ML20032D745
| ML20032D745 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 11/10/1981 |
| From: | METROPOLITAN EDISON CO. |
| To: | |
| Shared Package | |
| ML20032D738 | List: |
| References | |
| NUDOCS 8111170443 | |
| Download: ML20032D745 (14) | |
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TECHNICAL SPECIFICATION CHANGE REQUEST No. 108 The Licensee requests that the attached pages 111, 95a, 95b and 95c be added to Appendix A of the existing Technical Specifications for TMI-1.
II.
REASON FOR THE CHANGE REQUEST The attached pages are being added to the existing Technical Specifications for TMI-l'to permit the performance of a Low Power Natural Circulation Test (LPNCT) for the Rest. art of TMI-1 (Cycle 5).
Preliminary information con-cerning this test was provided'in our letter of May 6, 1981 (L1L 134) and
. was the subject of meetings with the NRC on April 22, 1981 and August 1981. ~ Additionally, presentations concerning the test were made before the ACRS on June 26, 1981 and July 10, 1981.
III.
SAFETY EVALUATION JUS 7_IFYING CHANGE i
(see Attachment 2)
IV.
AMENDMENT CLASSIFICATION (10 CFR 170.22)
This change requests an cdditional provision to the existing Technical Specifications involving a safety issue and is therefore, considered a Class III License Amendment. Therefore, enclosed, please find the pre-4 scribed remittance of $4,000.00.
V.
IMPLEMENTATION PERIOD This Technical Specification Change Request (TSCR) applies only to the Restart of TMI-1 for Cycle 5.
Upon completion of that test program, the resulting amendment should be cancelled.
i 8111170443 811110 PDRADOCK05000g P
TABLE OF CONTENTS Section h
3.16 SHOCK SUPPRESSORS '(SNUBBERS) 3-63 3.17 REACTOR-BUILDING AIR TEl(PERATURE 3-80 3.18 FIRE PROTECTION 3-86 3.18.1 FIRE DETECTION INSTRUMENTATION 3-86 3.18.2-FIRE SUPPRESSION WATER SYSTDi 3-88 3.18.3 DELUGE / SPRINKLER SYSTDiS 3-89 3.18.4 C0 SYSTDI 3-90 3.19 CONTAIhMENTSYSTEMS 3-95 3.20 5PECIAL TEST EXCEPTIONS 3-95a 3.20.1 LOW POWER NATURAL CIRCULATION TEST 3-95a 3.21 RADIOACTIVE ENVIRONMENTAL SPECIFICATIONS 3-96 3.21.1 RADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATIM 3-96 3.21.2 RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING 3-100 INSTRUMENTATION 3.22.1.1 LIQUID EFFLUENTS 3-106 3.22.1.2 DOSE 3-107 3.22.1.3 LIQUID WASTE TREATMEN!
3-109 3.22.1.4 LIQUID HOLDUP TANKS 3-110
)
l 3.22.2.1 DOSE PATE 3-111 3.22.2.2 DOSE, NOBLE GAS 3-112 3.22.2.3 DOSE, RADIOIODIMES, RADI0 ACTIVE MATERIAL IN PARTICULATE 3-113 FORM AND RADIONUCLIDES OTHER THAN NOBLE GASES 3.22.2.4 GASEOUS RADWASTE TREATMENT 3-115 3.22.2.5 EXPLOSIVE GAS MIXTURE 3-116
)
3.22.2.6 GAS STORAGE TANKS 3-117 3.22.3.1 SOLID RADIOACTIVE WASTE 3-118 3.22.4 TOTAL DOSE 3-119 j
3.23.1 MONITORING PROGRAM 3-120 j
3.23.2 IAND USE CENSUS 3-125 3.23.3 INTERLABORATORY COMPARISON PROGRAM 3-127 4
SURVEILLANCE STANDARDS 4-1 l
4.1 OPERATIONAL SAFETY REVIEW 4-1 4.2 REACTOR COOLANT SYSTEM INSERVICE INSPECTION 4-11 i
i 4.3 TESTING FOLLOWING OPENING OF SYSTDI 4-28 i
4.4 ' REACTOR BUILDING 4-29 4.4.1 CONTAINMENT LEAKAGE TESTS 4-29 4.4.2 STRUCTURAL INTEGRITY 4-35 4.4.3 HYDROGEN PURGE SYSTEM 4-37 4.5 DIERGENCY LOADING SEQUENCE AND POWER TRANSFER, EMERGENCY 4-39 CORE CCOLING SYSTEM AND REACTOR BUILDING COOLING SYSTDI PERIODIC TESTING 4.5.1 EMERGENCY LOADING SEQUENCE 4-39
- 4. 5. 2 -
DIERGENCY CORE COOLING SYSIDI 4-41 4.5.3 REACTOR BUILDING COOLING AND ISOLATION SYSTDi 4-43 4.5.4 DECAY HEAT RD10 VAL SYSTD1 LEAKAGE 4-45 i
4.6 DiERGENCY POWER SYSTEM PERIODIC TESTS 4-46 l
i Amendment No.
