ML19305C604
| ML19305C604 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 03/27/1980 |
| From: | Mills L TENNESSEE VALLEY AUTHORITY |
| To: | Rubenstein L Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19305C605 | List: |
| References | |
| NUDOCS 8003310140 | |
| Download: ML19305C604 (36) | |
Text
{{#Wiki_filter:400 Chestnut Street Tower II March 27, 1980 Director of Nuclear Reactor Regulation Attention: Mr. L. S. Rubenstein, Acting Chief Light Water Reactors Branch No. 4 Division of Project Management U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Rubenstein:
In the Matter of the Application of ) Docket No. 327 Tennessee Valley Authority ) Enclosed for your review are ten copies of the safety evaluation for the nine tests proposed for the Sequoyah Nuclear Plant unit 1 special low power test program (Enclosure 1). This safety evaluation repre-sents the results of review by both Westinghouse and TVA. In particular, the safety evaluation addresses specific exemptions and modifications to the technical specifications required for performance of the special test program, criteria for manual reactor trip and safety injection, and evaluations and analyses of Final Safety Analysis Report chapter 15 events. Also enclosed for your review are ten copies of the procedures that will be used for the special test program (Enclosure 2). These procedures have been reviewed and found acceptable by Westinghouse. Preliminary procedures were transmitted to you in my January 18, 1980, letter on the special test program. If you have any questions, please get in touch with M. J. Burzynski at FTS 354-2581. Very truly yours, TENNESSEE VALLEY AUTh0RITY \ 68 0
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. M. Mills, Shnager Nuclear Regulation and Safety Enclosures 9 pu6 ph l
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ECCLOSURE 1 s EVALUATIO:! 0F SPECIAL TEST PROGRA'I FOR SEQUOYAH NUCLEAR PLA:iT U:!IT 1
1.0 DESCRIPTION
OF TESTS 1.1 NATURAL CIRCULATION TEST (TEST 1) Objective - To demonstrate the capability to remove decay heat by natural circulation. Method - The reactor is at approximately 3% power and all Reactor Coolant Pumps (RCP's) are operating. All RCP's are tripped simul-taneously with the establishment of natural circulation indicated by the core exit thermocouples and the wide range RTD's. 1.2 NATURAL CIRCULATION WITH SIMULATED LOSS OF OFFSITE AC POWER (TEST 2) Objective - To demonstrate that following a loss of offsite AC power, natural circulation can be established and maintained while being powered from the emergency diesel generators. Method - The reactor is at approximately 3% power and all RCP's are . operating. All RCP's are tripped and a station blackout is simulated. AC power is returned by the diesel genera. tors and natural circulation is verified. 1.3 NATURAL CIRCULATION WITH LOSS OF PRESSURIZER HEATERS (TEST 3) Objective - To demonstrate the ability to maintain natural circulation and saturation margin with the loss of pressurizer heaters.
'ethod - Establish natural circulation as in Test 1 and turn off the iressurizer heaters at the main control board. Monitor the system s, pressures to detet'ine; the saturation margin, the depressurization rate, and the effects of charging / letdown flow and steam generator pressure on the saturation margin.
1.4 EFFECT OF STEAM GENERATOR SECONDARY SIDE ISOLATION ON NATURAL CIRCULATION (TEST 4) Objective - To determine the effects of steam generator secondary side isolation on natural circulation. Method - Establish natural circulation conditions as in Test 1 but at 1% power. Isolate the feedwater and steam line for one steam generator and establish equilibrium. Repeat this for one more steam generator until two are isolated. Return the steam generators to service in reverse order. I
- 1.5 NATURAL CIRCULATION AT REDUCED PRESSURE (TEST 5)
Objective - To demonstrate the ability to maintain natural circulation at reduced pressure and saturation margin. The accuracy of the saturation margin program of the plant computer will also be verified. Method - The test method is the same as for Test 3, with the exception that the pressure decrease can be accelerated with the use of auxiliary pressurizer sprays. The saturation margin will be decreased to approxi-mately 20 y, 1.6 COOLDOWN CAPABILITY OF THE CHARGING AND LETDOWN SYSTEM (TEST 6) Objective - To determine the capability of the charging and letdown system to cool down the RCS with the steam generators isolated and one RCP operating. Method - With the reactor shut down, trip three of the RCP's and isolate all four of the steam generators. Vary the charging and letdown flows and monitor the primary system temperatures to determine the heat removal capability. 1.7 SIMULATED LOSS OF ALL ONSITE AND OFFSITE AC POWER (TEST 7) Objective - To demonstrate that following a loss of all onsite and off-site AC power the decay heat can be removed by natural circulation using the auxiliary feedwater system in the manual mode. Method - The reactor is at approximately 1% power and all RCP's are running. All RCP's are tripped and a total station blackout is simu-lated. Instrument and lighting power is provided by the backup batteries since the diesels are shut down. 1.8 ESTABLISHMENT OF NATURAL CIRCULATION FROM STAGNANT q0NDITIONS (TEST 8) Objective - To demonstrate the establishment of natural circulation from
- agnant (no flow) conditions in the primary system using reactor power to simulate decay heat.
