6 REACTOR VESSEL ARRANGEMENT . IN-CORE INSTRUMENTATION SUPPORT PLATE - CEDM N0ZZLE fUI TU h .-l; ZZL Y. L [Ih CONTROL I i t %d 1 ELEMENT fLLL N 5 J { j WITHDRAWN STRUCTURE g. ) r c r c f I T i \\ ~~ 30 " l 0 4" 10 'N r T 7 i INLET ~ OUTLET \\ I I N0ZZLE ,s1 tlLill. N0ZZLE I rd I lii CORE lit j ii, SUPPORT SURVEILLANCE BARREL HOLDER 150" Ii CORE age % SHROUD i f I h } ' N i FUEL ASSEMBLY ji,, i,, ug ' - gm LOWER b SUPPORT SNUBBER \\Y,k I' STRUCTURE
- c' *:':: ::: * : !. !j; FLOW CORE STOP SKIRT O
95
FIGURE 5-7 TOP VIEW OF CORE SHOWING CORE Ex1T THERMOCOUPLE LOCATIONS AND IEMPERATURE DIFFERENCES FROM AVERAGE COLD LEG COLD LEG 2B I^ 536.5 *F SM.7 4 s to :: i4 I I l s s4.,~sjjjiTriA;7 i,a isszo zi z 4 /
- l~i I ! ! I i l'hs!'lN p
'i q i liil ,i, i N lo. .. 3l 'i.. ' 2., N w L:.: ..3 j,f ; / /l HOT LEG
- / / l Ed d'
HOT LEG 2 l l l s I i ", 4 j_ _L i u,J g,, l a+- 1 1 1l @lI I q-- --. 3 a.. , ta b l 1 $l ...g .. 3 l May-i.. i... ..si i II !I .. r 547.9 *F e\\ \\ 547.6 *F -i.J si.a n..I i \\\\l l l l l l l l l 0 c .,i ...I I,.. ! i..! ->. 4 \\\\\\\\l l l l l l l l e l l L.I a i, COLD LEG COLD LEG 2A 1B 536.8 *F 51 5.1 F CORE EXIT THERMOCOUPLE AVERAGE TEMPERATURE = 548.4 *F i l 1 96 1 1
FIGURE 5-8 RELATIVE LENGHT OF STEAM GENERATOR U-TUBES i e I : Z 2
- j/' ~ ~ ~ ~ ~; 5
/ .~~~f K L l + d mG il i ++l};[ i i %= %l*8 \\ /-W lla. .Y*,j. X f '\\ t /> < [~Q f l: !f I l l 1 l l 708 TOTAL LENGTH l i. l l 45' TOTAL LENGTH h i i i i Il t e-r i k e A f
- x. 4>
e e 6 97
FIGURE 5-9 REACTIVITY FOLLOWING ADDITION OF IEN 6ALLON BORON SLUG TO ONE LOOP OF REACTOR COOLANT SYSTEM i TEN GALLONS CONCENTRATED BORON SOLUTION ADDED TO SUCTION OF CHARGING PUMPS AT TIME ZERO. +5 m THIRD ' SECOND PASS ua FIRST PASS 8 PASS g 0 5 FOURTH { PASS 5 -5 4.8 MINUTES I l i i i l i 1 0 2 4 6 8 10 12 14 16 18 IIME (MINUTES)
s FIGURE 5-10 DETECTOR LOCATIONS WITH INPUT TO THE REACTIVITY COMPUTER REACTIVITY RECORDER COMPUTER n DETECTOR I PUMP pump l ** IB r w ~. /('?^ /../'^@8' SnAM REACTOR STEAM CENERATOR ( t. O if... Ql = lil: VESSEL l ll --. I 1 RATOR No.1 I f ;. / s' O. \\'[(,rMh. -- },7,$,O s' l --[,') (',').- DETECTOR PUMP No. 28 PUMP No. IA PRESSURIZER i 2