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3.20 SPECIAL TEST EXCEPTIONS 3.20.1 Low Power Natural Circulation Test (LPNCT)
Applicability During the performance of the Low Power Natural Circulation Test for Cycle 5 Restart. This Technical Specification is cancelled following the co=pletion of this test.
Obj ective To provide meaningful technical information concerning natural circulation at IMI-l and enhance operator training under both normal and certain degraded conditions.
Specification:
3.20.1.1 The limitations of Specifications 3.1.3.1 and 3.1.3.3 may be suspended during the LPNCT - Determination of Indicated Reactor Power Correction Factor provided:
The thermal power level does not exceed 5% of rated a.
Thermal Power.
b.
With the reactor coolant temperature below 525 F, continuous visual monitoring will be initiated within 15 minutes. A manual trip will be initiated if the RCS temperature drops below 520 F.
3.20.1.2 The li=itations of Specifications 2.1.2, 2.3.1, 3.5.1.1, 3.5.1.3 (0/ imbalance / flow & pump monitors) and 3.1.1.1 a. and b., may be suspended during the LPNCT - Establishment of Natural Circulation Flow / Determination of the Effect of Decreased OTSG Levels on Natural Circulation Flow Provided:
a.
The control rod index shall be maintained within the limits required for 100% FP operation specified in T.S. Figure 3.5-2A.
b.
The subcooling margin is greater than or equal to 50 F.
The EPS Overpower trip is less than or equal to 7% full power.
c.
d.
The average of the five highest core exit thermocouple tempera-tures is less than or equal to 610 F.
T hot is maintained less than or equal to 600 F e.
With the reactor coolant system outside the limits of any of the limits of a. through c.,
a manual trip of the reactor shall be initiated.
3.20.1.3 The limitations of Specifications 2.1.1, 2.1.2, 2.3.1, 3.1.1.1 a.
6 b. 3.5.1.1 and 3.5.1.3 (flux / flow / imbalance, puup monitors, low pressure, and variable low pressure) may be suspended during the LPNCT - Verification of the adequacy of the Pressurizer Heaters on the Emergency Bus provided:
3-95a
a.
The. control rod ~ index shall be maintained within the limits required for 100% full power operation specified in T.S. Figure 3.5-2A.
b.
The sabcooling margin is greater than or equal to 20 F.
c.
The RPS Overpower trip is less than or equal to 7% full power.
d.
The average of the five highest core exit thermocouple termpera-tures is less than or equal to 610 F.
e.
RCS pressure is between 1700 psig and 2300 psig.
f.
T is ai tained less than or equal to 600 F.
hot With the reactor coolant system outside the limits of any of the above limits of a. -
f., a manual trip to the reactor shall be initiated.
Bases During the performance of the special tests, the combination of administrative limits on reactor power <5% FP and subcooling margin 250 F, RPS overpower trip limit <7% FP and manual trip limits of core exit thermocouple temperature <610 F, loop I (hot) <600 F, loop Tave >525 F, and subcooling margin 320 F will insure that the integrity of the fuel cladding will be maintained.