Method - The reactor is critical in the HZP testing range and all RCP's are operating. Trip the RCP's and isolate the steam generators (feed and steam lines). When flow indicates zero, increase reactor power to approximately 1% and unisolate the steam generators. Natural circu-lation will be verified by observing the response of the hot leg temperatures and the core exit thermocouples. l l l l l
1.9 FORCED CIRCULATION C00LDOWN (TEST 9A) Objective - To determine the excore detector indicated power correction factor as a function of average cold leg temperature. This is required for Tests 4, 8, and 9B. Method - The reactor is at approximately 3% power and all RCP's are running. The primary system is cooled down in 10 F increments to 450 F with calorimetric data and the indice.ed excore power obtained for Data is also taken during the heatup to further correlation with T old. verify the results. 1.10 BORON MIXING AND COOLDOWN (TEST 9B) Objective - To demonstrate that the RCS can be uniformly borated while in natural circulation. Also demonstrate the capability to cool down the RCS in the natural circulation mode. Method - Establish natural circulation at approximately 3% power and with relatively deep rod insertion. Borate the RCS by approximately 100 ppm through the normal boration path maintaining reactor power with rod withdrawal. Once the system is mixed, cool down the RCS maintaining constant reactor power. I i - t
2.0 IMPACT ON TECHNICAL SPECIFICATIONS 2.0 Evaluation of the proposed tests indicates that 13 technical specification requirements will be violated and will therefore require exceptions during the performance of the special test program. The matrix below lists the technical specifications that will not be met for each test. Test 6 7 8 9A 9B Technical Specification 1 2 3 4 5 X X X X X X X X X 2.1.1 Core Safety Limits 2.2.1 Various Reactor Trips X X X X X X X Overtemperature AT X X X X X X X X X Overpower AT X X X X X X X X Steam Generator Level X X 3.1.1.3 Moderator Temperature X X Coefficient X X 3.1.1.4 Minimum Temgarature for Criticality X X X X 3.3.1 Various Reactor Trips Overtemperature AT X X X X X X X X Overpower AT X X X X X X X X X X X X X X X X Steam Generator Level X X 3.3.2 Safety Injection - All X X X X automatic functions X X X X X X Pressurizer X X X 3.4.4 3.5.1.2 Upper Head Injection X X X X X X X X X X System 3.7.1.2 Auxiliary Feedwater X X 3.8.1.1 AC Power Sources X X t 3.8.2.1 AC Oasite Power Distri-
. bution System X X 3.8.2.3 DC Distribution System X X 3.10.3 Special Test Exceptions -
Physics Tests X X X
~
2.1 Each technical specification requiring an exception is 1isted below alonF aith the reason for the exception and the basis for continued operation. 2.1.1 Reactor Core Safety Limits (T.S. 2.1.1) This specification restricts operation to within the nucleate boiling regime by limiting reactor coolant system average temperature as a function of reactor power and reactor coolant system pressure during 4 loop operation. In the natural circulation mode, this specification will not be met because the reactor coolant pumps will not be operating. l l s
A . During performance of the special tests, core exit thermo- - couple temperature will be limited to 610 F, core average temperature will be limited to 578 F, reactor coolant loop delta T will be limited to 65 F, and the reactor will be manually tripped if the margin to saturation reaches 150F. No boiling will be experienced in the core and the temperature limite of specification 2.1.1 will be met. 2.1.2 Reactor Trip System Instrumentation Setpoints (T.S. 2.2.1)
- 1. Overpower and Overtemperature AT The overtemperature and overpower AT trip functions obtain a temperature input from sensors located in the RTD bypass loops. During natural circulation, flow through the bypass loops will be extremely low and the temperature indication from these loops will be highly suspect. To prevent inadvertent tripping of the plant when in the natural circulation mode, these trip functions will be bypassed. Operator action will be relied on to provide the protective function previously furnished by the AT trips. This will be accomplished by manual reactor trip or safety injection actuation when:
Pressurizer Level (trip) < 17% span or an unexnlained decrease of more than 5% not concurrent with a T,y, change Pressu;izer Level (SI) < 10% span or an unexplained decrease of more than 10% not concurrent with a T,y, change Pressurizer Pressure (SI) - Decrease by 200 psi or more in an unplanned or unexplained manner Steam Generator Level (trip) < 5% narrow range span Steam Generator Level (SI) < 0% narrow range span or equivalent wide range level RCS subcooling (trip) <,150F T g ,g margin RCS subcooling (SI) < 10 F T,, margin NIS Power Range, 2 channels (trip) > 10% rated thermal power
- 2. Steam Generator Water Level This trip function ensures that there is a minimum volume of water in the steam generators above the tops
.~ - _ _. _ - _ -. - = . . ... -. . .. .- . . -.
of the U-tubes to maintain a secondary side heat
- sink. At times during the test program it may be difficult to hold the 12% margin between the normal operating level and the low-level trip setpoint at 21%
of span. The required setpoint (volume of water above the U-tubes) is dependent on the amount of core heat and the time required to establish adequate feed flow '
' tt the steam generators. With the plant limited to 5%
rated thermal power or less and being at BOL on Cycle 1, there will be little or no decay heat present; the heat source will be the core operating at the limited power level. Tripping the reactor on any of the different operable trip functions or the operator action points (see section on Operational Safety Criteria) will ensure that this requirement will be met. Thus, we find that it is acceptable to lower the trip setpoint to 5% span for all of the special tests. 2.1.3 Moderator Temperature coefficient (T.S. 3.1.1.3) The moderator temperature coe Nicient is limited to Opcm/ F or more negative to ensure that the value of the coefficient remains within the limiting conditions assumed for this parameter in the plant accident and transient analyses. When performing tests during which the primary system is cooled, the coefficient will become slightly positive. However,
- the isothermal temperature coefficient is expected to remain
, negative or approximately zero. This coefficient can be interpreted as the sum of the moderator and doppler coefficients; with.it at a zero or negative value, the i impact from rapid heatups and cooldowns will be minimized.
