ML20099D491
| ML20099D491 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 03/08/1985 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20099D448 | List: |
| References | |
| NUDOCS 8503130365 | |
| Download: ML20099D491 (13) | |
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I. TECHNICAL SPECIFICATION CHANGE REQUEST (TSCR) NO. 142 The Licensee requests that the revised pages attached replace the following pages of the existing Technical Specifications:
11, iv, 3-25, 3-26, 3-26a, 3-26b, 4-52, 4-52a II. REASON FOR CHANGE By letter dated July 18, 1984, NRC provided its evaluation of GPJJN's submittal (LIL 341) of January 26, 1984 regarding Decay Heat Removal and requested that GPUN include certain aspects of Standard Technical
, Specifications (STS) into the TMI-1 Technical Specifications or provide justification as to why the requirements are not necessary.
This TSCR provides additional technical specification requirements and clarification for maintaining decay heat removal capability below 250*F reactor coolant temperature as described in GPUN's letter of October 10, 1984, in order to assure redundant or diverse decay heat removal capability without reliance upon administrative requirements or management directives alone.
III. SAFETY EVALUATION JUSTIFYING CHANGE The changes made through this proposed revision incorporate additional requirements and clarification which meet the intent of STS (NUREG-0103, Rev. 4) in order to assure DHR capability for plant conditions with reactor coolant temperature less than 250*F. For this reason, these changes will have a beneficial effect on plant safety. Exceptions or modifications to certain aspects of the STS which NRC has requested to be included in the THI-1 T.S. are justified as follows:
- 1) This change is structured to conform to the TMI-1 T.S. format, applicable to the plant conditions which correspond most closely to STS modes of operation and meets the intent of STS to assure DHR capability.
- 2) Surveillance of RCP operation at power is performed continuously by the Reactor Protection System. Without a RCP in operation in each loop, the RPS will trip the reactor as specified in TMI-1 T.S.
Table 2.3-1. Surveillance requirements for reactor protection system instrumentation are given in TMI-1 T.S. 4.1.1. Procedures for these requirements are specified by THI-1 T.S. 6.8 to be implemented and maintained requiring review and approval prior to implementation and periodic review as set forth in administrative procedures.
Therefore, additional surveillance requirements for verification of RCP operation at power are not included as part of this change.
8503130365 850308 PDR ADOCK 05000209 P PDR
, 3) This change includes surveillance specifications that the means for decay heat removal below 250 F which are required to be operable be verified operable daily. Detailed surveillance procedures are
- required by Specification 6.8 in order to implement these surveillance requirements.
Certain surveillance specifications which are included in STS are not included in this change request. These specifications are unnecessary requirements as described in NUREG-1024, Section 3.9, in that they consume the time and attention of plant personnel that could be used for purposes more important to safety. Those j surveillances which have not been included are as follows:
a) surveillance requirements to verify reactor coolant loops in
] operation every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
i l b) surveillance requirements to verify operability of non-operating 1 reactor coolant pumps every 7 days by verifying the correct breaker alignments and indicated power availability.
c) surveillance requirements to verify that a DHR loop is in operation every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Such requirements would be an unnecessary administrative burden and j these conditions would be obvious to the control room operators.
- 4) This change recognizes heat losses to the Reactor Building atmosphere as an acceptable means of decay heat removal at decay heat generation rates below 188 KW with the RCS full and below 100 KW with the RCS drained down for maintenance (TMI-1 calculation,
'l C3320-85-001). When decay heat load is very low, as with the present plant conditions at TMI-1, heat loss to ambient is l sufficient to provide adequate decay heat removal capability. This cooling method requires no active components but relies upon basic heat transfer principles.
- 5) This change requires equipment to be in operation only when needed to circulate reactor coolant in order to maintain reactor coolant temperature at least 10*F below saturation temperature. While STS
- Sections 3.4.1.3, 3.4.1.4 and 3.9.8.1 require equipment to be in 1
operation allowing only a maximum of one hour down time with certain stipulations applied, this change allows equipment to be secured for longer periods where conditions permit.