During the performance of the special tests, removal of forced flow will be initiated at a reactor thermal power level of approximately 3% FP with an administrative control limit of 5% FP and RPS overpower trip setting of 7%
FP.
Table 14-13 of the Three Mile Island, Unit 1 FSAR provides calculated values that indicate that the expected natural circulation flow rates at less than or equal to 5% FP will be in excess of flow rates required for the removal of core decay heat without formation of voids in the core or reactor outlet piping.
The power imbalance portion of overpower trip based on reactor coolant flow and reactor power imbalance protects the core from center-line fuel melt by limiting the maximum linear heat rate (Kw/ft) in the fuel at high reactor thermal power levels. The overpower trip setting at 7% FP will prevent exceeding the Kw/ft limit regardless of the core imbalance.
The low pressure and variable low pressure trip setpoints have been established to maintain the D:iB ratio 1 1.3 for these design accidents that result in RCS pressure reductions. To prevent an inadvertent reactor trip during the perform-ance of 3.20.1.3,the variable low pressure trip will be bypassed.
However, constant operator surveillancc of reactor power level, reactor coolant system 'emperatures and pressure, and subcooling margin and operator action will t
be relied on to provide the protection function normally furnished by this RPS trip.
This will be accomplished by manual reactor trip if any cf tne limits specified are reached.
The low pressure trip prevents operation at pressures which might reduce DNBR margin and provides for reactor trip prior to ESAS actuation on low RCS pressure.
In order to retain some automatic low pressure protection and still provide 3-95b
a operation flexibility.and prevent an inadvertent reactor trip during the performance of 3.20.1.3 only, the low pressure trip setpoint will be, lowered t'
to 1700 psig.
Based on TMI-1, Cycle 5 Physics Test Manual predictions, it is expected that a slightly negative moderator temperature coefficient will exist for the conditions of the. low power natural. circulation testing. This prediction will.
be verified based on the hot zero power moderator temperature coefficient mea-sured during the.zero power physics testing program prior. to power escalation.
a
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This technical' specification delineates the conditions of the unit instrumentation and safety circuits necessary to assure reactor safety. Technical Specification 3.5.1.1 states that the reactor shall not be in a startup mode or in a critical state unless the requirements of Table 3.5-1, columns A and B are met.
Parameter indications from these specified instrument channels will be available at all times but the trip functions for flux / imbalance / flow, power / number of pumps, 1
and pressure / temperature (variable low pressure) will be defeated during pre-
)
viously specified tests.
Technical Specification.3.5.1.3 states that if the number of protection channels operable is less than the limit given in Table 3.5-1, column A, operation shall be limited as'specified in Column C.
In the cases of the above three protection
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channels defeated, operation would be. limited to Hot Shutdown.
I It should be noted that during natural circulation testing, the sensitivity of the reactor power (power range) indicator will be increased such that the actual power is 10 times less than the " indicated" reactor power (i.e. 50%
power read on the meter equals 5% power).
This will necessitate bypassing the planned anticipatory trips for loss of turbine (< 20%. indicated reactor power) and loss of both main feed pumps (< 7% indicated reactor power).
-i Technical Specification Section 3.1.'9 " Low Power Physics Testing Restrictions" i
do not apply during this test.
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References (1) FSAR Chapter 14 l
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- S ATTACHMENT 2 2.0 _ IMPACT ON TECHNICAL SPECIFICATIONS Evaluation of the proposed low power natural circulation test program indicates that eight (8) technical specification requirements will require exceptions during the performance of this test program. The matrix below lists the technical specifications which will require exceptions for each test.