- In addition, the effect of a small positive moderator temperature coefficient has been considered in the analyses performed for the test conditions.
2.1.4 Minimum Temperature for criticality (T.S. 3.1.1.4) The minimum temperature for criticality is limited to 541 F i by specification 3.1.1.4 and 531 F by specification 3.10.3. To perform tests 4, 9A, and 9B, it is egpected that the RCS average temperature will drop below 531 F. We have deter-mined that operation with T as low as 425 F is acceptable assuming that: """ ! 1. Control Bank D is inserted to no deeper than 100 steps i withdrawn, and l
- 2. Power range neutron flux low setpoint and intermediate i
range neutron flux' reactor trip setpoints are reduced i from 25% RTP to 7% RTP. This will considerably reduc 6 tu consequences of possible ~ transients by (1) reducing in & tdu?. control rod worths ; (Bank D) on unplanned withdrawal, (2) reducing bank worth (Bank D) on unplanned withdrawal, (3) maximizing reactivity
- . . -a
incartion ecp:bility consictcnt with opsrstionsi rcquirs-ocnto, (4) limiting maximum pow:r to o vary low vclua on an unplanned power excursion, and (5) allowing the use of the "at power" reactor trips ae backup trips rather than as primary trips. 2.1.5 Reactor Trip System Instrumentation (T.S. 3.3.1) The overpower AT, overtemperature AT, and the steam generator water level trips will not meet the operability requirements of this specification. However, this speci-fication can be excepted for the reasons previously noted under Reactor Trip System Instrumentation Setpoints (T.S. 2.2.1). 2.1.6 Engineered Safety Feature Actuation System Instrumentation (T.S. 3.3.2) In order to prevent inadvertent safety injection during the performance of the special tests, the automatic injection functions will be blocked. The reactor trip, control room trip indications / alarms, and manual safety injection initiation will be operable. If the ESF instrumentation detects a condition which is over the trip setpoint, the safety injection signal will provide a reactor trip and control room indication / alarm. The automatic injection will be prevented by forcing the logic to see that the reactor trip breakers are open. If the operator determines that the actual situation requires injection, he will manually initiate safety injection using the control room hand switches. We believe that this mode of operation is acceptable for the short period of time these tests will be carried out based on'the folloying:
- 1. Close observatfor, of control room trip indications / alarms and rigid adherence to the action points specified in the section on Operational Safety Criteria will ensure manual safety injection actuation.
- 2. Little or no decay heat is present in the system, thus safety injection serves primarily as a pressurization function (shutdown margin capability is considerably more than 1.6% AK/K for control rods at or above the insertion limits).
Blocking these functicns will allow the performance of these tests at low power, pressure, or temperature while close operator surveillance will ensure initiation of safety injection, if required. 2.1.7 Pressurizer (T.S. 3.4.4) The pressurizer provides the means of maintaining pressure control for the plant. Normally this is accomplished through the use of pressurizer heaters and spray. In several tests the pressuri,-er heaters will be either turned off or rendered inoperable by loss of power. This mode of operation is acceptable in that pressure control will be maintained through the u a of pressurizer level and ' charging / letdown flow.
2.1.8 Upper Head Injection Accumulators (T.S. 3.5.1.2) During the performance of these tests the UHI system will be locked out. This is to prevent inadvertent actuation of the system. The UHI system provides borated water in the event of significant depressurization indicating a LOCA. It has been noted that with little or no decay heat at low power levels, this system provides little or no benefit in the event of a LOCA. Based on this knowledge, We have determined that it is acceptable to perform these , special tests with the UHI system locked out. 2.1.9 Auxiliary Feedwater System (T.S. 3.7.11 The auxiliary feedwater system will be rendered partially inoperabis for two tests. The two tests simulate some form of loss of AC power, i.e. , motor driven auxiliary feedwater pumps inoperable. Westinghouse has determined that this is acceptable for these two tests because of the little or no decay heat present allowing sufficient time (s 30 minutes) for operating personnel to rack in the pump power supplies and regain steam generator level. 2.1.10 Power Sources (T.S. 3.8.1.1, 3.8.2.1, and 3.8.2.3) Technical Specification 3.8.1.1 requires two physically independent circuits between the offsite transmission network and the onsite Class lE distribution systems and requires that four separate and independent diesel generator sets be operable. Technical Specification 3.8.2.1 requires that the 6900- and 480-volt shutdown boards and 120-volt AC vital instrument boards be operable and energized. Technical Specification 3.8.2.3 requires that the DC vital battery channels be energized and operable. Each channel shall include one 125-volt DC board, one 125-volt DC battery bank, and a full capacity charger. In test 2, the test switches on the 6900-volt shutdown boards logic relay panels 2A-A and 2B-B for diesels 2A-A and 2B-B are placed in the test positions. This action prevents the unit.two diesels from responding to a blackout signal. The circuit breakers between the unit one unit boards and 6900-volt shutdown boards lA-A and 1B-B are opened to simulate a loss of offsite power. If for some reason either of the diesels fail to start, power will be returned to the effected board by closing either the normal or alternate feeder breaker. l l 1
. Test 7 simulates a loss of all onsite and offsite AC power by selectively de-energizing system components and equipment. As par't of the test, the Class lE distrioution is aligned so that the only supply to the 125-volt vital AC instrument boards will be the 125-volt vital batteries.