F IGURE-5-11 SIMULATION OF A BTP RSB.S-1-SCENARIO DECAY-HEAT. '10 9 8 e' 7 w 2 O Q. 5 6 5 C 5 =w Ua: E 4 CORE THEPJiAL OUTPUT BASED UPON THE STANDARD DECAY HEAT 3 CURVE IN PERCENT OF INITIAL POWER. 2 i 1 0 1 I i 1 0 2500 5000 7500 10000 12500 15C00 TIME (SECONDS) 100
F IGURE 5-12 SIMULATION OF A BTP RSB 5-1 SCENARIO 10 m 9.. 8 -. 7 3 6 N tu (1. g 5 a u. FULL NATRUAL CIRCULATION FLOW ESTABLISHED 4 4 3 ~ 2 1 0 I I i t i 0 2500 5000 7500 10000 12500 15000 TIME (SECONOS) 101
F I GUR E 5-13 SIMULATION OF A BTP RSB 5-1 SCENARIO ER N SSURE 2500 2400 2300 _ CO MENCE SIMULTANEOUS PLANT BORATION AND / C00LDOWN USING TWO CHARGING PUMPS. 2200 BORATION COMPLETE, SECURE ONE CHARGING PUMP AND CONTINUE C C00LDOWN. G 2100 S S 2000 4 t3 e: 1900 1800 1700 1600 1500 I I i 1 O 2500 5000 7500 10000 12500 15000 TIME (SECONOS) 102
FIGURE 5-14 SIMULATION OF A BTP RSB S-1 SCENARIO 1200 ESS WIZER BAC M M MS 'M 1080 960 ~ 840 C u x 3 720 y E 600 "c Wr 480 _ ONLY lE PRESSURIZER HEATERS FROM THE OPERABLE DIESEL GENERATOR ARE 3 AVAILABLE FOR USE. TOTAL HEATER 360'_ CAPACITY IS 200kW. 240 120 O I i t O 2500 5000 7500 10000 12500 15000 TIME ISECONOS1 103
F I GUR E 5-15 SIMULATION 0F A BTP RSB 5-1 SCENRRID-100 m 90 80 70 5 60 4 E SECURE ONE CHARGING PUMP J 50 AND CONTINUE PLANT C00LDOWN. 40 30 COMMENCE SIMULTANEOUS PLANT B0 RATION AND C00LDOWN USING TWO CHARGING PUMPS. 20 _ 10 _ 0 t I I i 1 0 2500 5000 7500 10000 12500 15000 TIME (SECONOS1 104
F I GUR E-5-16 200 9 180 e 160 140 E 120 b COMMENCE SIMULTANEOUS PLANT 2o B0 RATION AND C00LDOWN g 100 80 k 60 Y BORATION COMPLETE, CONTINUE PLANT C00LDOWN 20 0 I i I I O 2500 5000 7500 10000 12500 15000 TIME ISECONOS1 105
F I GUR E 5-17 SIMULATION OF A BTP RSB 5-1 SCENARIO 700 s 650 e 600 550 L HOT LEG TEMPERATURE u_ e 500 N 2 [ E 450 COLD LEG TEMPERATURE ?= t 400 1 350 l 300 250 l 200 I I i l O 2500 5000 7500 10000 12500 15000 TIME (SECONOS1 106
FIGURE 5-18 ' SIMULATION OF A BTP RSB 5-1~ SCENARIO- .700 a SSO a 600 N 550 L HOT LEG TEMPERATURE E e E 500 U >? / g 450 COLD LEG TEMPERATURE t 5 400 350 l 300 250 s 200 1 I O 2500 5000 7500 10000 12500 15000 TIME tSECONOS1 107
FIGURE 5-19 SIMULATION-OF A BTP RSB 5-1 SCENARIO-700 3 650 O 600 L_ 550 u. e 500 Ea { 450 y 5 400 350 t 300 t' 250 _ 200 0 2500 5000 7500 10000 12500 15000 TIME (SECONOSI 108
] t.
- FIGUREL 5-20..