Test Number Technical Specification 1-2 3
4 5
5 2.1 Safety Limits, Reactor Core 2.1.1 Reactor system pressure and -
X coolant temperature 2.1.2 Combination of reactor thermal X
X X
power and reactor power imbalance for specified flow 2.3 Limiting Safety System Settings 2.3.1 Protection Instrumentation-kPS Trip Setting Limits 0/ imbalance / flow X
X X
Pump monitors X
X X
Low Pressure X
Variable low pressure X
3.0 Limiting Conditions for Operation j
3.1.1.1 Reactor coolant pump operation X
X X
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3.1.3 Minimum conditions for criticality 3.1.3.1 FCS te=perature shall be above X
i 525 F l
3.1.3.3 RCS < 525 F the reactor shall be X
subcritical. by an amount >
f reactivity insertion due to de-pressurization 3.5.1 Operation safety instrumentaticn I
3.5.1.1 Requirements of Table 3.5.1 X
X X-l 3.5.1.3 Less than minimum number of pro -
X X
X tections channels operable as given in Table 3.5-1, column A operation shall be limited to column C.
X - Exception to Technical Specification Required l
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'I 2.1 Each techafcal specification requiring exception is listed below along with the reason for the exception and the basis for continued operation.
The. discussions of TS 3.1.3.1 and 3.1.3.3 in Section 2.1.4 apply only to Test #1 - Determination of Indicated Reactor Power Correction Factor.
The remaining discussions in Sections 2.1.1, 2.1.2, 2.1.3 and 2.1.5 apply to Test #4 - Establishment of Natural Circulation Flow, Test #5 - Verification of the Adequacy of the Pressurifer Heaters on the Emergency Bus and Test
- 6 - Determination of the Effect of Decreased OTSG Levels on Natural Circulation Flow. Test #2 - E' vgency Feedwater Actuation Test and Test
- 3 - Verification of Two Hour / '
Supply Capability of the Bottled Backup Air Supply require no technical 1ecification exceptions.
2.1.1 Reactor Core Safety Limits iTS 2.1)
Technical Specification 2.1.1 states that the reactor system pressure
)
and coolant temperature shall not exceed the safety limit as defined by the locus of points established in Figure 2.1-1.
Technical Specification 2.1.2 limits the combination of reactor thermal power j
and reactor power imbalance to the safety limit as defined by the locus of points for the specified reactor coolant flow races set forth in Figure 2.1-2.
These two technical specifications restrict reactor operation to within the nucleate boiling regime.
During the performance of the special tests, the combination of adginistrative limits on reactor power < 5% FP and subcooling margin 2; 50 F, RPS overpower trip limit < 7% FP and manual trip limits of core axigthermocoupletemperature<[6f0F,loopThot<600F,loopTave2; 525 F, and subcooling margin 2;20 F along with the automatic low pressure trip will insure that the integrity of the fuel cladding will be maintained.
l 2.1.2 Limiting Safety System Settings (TS 2.3)
Technical Specification 2.3.1 states that the reactor protection system trip setting limits and the permissible bypasses for the instrument channels shall be as stated in Table 2.3.1 and Figure 2.3-2.
The overpower trip based on reactor coolant flow and reactor power imbalance protects the core from DNB due to a loss-of-coolant flow accident from high reactor thermal power levels.
The overpower trip 1
based on pump monitors protects the core from DNB due to loss of reactor coolant pump (s) and restricts the power level for the number cf pumps in operation.
During the performance of the special tests, removal of forced flow will be initiated at a reactor thermal power level of approximately j
3% FP with administrative control limit of 5% FP and a RPS over'ower trip setting of 7% FP.
Table 14-13 of the Three Mile Island, Unit 1 FSAR provides calculated values that indicate that the expected natural circulation flow rates at less than or equal to 5% FP will be in excess of flow rates required for the removal of core decay heat without formation of voids in the core or reactor outlet piping.
l 1
i The' power imbalance portion of overpower trip based on reactor coolant flow and reactor power imbalance protects the core from center-line fuel melt by limiting the maximum linear heat rate (Kw/ft) in the fuel at high reactor thermal power levels. The overpower: trip setting at 7% FP will prevent exceeding the Kw/ft limit regardless of the core imbalance.
The low pressure and variable low pressure trip setpoints have been established to maintaan the DNB ratio > 1.3 for those design accidents that result in RCS pressure reductions. To prevent an inadvertent reactor trip during the performance of Test #5, the variable low pressure trip will be bypassed. However, constant operator surveillance of reactor power level, reactor coolant system temperatures and pressure, and subcooling margin and operator action will be relied on to provide the protection function normally furnished by this RPS trip.