This will be done opening the circuit breaker between the chargers and the battery boards. The 480-volt supply to the inverters will also be opened. At this point, the loads supplied by the 120-volt vital AC instrument boards will be supplied only by the batteries. If a problem develops with the batteries, power may be restored to the battery boards by reclosing the circuit breakers from the chargers or by closing the breakers for the 480-volt AC supply to the vital inverters. 2.1.11 - Special Test Exceptions - Physics Tests (T.S. _3.10.3) This specification allous the minimum temperature for criticality to be as low as 531 F. Since# it is expected that RCS T will be taken as low as 450 F, this speci-fication will be excepted. See Section 2.1.4 for basis of acceptability. 2.1.12 Technical Specifications Not Excepted While not applicable at power levels below 5% RTP, the following technical specification limits can be expected to be exceeded:
- 1. Heat Flux Hot Channel Factor - nF (Z) (T.S. 3.2.2)
At low temperatures and flows FnIZ) can be expected to be above normal for 5% RTP wIth RCP's running. However, at such a low power level no significant deviations in burnup or Xe peaks are expected.
- 2. RCS Flow Rate and R - (F Atlowtemperaturesand$$o)w(T.S.3.2.3)
F AH can be expected to be higher than if pumps are running. However, no significant consequences for full power operation are expected.
- 3. Quadrant Power Tilt Ratio (T.S. 3.2.4)
With no, one, or two pumps running and critical, core power distributions resulting in quadrant power tilt may form. At low power levels and for short periods of time, these tilts will not significantly influence core burnup.
- 4. DNB Parameters (T.S. 3.2.5)
In the performance of several tests the plant will be depressurized below 2230 psia. At low operating power levels this depressurization is not significant as long as subcooling margin is maintained. i I l l
. 3.0 OPERATIONAL SAFETY CRITERIA 3.0 During the performance of these tests, the operator must meet the following~ set of criteria for operation.
3.1 For all tests
- a. Primary System Subcooling (T sat Margin) > 20 F
- b. Steam Generator Water Level > 33% Narrow Range Span -
- c. Pressurizer Water Level (1) With RCP's Running > 25% Span (2) Natural Circulation > Value when RCP's tripped
- d. Loop AT < 65 F
- e. T 7 578 F '
- f. CSEE Exit Temperature (highest) 5610F
- g. Power Range Neutron Flux Low Setpoint and -
Intermediate Range Neutron Flux Reactor Trip Setpoints < 7% RTP
- h. Control Bank D 100 steps withdrawn or higher 3.2 Reactor trip and test termination must occur if any of the following conditions are met
- a. Primary System Subcooling (T sat K*#8i ") I 15 F
- b. Steam Generator Water Level < 5% Narrow Range Span
- c. NIS Power Range, 2 Channels > 107 RTP
- d. Pressurizer Water Level < 17% Span or an unexplained decrease of more than 57 not concurrent with a T- change
- e. Any Loop AT > 6$'E
- f. T > 578 F '
- g. CSEE Exit Temperature (highest) > 610 F
- h. Uncontrolled rod motion 3.3 Safety injection must be manually initiated if any of the following conditions are met
- a. Primary System Subcooling (T Margin) 5,10 F
- b. Steam Generator b'ater Level sat < 0% Narrow Range Span or Equivalent Wide Range Level
- c. Containment Pressure > 1.54 psig
- d. Pressurizer Water Level < 10% Span or an unexplained decrease of more than 107 not concurrent with a T change
- e. Pressurizer Pressure De:gEIses r by 200 psi or more in an unplanned or unexplained manner Safety injection termination must be in accordance with the termination
! criteria set forth in the plant Emergency Operating Instructions. i l os .
3.4 These operating and function initiating conditions are selected to
- ensure that the base conditions for safe operation are met, i.e.,
- a. Sufficient margin to saturation temperature at system pressure to ensure adequate core cooling,
- b. Sufficient steam generator level to ensure an adequate secondary side heat sink,
- c. Sufficient level in the pressurizer to ensure coverage of the heaters to maintain pressure control,
- d. Sufficient control rod worth to ensure adequate shutdown margin and minimize impact of uncontrolled bank withdrawal, and
- e. Limit maximum possible power level in the event of an uncontrolled power increase.
f
4.0 EVALUATION / ANALYSIS OF FSAR EVENTS In this section the safety effects of those special test conditions which are outside the bounds of conditions assumed in the FSAR are evaluated. The interaction of these conditions with the transients analyses in the FSAR are discussed. For those transients for which it is not obvious that the FSAR is limiting, new analyses are provided. 4.1 Evaluation of Transients The effect of the unusual operating conditions on the transients analyzed in the FSAR are evaluated. 4.1.1 CONDITION II - Incidents of Moderate Frequency 4.1.1.1 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from a Subcritical Condition Restriction of control rod operation to manual control and constant operator monitoring of rod position, nuclear power, and temperatures greatly reduce the likelihood of an uncontrolled RCCA withdrawal. Furthermore, during all but one test (9B) the total inserted control rod worth will be less than that required to go prompt critical. Operation without reactor coolant pumps, and in some cases with a positive moderator temperature
, reactivity coefficient, tend to make the consequences of RCCA withdrawal worse compared to the operating conditions assumed in the FSAR. For these reasons the operating procedures require that following any reactor trip at least one reactor coolant pump will be restarted and the reactor boron concentration will be such that it will not go critical with less than 100 steps with-drawal on D Bank. Thus, the only opportunity for a rod withdrawal from suberitical or zero power conditions without RCP flow is during test No. 8. An analysis of this event is presented in Section 4.2.1.