, Depending on the decay heat generation rate, shutdown of the forced l circulation equipment may allow reactor coolant system temperature 4
to increase very slowly some time after shutdown compared to the
, conditions ininediately following plant shutdown from power operation. As described in 4 above, if the period of plant shutdown I i is sufficiently long, losses to ambient may be adequate to allow ,
j reactor coolant temperature to stabilize without forced circulation.
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! 6) This change recognizes the acceptability of cooldown by natural
, circulation as an acceptable means of decay heat removal. The
- adequacy of natural circulation cooldown as a stable means of decay heat removal is described in Appendix 1 of BAW-10069, by Babcock and
- Wilcox.
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- 7) This change allows a limited period of up to 7 days for which the requirement for a redundant or diverse means of DHR may be suspended
- in order to provide for preventive or corrective maintenance that l may be necessary to ensure the cont'inued reliability of the l
- preferred means of DhR capability. This provides the assurance also
- that appropriate action will be taken to restore the preferred means j of DHR capability to operable status in a timely manner, without prohibiting maintenance which is needed to decrease the likelihood of actual in-service failures. A period of up to 7 days is i
! justified considering'the low probability and minimal consequences of such a system failure.
j 8) This change recognizes that the DHR system is not the only system
! capable of providing a flow of borated cooling water through the reactor vessel during cold shutdown and refueling. In addition to the ' equipment allowed by STS in fulfilling the requirements for DHR capability, this change allows the use of a flow path from the BWST
! with BWST level greater than 44 ft. as an alternate flow path whenever such means are determined to be capable of maintaining RCS temperature at least 10 F below saturation. With such a volume of cooling water available for circulation through the reactor vessel as the only source of cooling, DHR capability can be maintained for i
a period of several weeks. The length of time such an alternate flow path would be allowable in lieu of the operability of the redundant primary means of DHR is predictable using calculations based on actual plant data or through plant testing at the time the system is declared operable.
Surveillance requirements are included to verify the operability of the circulating path daily whenever the flow p'ath is required to be operable.
These changes as discussed above are justified in that the requirements
! embodied in this TSCR provide for continuous decay heat removal
- capability, provide additional guidance, and specify a level of i redundancy while allowing systems to be taken out of service for proper maintenance to be performed.
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IV. NO SIGNIFICANT HAZARDS CONSIDERATIONS These proposed changes provide additional operational requirements to assure redundant or diverse decay heat removal capability. Additional limiting conditions for operation and additional surveillance requirements are included. Therefore, operation of TMI-1 in accordance with this TSCR:
- 1) does not involve a significant increase in the probability or consequences of an accident previously evaluated,
- 2) does not create the possibility of a new or different kind of accident from any accident previously evaluated, and
- 3) does not involve a significant reduction in a margin of safety.
Therefore, significant safety hazards are not associated with this change.
V. IMPLEMENTATION It is requested that the amendment authorizing this change become effective 120 days after receipt, to allow for the necessary procedural revisions to be put in place.
VI. AMENDMENT FEE (10 CFR 170.21)
Pursuant to the provisions of 10 CFR 170.21, attached is a check for
$150.00.