This will be acco=plished by manual reactor trip if any of the limits specified are reached.
The low pressure trip prevents operation at pressures which might reduce DNBR margin and provides for reactor trip prior to ESAS actuation on low RCS pressure.
In order to retain some automatic low pressure protection and still provide operation flexibility and prevent an inadvertent reactor trip during the performance of Test #5 only, the low pressure trip setpoint will be lowered to 1700 psig.
2.1.3 Limitine Conditions for operation (TS 3.0)
Technical Specification 3.1.1.1.a states that the reactor coolant pump combinations permissible for given reactor thermal power levels shall be as shown in Table 2.3.1.
The previous discussions of the overpower trip based on reactor coolant flow and reactor power imbalance and the overpower trip based on pump monitors are also applicable to this technical specification.
2.1.4 Minimum Conditions for Criticality (TS 3.1.3)
Technical Specification 3.1.3.1 states that the reactor coolant temperature shall be above 525 F except for low power physics testing.
This technical specification limits to an acceptable level the magnitude of any power excursions resulting from reactivity insertion due to a positive moderator temperature coefficient.
Based on TMI-1 Cycle 5 Physics Test Manual predictions, it is expected that a slightly negative moderator temperature coefficient will exist for the conditions of the low power natural circulaticn testing.
This prediction will be verified based on the hot zero power moderator temp-erature coefficient measured during the zero power physics testing prcgram prior to power escalation.
It is anticipated that the 525 F limit will not be exceeded. However, if it is, visual monitoring will be initiated within 15 minutes to acnitor reactor coolant temperature and a manual trip will be initiated if temp-erature decreases below 520 F.
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'i Technical Specification 3.1.3.3 states that whenever reactor coolant temperature is less than 525 F the reactor shall be sub-critical by an amount equal to or greater than the calculated reactivityinsertionduetodepressurizationwhgchisapproximately 0.1% Ak/k..
It is not anticipated _that the 525 F limit will be exceeded and operator actions discussed above will be ~ in eff 7ct.
Although ghe reactor would not be placed in a suberitical co.1dition until 520 F is exceeded,' ample shutdown nargin will exist due to limits on control rod insertion.
2.1.5 Operational Safety Instrumentation (TS 3.5)
I This technical specification delineates the conditions of the unit instrumentation and safety circuits necessary to assure reactor safety.
Technical Specification 3.5.1.1 states that the reactor shall not be in a startup mode or in a critical state unless the re-quirements of Table 3.5-1, columns A and B are met.
Parameter indications from these specified instrument channels will be l
available at all times but the trip functions for flux / imbalance / flow, power / number of pumps, and pressure / temperature (variable low pressure) j will be defeated during previously specified tests.
Technical Specification 3.5.1.3 states that if the number of protection channels operable is less than the limit given in Table 3.5-1, Column A, operation shall be limited as specified in Column C.
In the cases of the above three protection channels defeated, operation would be limited to Hot Shutdown. Justification for continued reactor operation has previously been discussed in Sectbns 2.1.1 and 2.1.2 of this report.
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's 3.0 Evaluation /Ahalysis of FSAR Events The low power natural circulation tests will require plant initial conditions for some limited period of time that are not considered normal operating conditions. These initial test conditions require that exception.
be taken to certain specified Technical Specification Limits, normal operating procedures and normal Reactor Protection System trip functions.
The events analyzed in Chapter 14 of the FSAR have been re-examined and their consequences evaluated relative to the expected plant conditions during this testing sequence. A discussion of the impact of testing conditions on the consequences of each transient is provided.
In general, events are bounded by conditions presented in the FSAR or can readily be determined to have consequences less severe than those specifically analyzed in the FSAR.
3.1 Evaluation of Transients Each FSAR event is examined for the influence of the Natural Circulation Test conditions on the transient consequences.
3.1.1 Core and Coolant Boundary Protection Analysis 3.1.1.1 Uncompensated Operating Reactivity Changes Reactivity changes associated with fuel depletion and the builduo of fission product concentrations are associated witn long term normal operation.