4.1.1.2 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power The same considerations discussed in Paragraph 4.2.1.1 apply here. In addition, the low operating power and the Power Range Neutron Flux Low and Intermediate Range Neutron Flux trip setpoints (7% RTP) act to mitigate this incident, while lack of the overtemperatura AT trip removes some of the protection provided in the FSAR case. An analysis is discussed in Paragraph 4.2.2. 4.1.1.3 Rod Cluster Control Assembly Misaligcment The FSAR discussion concerning static RCCA misalignment applies to the test conditions. The consequences of a st
. dropped RCCA would ba o dscrecca in p:ver. Thus no increase in probability or severity of this incident . is introduced by the test conditions.
4.1.1.4 Uncontrolled Boron Dilution The consequences of, and operator action time require-ments for, an uncontrolled boron dilution under the test conditions are bounded by those discussed in the FSAR. The fact that the control rods will never be inserted to the insertion limits, as well as the Power Range Neutron Flux Low Setpoint and the constant operator monitoring of reactor power, temperature and charging system operation, provides added protection. 4.1.1.5 Partial Loss of Forced Reactor Coolant Flow Because of the low power limits the consequences of loss of reactor coolant pump power are trivial; they are bounded by normal operating conditions for these tests. 4.1.1.6 Startup of an Inactive reactor Coolant Loop When at least one reactor coolant pump is operating, the power limit for these tests results in such small temperature differences in the reactor coolant system that startup of another loop cannot introduce a signi-ficant reactivity disturbance. In natural circulation operation, inadvertent startup of a pump would reduce the core water temperature and thus provide a change in reactivity and power. Because of the small moderator reactivity coefficient at beginning of life the power increase in the worst condition would be small and gradual and the flow to power ratio in the core would be increasing. The Power Range Neutron Flux Low Setpoint (7% RTP) reactor trip provides an upper bound on power. Because of the increase in flow-to-power ratio and because of the low setpoint on the reactor trip, DNB is precluded in this transient. 4.1.1.7 Loss of External Electrical Load and/or Turbine Trip Because of the low power level, the disturbance caused by any loss of load is small. The FSAR case is bounding; the turbine generator will not be in operation during the special test program. 4.1.1.8 Loss of .9ormal Feedwater i Because of the low power level, the consequences of a loss of feedwater are bounded by the FSAR case. In the case of loss of all feedwater sources, if the reactor is not shutdown manually, it would be tripped on Low-Low Steam Generator Water Level. Ample time is available to reinstitute auxiliary feedwater sources.
#1 n ,
4.1.1.9 Loss of Offsite Power to the Station Auxiliaries (Blackout) Because of the low power level, the consequences of a loss of offsite power are bounded by the FSAR case. 4.1.1.10 Excessive Heat Removal Due to Feedwater System Malfunctions The main feedwater control valves will not be used while the reactor is at power or near criticality on these tests. Thus, the potential water flow is restricted to the main feedwater bypass valve flow or auxiliary feed-water flow, about 15% of normal flow. The transient is further mitigated by the low operating. power level, small moderator temperature reactivity coefficient, the low setpoints on the Intermediate and Power Range Nuclear Flux trips, and close operator surveillance of feed flow, RCS temperatures, RCS pressure, and nuclear power. The case of excess feedwater with very low reactor coolant flow is among the cooldown transients discussed in more detail in Section 4.2.3. 4.1.1.11 Excessive Load Increase Incident
. The turbine will not be in use during the performance of these tests, and load control will be limited to operation of a single steam dump or steam relief valve. The small moderator temperature reactivity coefficient also reduces the consequences of this transient. Close operator surveillance of steam pressure, cold leg temperature, pressurizer pressure, and reactor power, with specific initiation criteria for manual reactor trip, protect against an unwanted reactor power increase. In addition, the low setpoints for Power-Range and Intermediate-Range Neutron Flux reactor trips (7% RTP) limit any power transient. In addition, modification of the high steamline flow setpoint allows a reactor trip on Low Steam Pressure only. Analyses are discussed in Section 4.2.3.