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S:ction TABLE OF CONTENTS Pag 3 2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2-1 2.1 Safety Limits, Reactor Core 2-1 2.2 Safety Limits, Reactor System Pressure 2-4 ,
2.3 Limiting Safety System Settinor, Protection 2-5 Instrumentation
- 3 LIMITING CONDITIONS FOR OPERATION 3-1 3.0 General Action Requirements 3-1 3.1 Reactor Coolant System 3-la 3.1.1 Operational Components 3-la 3.1.2 Pressurization, Heatup and Cooldown Limitations 3-3 3.1.3 Minimum Conditions for Criticality 3-6 3.1.4 Reactor Coolant System Activity 3-8 3.1.5 Chemistry 3-10 3.1.6 Leakage 3-12 3.1.7 Moderator Temperature Coefficient of Reactivity 3-16 3.1.8 Single Loop Restrictions 3-17 3.1.9 Low Power Physics Testing Restrictions 3-18 3.1.10 Control Rod Operation 3-18a 3.1.11 Reactor Internal Vent Valves 3-18b 3.1.12' Pressurizer Power Operated Relief Valve (PORV) 3-18c and Block Valve 3.1.13 Reactor Coolant System Vents 3-18f 3.2 Makeup and Purification and Chemical Addition Systems 3-19 3.3 Emergency Core Cooling, Reactor Building Emergency 3-21 Cooling and Reactor Building Spray Systems 3.4 Decay Heat Removal Capability 3-25 3.4.1 Reactor Coolant System Temperature Greater 3-25 Than 250*F 3.4.2 Reactor Coolant System Temperature 250*F or Less 3-26 3.5 Instrumentation Systems 3-27 3.5.1 Operational Safety Instrumentation 3-27 3.5.2 Control Rod Group and Power Distribution Limits 3-33 3.5.3 Engineered Safeguards Protection System Actuation 3-37 Setpoints 3.5.4 Incore Instrumentation 3-38
- 3.5.5 Accident Monitoring Instrumentation 3-40a 3.6 Reactor Building 3-41 3.7 Unit Electrical Power System 3-42 3.8 Fuel Loading and Refueling 3-44 3.9 Radioactive Materials 3-46 3.10 Miscellaneous Radioactive Materials Sources 3-46 3.11 Handling of Irradiated Fuel 3-55
- 3.12 Reactor Building Polar Crane 3-57 3.13 Secondary System Activity 3-58 3.14 Flood 3-59 3.14.1 Periodic Inspection of the Dikes Around TMI 3-59 3.14.2 Flood Condition for Placing the Unit in Hot Standby 3-60 3.15 Air Treatment Systems 3-61 3.15.1 Emergency Control Room Air Treatment System 3-61 3.15.2 Reactor Building Purge Air Treatment System 3-62a 3.15.3 Auxiliary and Fuel Handling Exhaust Air Treatment 3-62c Amendment No. $ , Jd, 7/, Sf, 98 11
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- S3ction TABLE OF CONTENTS _Page j 4.7 Reactor Control Rod System Tests - 4-48 2
4.7.1 Control Rod Drive System Functional Tests 4-48 i 4.7.2. Control Rod Program Verification 4-50 i i 4.8 Main Steam Isolation Valves 4-51 l 4.9 Decay Heat Removal Capability - Periodic Test 4-52 4.9.1 Emergency Feedwater System - Periodic Testing 4-52
- (Reactor Coolant Temperature Greater Than 250*F) 4.9.2 Decay Heat Removal Capability - Periodic Testing 4-52a
! (Reactor Coolant Temperature 250*F or Less)
- 4.9.3 Acceptance Criteria 4-52a
- 4.10 Reactivity Anomalies 4-53
, 4.11 Reactor Coolant System Vents 4-54
! 4.12 Air Treatment Systems 4-55 i 4.12.1 Emergency Control Room Air Treatment System 4-55 i 4.12.2 Reactor Building Purge Air Treatment System 4-55b j 4.12.3 Auxiliary & Fuel Handling Exhaust Air Treatment System 4-55d j 4.13 Radioactive Materials Sources Surveillance 4-56 i 4.14 Reactor Building Purge Exhaust System 4-57 1
4.15 Main Steam System Inservice Inspection 4-58 i 4.16 Reactor Internals Vent Valves Surveillance 4-59 l 4.17 Shock Suppressors (Snubbers) 4-60
! 4.18 Fire Protection Systems .
4-72 1
4.18.1 Fire Protection Instruments 4-72 4.18.2 Fire Suppression Water System 4-73
- 4.18.3 Deluge / Sprinkler System 4-74 ;
j 4.18.4 CO2 System 4-74 :
j 4.18.5 HaTon Systems 4-75 l 4.18.6 Hose Stations 4-76
- 4.19 OTSG Tube Inservice Inspection 4-77 4.19.1 Steam Generator Sample Selection & Inspection Methods 4-77 l 4.19.2 Steam Generator Tube Sample Selection & Inspection 4-77
! 4.19.3 Inspection Frequencies 4-79 l 4.19.4 Acc.eptance Criteria 4-80
! 4.19.5 Reports 4-81 4 4.20 Reactor Building Air Temperature 4-86 1
4.21.1 Radioactive Liquid Effluent Instrumentation 4-87
- 4.21.2 Radioactive Gaseous Process & Effluent Monitoring 4-90 l Instrumentation j 4.22'.1.1 Liquid Efflu sts 4-97 ,
- 4.22.1.2 Dose 4-102 j 4.22.1.3 Liquid Waste Treatment 4-103 l 4.22.1.4 Liquid Holdup Tanks 4-104 i 4.22.2.1 Dose Rate 4-105 i 4.22.2.2 Dose. Noble Gas 4-110
, 4.22.2.3 Dose, Radiciodines, Radioactive Material in Particulate 4-111 ;
l Form & Radionuclides Other Than Noble Gases j 4.22.2.4 Gaseous Radwaste Treatment 4-112 I j 4.22.2.5 Explosive Gas Mixture 4-113 4.22.2.6 Gas Storage Tanks 4-114 !