No additional compensation for these reactivity changes is expected to be required.
3.1.1.2 Startup Accident The reactivity excursion due to an advertent CRA group withdrawal during the criticality procedure is bounded by the FSAR analysis.
Requirements for the immediate start of one reactor coolant pump in each loop following manual or automatic reactor trip and the subsequent start of the remaining two reactor coolant pumps prior to approach to criticality result in a normal RCS configuration during any approach to criticality during this test program.
Additional prctection from a large reactivity excursion is provide.d by a lowered high flux trip setpoint, Manual control of the control rods during approach to critical reduces the likelihood of a startup accident.
Rod Withdrawal Accident At Rated Power Operation 3.1.1.3 Power Operation is limited to approximately 5% FP with ob the high flux trip set to 7% of full power for the entire low power test program. The FSAR considers low power operation with forced RC flow which bound the test conditions for Tests 1 through 3.
Because of the operational and trip limits on maximum reactor power and the resulting low heat fluxes produced in the core, there is not concern
o for clad integrity during tne natural circulation conditions in Test 4 through 6.
3.1.1.4 Moderator Dilution Event The moderator dilution event is a slow reactivity excursion with protection generally provided by the high RC pressure trip.
The boron concentration will be greater than normal and with the increased post trip suberitical margin, longer times are expected for o erator action to c
terminate the source of dilution.
DNB considerations are similar to the CRA withdrawal event for the natural circulation conditions of Test 4 through 6.
The consequences of the moderator dilution event and operator action requirements are therefore bounded by the FSAR analysis.
3.1.1.5 Cold k'ater Accident (Pump Startup)
The startup of a reactor coolant pump while in natural circulation operation would reduce the core average moderator temperature and provide a small change in core reactivity and power. Any power excursion is limited by the low RPS high flux trip gof 7* FP, and the reactivity change is limited by the slightly negative moderator temperature coefficient associate with this tire in core life.
Increases in the ratio of flow and pressure to the power being produced preclude DNB during this transient.
3.1.1.6 Less of Coolant Flow Very low power operation makes the consequences of the loss of forced coolant flow inconsequential. Test 4 is an intentional loss of flow to the natural circulation mode of operation and will be performed with close operator monitoring. Tests 5 and 6 are performed while in natural circulations so the loss of forced coolant flow transient does not apply.
However, the RCS will be placed in an off-normal configuration for natural circulation conditions for the performance of these tests and operator action will be required to return the system to a normal configuration should natural circulation conditions degrade to a point
(
past that allowed in the test procedure.
3.1.1.7 Stuck-Out, Stuck-In. or Dropped Control Rod Accident This event results in a primary pressure reduction. Except fc,r Test 5, normal RPS trip settings for high pressure, law pressure and variable low pressure will be in effect through the lower power test program. Power is maintained low and the worth of the dropped rod is less due to the additional boration during the tests.
No increase in probability or severity of this event is introduced by these tests conditions and the consequences of such an event are considered bounded by the FSAR.
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x for Test 5, in order to provide additional operating flexibility and prevent an' inadvertent reactor trip, t.he variable low pressure trip will be bypassed and the low pressure trip reset to 1700 psig. However, constant operator surveillance of reactor power level, reactor-coolant system temperatures and pressure, and subcooling-margin and manual trip of the reactor if any trip limits are. reached will be relied on for some low pressure protection during this test.
3.1.1.8 Loss of Electrical Power Because of the low power level for the test program, the consequences of this accident with respect to the maximum system pressure are bounded by the FSAR analysis.
Because of the low power level and essentially no decay heat load, the consequences of this accident with respect to possible fuel damage are considered bounded by the FSAR analysis provided operator actions are taken to prevent the EFW system from rapidly overfeeding the OTSGs and yet insure that OTSG 1evels of 507. on the Operator Range' are established and then maintained, either manually or 3
automatically.
These actions would results in a primary plant response that is bounced by the FSAR analysis where full decay heat load was considered and thus no operator action to control the EFW system was necessary.
e 3.1.1.9 Steam Line Failure The steam line break event considers the effects f core overcooling, containment pressure response, and fuel failure (DN3) potential.