4.1.1.12 Accidental Depressurization of the Reactor Coolant System Close operator surveillance of pressurizer pressure and of hot leg subcooling, with specific initiation points for manual reactor trip, provides protection against DNB in the event of an accidental depressurization of the RCS. In addition, automatic reactor trip caused by the Low Pressurizer Pressure Safety Injection signal would occur even before core outlet subcooling reached the 15 F minimum limit specified for manual trip. During test 3 and 5, when this trip is bypassed to allow deliberate operation at low pressure, the pressurizer PORV block valves will be closed to remove the major credible source of rapid inadvertent depressurization. (The Low Pressure trip is automatically reinstated when pressure goes above i 1970 psi and the PORV block valves will be reopened ! at that time.) l l l p -
4.1.1.13 Accidental Depressurization of the Main Steam System The FSAR analysis for accidental steam system depressurization indicates that if the transient starts at hot shutdown conditions with the worst RCCA stuck out of the co ., the negative reactivity introduced by Safety Injection prevents the core from going critical. Because of the small moderator temperature reactivity coefficient which will exist during the test period, the reactor would remain suberitical even if it were cooled to room temperature even without Safety Injection. Thus the SAR analysis is bounding. 4.1.1.14 Spurious Initiation of the Safety Injection System at Power In order to reduce the possibility of unnecessary thermal fatigue cycling of the reactor coolant system components, the actuation of high head charging in the safety injection mode, and of the safety injection pumps, by any source except manual action will be disabled. Thus, the most likely sources of spurious Safety Injection, i.e., spurious or " spike" pressure or pressure-difference signals from the primary or secondary systems, have been eliminated. 4.1.2 CONDITION III - Incidents of Moderate Frequency 4.1.2.1 Loss of Reactor Coolant from Small Ruptured Pipes A review of the plant small break loss of coolant accident behavior during the low power testing sequence indicates that without automatic Safety Injection the system inventory and normal charging flow will provide the short term cooling for the small break transient. A sample calculation for a 2-inch break shows that the core remains covered for at least 5000 seconds. This is sufficient time for the operator to manually initiate SI and align the system for long term cooling.
, The magnitude of the resulting clad heatup transient during a LOCA event from these conditions is significantly reduced from the FSAR basis scenario due to the low decay heat and core stored energy which will exist at the time of these tests.
4.1.2.2 Minor Secondary System Pipe Breaks The consequences of minor secondary system pipe breaks are within the bounds of the transients discussed in Paragraph 4.2.3 (Analysis of Cooldown Transients). l 4.1.2.3 Single Rod Control Cluster Assembly Withdrawal at Power l j The single rod control cluster assembly withdrawal j at power under natural circulation flow was l l?
not analyzed. The FSAR analysis shows that assuming
- limited parameters for normal operation with forced flow, a maximum of 5% of the fuel rods could expetience a DNBR of less than 1.3 following a single RCC withdrawal. As pointed out in the FSAR, no single electrical or mechanical failure in the control system could cause such an event. In addition, the probability of this event happening during these tests is reduced by:
- 1. Close operator surveillance of control rod motion with instructions to trip the reactor if uncontrolled rod motion occurs.
- 2. Rod control is restricted to operator manual control therefore any rod operation will be under close operator surveillance and any rod motion which occurs without operator demand will be interpreted as uncontrolled rod motion.
> 3. Close operator surveillance of reactor power and hot leg temperature.
In summary, due to the extremely low probability of single rod control cluster assembly withdrawal at power during this test sequence, analysis of this event is not warranted. 4.1.2.4 Other Incidents of Moderate Frequency The consequences of an inadvertent loading of a fuel assembly into an improper position, complete loss of forced reactor coolant flow, and waste gas decay tank rupture, have been reviewed. The consequences of each of these events occurring during test conditions are bound by those described in the FSAR. 4.1.3 CONDITION IV - Limiting Faults 4.1.3.1 Major Rupture of Pipes Containing Reactor Coolant Up to and Including Double Ended Rupture of the Largest pipe in the Reactor Coolant System (Loss of Coolant Accident) A review of the plant loss of coolant accident behavior during the low power testing sequence indicates that without automatic safety injection there is sufficient cooling water readily available to prevent the fuel rod cladding from over heating. During the large break event, the reactor coolant system (RCS) will begin to depressurize to the containment, removing stored energy and decreasing the RCS inventory. The cold leg accumulators will begin injecting into the RCS, replacing inventory. As \ to
a result, the vessel is filled to the bottom of the nozzles (at no time is the core uncovered). At 100 seconds, there is enough water in the reactor vessel below the nozzles to keep the core covered for 1.7 hours. This is sufficient time for the operator to manually initiate SI and align the system for long term cooling. The magnitude of the resulting clad heatup transient during a LOCA event from these conditions is significantly - reduced from the FSAR basis scenario by the low decay heat and core stored energy existing at .the time of the performance of these tests. 4.1.3.2 Major Secondary System Pipe Rupture During a cooldown transient caused by the secondary system, protection against excessive reactor power is provided by the following: the small moderator tempera-ture reactivity coefficient; close operator surveillance of pressurizer pressure, cold leg temperature and reactor power, with specific criteria for manual reactor trip and safety-injection; low trip setpoints on the intermediate range and power range Neutron ' Flux trips; reactor trip and MSIV closure on low steam pressure; reactor trip on low pressurizer pressure. Following reactor trip, assuming the most reactive RCCA stuck ' out of the core, the reactor would remain subcritical even if it were cooled to room temperature. Transient analyses for a steam pipe rupture are provided in Section 4.2.3. The consequences of a main feedline rupture are bounded in the cooldown direction by the steam pipe rupture discussion. Because of the low operating power, the heatup aspects of a feedline rupture are bounded by the FSAR discussion. 4.1.3.3 Steam Generator Tube Rupture The steam generator tube rupture event may be categorized by two distinct phases. The initial phase of the event is analogous to a small LOCA event. Prior to operator-controlled system depressurization, the steam generator tube rupture is a special class of the small break LOCA transients, and the operator actions required to deal with the situation during this phase are identical to those required for mitigation of a small LOCA. Hence, evaluation of the steam generator tube rupture during this phase is wholly covered by the safety evaluation of the small LOCA. After the appropriate operator actions have taken place to deal with the initial LOCA phase of the event, the remainder of the steam generator tube rupture accident
/7
mitigation would consist of those operator actions
- required to isolate the faulted steam generator, cool-down the RCS, and depressurize the RCS to equilibrate primary RCS pressure with the faulted steam generator secondary pressure. These actions require utilization of the following systems:
- 1. Auxiliary feedwater control to faulted steam generator.