n-SIMULATION OF A BTP RSB 5-1-SCENARIO 0F W GIN OF SUBC00 LING 500 3 450 O-1 400 350 L u. 300 e u; O z -e 250 ee r. 200 150 I 3 100 \\ N 50 0 i i i i i 0 2500 5000 7500 10000 12500 15000 ( TIME ISECONOS) ? 109
FIGURE 5-21 SIMULATION OF A BTP RS8 5-1 SCENARIO STM GEN 1 NARROW RANGE LEVEL-100 'i 90 e 80 k g i 70 l l Ill )fg jl l t, 1 I s 3 60 I I f i u l 5 S I y 50 D a 40 s 30 20 I s l 10 0 1 1 I 1 s C 2500 5000 7500 10000 12500 15000 TIME tSECONOS1 i 110
'i
- ,37-- --
,, ~ ' x Q. y:. ": 7,. 3 FIGURE 5-22 cm t~ e SIMUERTION OF A BTP RSS 5-1 SCENAR,10 STri GEN 2 NARROW R4NGc^ LEVEL- t, 100 -.__.'2 s. i-90 o t e .t b 80 t p y ( \\ .y. h 70 4 ~ ) i i -l w i [ s ell i r I i i 5 60
- f. (
i d u T l
- 5. _
s' ew lc 1 I. 50 ~ w> w 40 w 30 20 10,_ 0 t I I i 1 0 2500 5000 7500 10000 12500 15000 TIME (SECONOS1 111
g:j., +3 W i; y1 ' kr (,i f. 1 )' t N/ -{. s,' FIGURE 5-23 - i ~- "J'.-J' SIMULRTION OF A BTP RSB 5-1 SCENRRIO STM GEN 1 PRESSURE- 'l100 ~ 1 ~ 1000 b-- 9*, p,. x, "? / C00LDOWN USING BOTH ATMOSPHERIC DUMP VALVES. . 900 ; ,a \\ s. g -, :~ ~, ' ( -800 I s e 's 700 s .s a. i i, . t6 e a w 600 m wt m eg 500 400 4 300 l 200 l T i i 100 t i i i 1 0 2500 5000 7500 10000 12500 !?iOOO TIF.E (SECON05) 112
F I GUR E 5-24 sit 1ULATION OF A BTP~RSB 5-1 SCENARIO STM GEN 2 PRESSURE 1100 = 1000 h-o C00LDOWN USING BOTH ATMOSPHERIC 900 / DUMP VALVES 9 / 800 l l E ~ 700 m b w g 600 u a. 500 400 300 h 200 _ 100 i t i t t 0 2500 5000 7500 10000 12500 15000 TIME (SECONOS) 113
FIGURE 5-25 SIMULATION OF A BTP RSB 5-1 SCENARIO STM GEN 1 AUXILIARY FEE 0 WATER FLOW 500._ i 450 w STEAM GENERATOR WATER LEVEL MAINTAINED 400 VIA MANUAL "0N/0FF" TYPE CONTROL OF THE 3 AUXILIARY FEEDWATER SYSTEM. 350 E 300 l l 1 J 250 I 200 ll ll i ll i! 150 lj I !j. I i, 100 i ~ f l l 50 l l l l 0_ 1 i l t t i t 0 2500 5000 7500 10000 12500 15000 TIME (SECONOS) 114
FIGURE 5-26 SIMULATION OF A BTP RSB 5-1 SCENARIO 500 STM GEN 2 AUXILIARY FEE 0 HATER FLOW t 450 t SEE NOTE ON FIGURE 5-25. 400 350 E 300 c e l i l 2 1 o l J 250 u l i l 200 i l l $ 1 il 150 ll i! l O i l l 100 s { f y 50 _ l l i i i O I i i i O 2500 5000 7500 10000 12500 15000 TIME (SECONOS1 115
r- 'I
6.0 REFERENCES
i 1. " Natural Circulation Test Program, San Onofre Nuclear Generating Station Unit 2, Safety Evaluation," CEN-201(S), April 1982. 2. USNRC Memorandum dated 19 October 1982, Docket File No. 50-3 361,
Subject:
San Onofre Unit 2 - Roron Mixing and Natural Circulation Tests. 21 3. " Final Safety Analysis Report for San Onofre Nuclear Generating Station, Units 2 & 3," Docket 50-161/362, Amendment 30, July 1982. 4 " Technical Specifications, San Onofre Nuclear Generating Station, Unit No. 2," NUREG-0741, February 1982. 5. " Guidance for Residual Heat Removal," Regulatory Guide 1.139, U.S. Nuclear Regulatory Commission, May 1978.- 6. " Safety Evaluation Report Related to the Operation of San Onofre Nuclear Generating Station, Units 2 & 3," NUREG-0712, February 1981. 7. "LOFTRAN Code Description," T.W.T. Burnett, C.J. McIntyre, and J.C. Buker, WCAP-7907, October 1972. = 8. Baron Injection Transient (BITRAN). A simplified code developed to efficiently model natural circulation ev'ents, Brookhaven National Laboratory (unpublished). 9. " Verification of C-E Mixing Model for Pressurized Thernal Shock Events with CREARE 1/5 Scale Data," CENPS0-204, September 1982. 10. " Analysis and Evaluation of St. Lucie Unit 1 Natural Circulaton s Cooldown," NSAC-16/INPD-21, December 1980. 116 )
W ENCLOSORE 2 4 9 f .}}