Plant conditions during testing will assure a greater suberitical margin upon trip due to the high boron concentration. Reactivity feedback due to the potentially large overcooling from the blowdown of the large secondary inventory is minimized by the slightly negative moderator coefficient of reactivity..Due to the increased shutdown margin and moderator coefficient of reactivity no return to power is anticipated following a major steam line break.
Considering the icw superheat available, the containment environment rasponse is bounded by the FSAR analysis. Due to the low power level limits and resulting low heat fluxes, no fuel failure is expected.
Natural circulation flows would tend to retard excessive cooldown.
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Operator surveillance during testing, combined with requirements for manual trip based on conditions described provide added protection.
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~~4 3.l.1.10 Steam Generator Tube Failure The steam generator tube rupture event in its early phase of the transient is bounded by conditions and actions required for a very small LOCA.
Operator actions are required for RCS depressurization and identification and isolation of the affected steam generator.
The instrumentation and equipment required to monitor and perform these functions will remain available during the testing period. Therefore, the ability to mitigate the consequences of a tube leak is not improved by test condi-tions and the consequences calculated in the FSAR analysis remain valid.
3.1.2 Standby Safeguards Analysis 3.1.2.1 Fuel Handling Accident The FSAR analysis of the fuel handling accident is bounding.
System configuration during the test program would preclude the occurrence of this accident.
3.1.2.2 Rod Ejection Accident Control road insertion during low power testing will be limited to the full power rod insertion limits, thereby limiting the ejectable CRA to values lower than those considered in the FSAR analysis. These low ejected rod worths combined with the low RPS high flux trip setpoint provides added protection for any reactivity excursion by initiating an earlier trip.
3.1.2.3 Loss of Coolant Accident For the proposed series of low-power natural circulation tests, initial and expected transient conditions were evaluated to determine whether or not they lie within the bounds of conditions assumed for the hypothetical LOCA's analyzed as described in the TMI-l FSAR, Chapter 14.
Initial conditions for the tests were found to be within bounds of the FSAll analysis. However, as described below, the plant can undergo transients and evolve to conditions during the tests which can make it difficult to detect or control concurrently occurring LOCA's, particularly small LOCA's.
3.1.2.3.1 Large LOCA's (> 0.5 ft Breaks)
Large LOCA's are Inpid, violent events which can easily be detected by the operators, even if the RCS is undergoing transients of the type anticipated during the tests. Initial and transient conditions
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are within the bounds of the assumptions used for FS AR large LOCA analyses.
The ECCS is designed to respond automatically to these accidents.
Immediate operator action is not required to recognize and maintain or restore safe plant conditions but followup operator action is required to place the unit in the long term core cooling mode should a 'large LOCA occur.
2 3.1.2.3.2 Small LOCA's (< 0.5 ft Breaks)
With the plant at initial test r.onditions, but befcre the actual test transient begin, FSAR assumptions and analyses are bom. ding.
Should a small LOCA occur when the tests are underway, FSAR assumptions are also bounding. The transients induced as part of the proposed testing could, in some situations, tend to disguise the early transient behavior of a small LOCA, should such an accident occur cot: current with the test.
However, the existing Small Break Operating Guidelines (SBOG) fully describe all possible symptoms for small LOCA. These guidelines will enable operators to distinguish between expected system behavior resulting from the tests, and system aberrations which could be caused by small LOCA.
For these small breaks which may require operator action to maintain plant safety, the SB0G provide adequate instructions for dealing with all circumstances during which LOCA may occur. This includes the anticipated conditions during this testing.
In addition, special precautions provided to the operator in the test procedures developed for this test program will provide additional safety margins and will specifically direct s
operator'r attention to paraueters which are not only important as far as evaluating the test, but are also important to the detection of possible small LOCA.
3.1.2.4 Maximum Hypothetical Accident The dose consequences of this event are bounded by the FSAR.
3.1.2.5 Waste Gas Tank Rupture Dose consequences of this event are bounded by the FSAR.
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