- 2. Steam line isolation of the faulted steam generator.
- 3. Steam relief capability of at least one nonfaulted steam generator.
- 4. RCS depressurization capability.
Evaluation of the TVA special test procedures has verified that all of the above systems are immediately available for operator control from the control room. Therefore, it is concluded that the ability to mitigate the steam ge er.to tu e r:pt"re avea.t is not compromised by the modifications required for operation at 5% power during the proposed tests, and that the analyses performed for the SAR regarding this event remain bounding. 4.1.3.4 Single Reactor Coolant Pump Locked Rotor Because of the low power level, the locking of a single reactor coolant pump rotor is inconsequential. 4.1.3.5 Fuel Handling Accident The FSAR analysis of a fuel handling accident is bounding. 4.1.3.6 Rupture of a Control Rod Drive Mechanism Except for a short period during test 9B, the control rod bank insertion will be so limited (i.e. , only Bank D inserted, with at least 100 steps withdrawn) that the worth of an ejected rod will be substantially less than the delayed neutron fraction. Thus, the power rise following a control rod ejection would be relatively gradual and terminated by the Power Range and Intermediate Range Neutron Flux reactor trips. During test 93, the control rods will be inserted further while the reactor coolant pumps are running. In this case, the bank insertion will be less than the insertion limit at normal no-load conditions, thus the ejected rod worth would be lower and the consequences of an ejected rod less severe than those described in the FSAR. In test 9B, immediately following RCP trip I l I ! I l f if I l
\
and coastdown, the control rod banks will be withdrawn
- gradually as boron is added until they reach the condition of only Bank D in, withdrawn to at least 100 steps. While the core power transient and power distribution following an RCCA ejection at this time would be less severe than those shown in the FSAR, the result of combining these ameliorating effects with the effect of the natural circulation flow rate on clad-to-water heat transfer and RCS pressure have not been analyzed. The extremely low probablility of -
an RCCA ejection during the brief period in the test sequence does not warrant such an analysis. 4.2 Analysis of Transients 4.2.1 ANALYSIS OF TRANSIENTS An analysis was performed to bound the transient for test 8, i.e., the only case in which the reactor will be near critical but not at power without reactor coolant pumps running. The methods and assumptions used in the FSAR, Section 15.2.1.1 were used with the following exceptions:
- 1. Reactor coolant flow was 0.1% of nominal.
' Control rod incremental worth and total worth were upper 2.
bcond values for the D bank initially 100 steps withdrawn.
- 3. Moderator temperature reactivity coefficient was an upper bound (positive) for any core average temperature at or above 425 y,
- 4. The lower bound for total delayed neutron fraction for the beginning of life for Cycle 1 was used.
- 5. Reactor trip was initiated at 10% of full power.
- 6. DNB was assumed to occur spontaneously at the hot spot, at the beginning of the transient.
The resulting nuclear power peaked at 44% of full power, as is shown in Figure 4.2.1. The peak clad temperature reached was under 1200 F, as is shown in Figure 4.2.2. No clad failure is expected as a result of this transient. 4.2.2 ANALYSIS OF RCC BANK WITHDRAWAL AT POWER Analyses of RCCA bank withdrawal transients were performed for natural circulation conditions. The transients were i assumed to start from steady-state operating conditions at either 1% or 5% of full power, and with either all steamline isolation valves open or two of those valves closed. A l
.1
range of reactivity insertion rates up to the maximum for two banks moving was assumed for cases with all steamlines open, and up to the maximum for one bank moving for the cases with two steamlines isolated. Both maximum and minimum bounds on reactivity feedback coef ficients for beginning of life, Cycle 1, were investigated. In all cases, reactor trip was initiated at 10% nuclear power. Reactor conditions at the time of maximum core heat flux are shown in Figures 4.2.3 and 4.2.4 as functions of the reactivity insertion rate for three, four-loop active cases. For high reactivity insertion rates, the minimum reactivity coef ficient cases give the greatest heat flux af ter the trip setpoint is reached, and have the lowest coolant flow rate at the time of peak heat flux. For these cases, even the slowest reactivity insertion rates studied did not result in any increase in core inlet temperature at the time of peak heat flux. For maximum feedback cases, however, the transients for very low reactivity insertion rates go on for so long that the core inlet temperature finally increases before trip, i.e., af ter approximately one and one-half minutes of continuous withdrawal. Thus, the cases shown bound the worst cases. 4.2.3 ANALYSIS OF C00LDOWN TRANSIENTS Cooldown transients include feedwater system malfunctions, excessive steam load increase, accidental depressurization of the main steam system, and minor and major secondary system pipe ruptures. Attention has been focused on the possibility and magnitude of core power transients resulting from such cooldowns before reactor trip would occur. (Following reactor trip, no cooldown event would return the reactor to a critical condition.) During natural circulation operation, approximately one to two minutes would elapse following a secondary side event before cold primary water reached the core; thus, considering the close and constant surveillance during these tests, ample time would be available for the operator to resnond to such an event. Analyses were also performed to deter-mine the extent of protection provided by automatic l protection systems under trip conditions. l 4.2.3.1 Load Increases A load increase or a small pipe break, equivalent to the opening of a single power-operated steam pressure relief valve, a dump valve, or a safety valve, would cause an increase of less than four percent in reactor power, with a corresponding increase in core flow with nator.1 circulation, assuming the bounding negative i l l i 5k
moderator temperature coefficient for the beginning
- of life, Cycle 1. Thus no automatic protection is required, and ample time is available to the operator to trip the reactor, isolate feedwater to the faulted steam generator, and isolate the break to the extent possible. Calculated results for the sudden opening of a single steam valve, assuming the most negative BOL Cycle 1, moderator reactivity coef ficient are shown in Figures 4.2.5 and 4.2.6 (5% initial power, steam dump valve), and in Figures 4.2.7 and 4.2.8 (1%
initial power, two steam generators isolated, pressure relief valve). 4.2.3.2 High Flux Protection Reactor trip on high nuclear flux provides backup protection for larger pipe breaks or load increases. Analyses were performed to determine the worst core conditions that could prevail at the time of high-flux trip, independent of the cause. The following assumptions were used:
- 1. Upper-bound negative moderator isothermal temperature coefficient, vs. core average temperature, for beginning of life, Cycle 1.
- 2. Lower-bound fuel temperature - power reactivity coefficient.
- 3. Initial operation with core inlet temperature 5550F.
- 4. Initial powers of 0% and 5% of full power were analyzed.
- 5. Hot leg coolant at incipient boiling at the time of reactor trip. This results in some boiling in the reactor. The negative reactivity intro-duced by core boiling would effectively limit power; this negative reactivity was conservatively neglected.
- 6. Uniform core inlet temperature and flow.
- 7. Reactor trip equivalent to 10% of full power at the initial inlet temperature. The power as measured by the NIS is assumed to be diminished
! from the true power by 1% for each 1 F decrease I in reactor inlet temperature, resulting in a true i power of greater than 10% at the time of trip.
- 8. Core flow rate as a function of core power was conservatively assumed equal to the predicted I 6'o. under steady-state operating. conditions.
l 1 l l at l
Analyses of core conditions based on these assumptions indicate that the DNB criterion of the FSAR is met. 4.2.3.3 Secondary Pressure Trip Protection Large steamline ruptures devnstream of the steamline check valve's will affect all loops uniformly since the lines enter a common header just upstream of the HP turbine. Since the high steamline flow setpoint is set to zero flow (that is, the bystable is in the tripped condition), reactor trip and MSIV isolation will be activated by two out of four low steamline pressure signals. Low pressurizer pressure and Power Range Neutron Flux trip setpoints serve as further backups. The response to a double ended main steamline rupture downstream of the check valves is shown in Figures 4.2.9 and 4.2.10. An initial power of 57 and natural circu-lation conditions are assumed. In the example shown, the main steamline isolation valve on loop one was assumed to fail to close. No power excursion resulted, and the reactor remained subcritical after the trip. For large steamline ruptures inside of the steamline check valves, the normal protection is provided by the steamline delta-P trip and isolation. However, this signal has been disabled for these tests since the steamline pressure difference is expected to exceed the trip setpoint range. For this transient, the Power Range Neutron Flux trip serves as a backup to manual protection. An example case is shown in Figures 4.2.11 and 4.2.12. For this case, operation of 1% power was assumed with steam generators three and four isolated and with natural circulation. A double-ended rupture of the main steamline upstream , of the steamline venturi was assumed. The transient was allowed to continue without manual trip until the Power Range Neutron Flux trip was reached at 104 seconds. Ja
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= 1000. 00 -
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# 5% POWER, ALL LOOPS ACTIVE 4
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/ FOLLOWING THE OPENING OF A STEAM DUMP VALVE FROM 5% POWER, ALL LOOPS ACTIVE e
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/ FOLLOWING THE OPENING OF A SINGLE STEAM RELIEF VALVE AT 1%
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FIGURE 4.2.8 TRANSIENTS IN THE PRESSURIZER AND STEAM GENERATOR
' FOLLOWING THE OPENING OF A SINGLE STEAM RELIEF VALVE AT
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,' FOLLOWING A DOUBLE ENDED RUPTURE OF A MAIN STEAM-l
! LINE DOWNSTREAM OF THE STEAMLINE ISOLATION AND CHECK VALVES AT 5% POWER, ALL LOOPS ACTIVE
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TIME. (SEC)
, FIGURE 4.2.10 TRANSIENTS IN THE PRESSURIZER AND STEAM CENERATOR FOLLOWJNG A DOUBLE ENDED RUPTURE OF A MAIN STEAMLINE DOWNSTREAM OF THE STEAMLINE ISOLATION AND CHECK l
VALVES AT 5% POWER, ALI. LOOPS ACTIVE
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o ru o m n s - - - TIME (SEC) . FIGURE 4.2.11 TRANSIENTS IN THE REACTOR CORE AND COOLANT LOOPS FOLLOWING A DOUBLE ENDED RUPTURE OF A t L l MAIN STEAMLINE UPSTREAM OF Ti!E STEAMLINE 1 l VENTURI AT 17. POWER, TWO LOOPS ISOLATED l l I ! l l
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l TIME (SEC)
/
. FIGURE 4.2.12 TRANSIENTS IN THE PRESSURIZER AND STEAM GENERATOR FOLLOWING A DOUBLE ENDED RUPTURE OF A MAIN STEAMLINE UPSTREAM OF THE STEAMLINE VENTURI AT 1% POWER, TWO LOOPS ISOLATED I
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