4.22.3.1 Solid Radioactive Waste 4-115
- 4.22.4 Total Dose 4-116 i 4.23.1 Monitoring Program 4-117 i 4.23.2 Land Use Census 4-121 j 4.23.3 Interlaboratory Comparison Program 4-122 f Amendment No. g )6, )6 f!', f7',%J6J6',)MI, 97 4
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3.4 DECAY HEAT REMOVAL C/@ ABILITY Applicability
, Applies to the operating status of systems and components that function to remove decay heat when one or more fuel bundles are located in the reactor
- vessel.
Objective To define the conditions necessary to assure continuous capability of decay
! heat removal **
Specification 4
3.4.1 Reactor Coolant System temperature greater than 250*F.
3.4.1.1 With the Reactor Coolant System temperature greater than 250'F, three independent EFW pumps and associated flow paths shall be OPERABLE with:
- a. Two EFW pumps, each capable of being powered from an OPERABLE emergency bus, and one EFW pump capable of being powered from an OPERABLE steam supply system. Specification 3.0.1 applies.
- b. With one pump or flow path
- inoperable, restore the inoperable pump
- or flow path to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in COLD i SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. With more than one EFW pump or flow path
- inoperable, restore the inoperable pumps or flow paths
- to OPERABLE status or be subcritical within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in COLD SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
l c. Four of six turbine bypass valves OPERABLE.
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- d. The condensate storage tanks (CST) OPERABLE with a minimum of 150,000 gallons of condensate available in each CST. With a CST inoperable, i restore the CST to operability within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. With more than one CST inoperable, restore the inoperable CST to OPERABLE status or be subcritical within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in at least l
HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in COLD SHUTDOWN within the
- following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Specification 3.0.1 applies.
' *For the purpose of this requirement, an OPERABLE flow path shall mean an unobstructed path from the water source to the pump and from the pump to a steam generator.
- **These requirements supplement the requirements of Sections 3.1.1.1.c.
[ 3.1.1.2, 3.3.1 and 3.8.3.
l 3-25 AmendmentNo./,7/,98 h
3.4.1.2 With the Reactor Coolant System temperature greater than 250*F, all l eighteen (18) main steam safety valves shall be OPERABLE or, if any are not OPERABLE, the maximum overpower trip setpoint (see Table 2.3-1) shall be reset as follows:
Maximum Number of Maximum Overpower Safety Valves Disabled on Trip Setpoint Any Steam Generator (% of Rated Power) 1 92.4 2 79.4 3 66.3 With more than 3 main steam safety valves inoperable, restore at least fifteen (15) main steam safety valves to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
3.4.2 Reactor Coolant System temperature 250*F or less.
3.4.2.1 With Reactor Coolant temperature 250*F or less, at least two of the following means for maintaining decay heat removal capability shall be OPERABLE and at least one shall be in operation except as allowed by Specifications 3.4.2.2, 3.4.2.3 and 3.4.2.4.
- a. Decay Heat Removal String "A".
- b. Decay Heat Removal String "B".
- c. Reactor Coolant Loop "A", its associated 0TSG, and its associated main or emergency feedwater flowpath,
- d. Reactor Coolant Loop "B", its associated OTSG, and its associated
! main or emergency feedwater flowpath.
3.4.2.2 Operation of the means for decay heat removal may be suspended provided the core outlet temperature is maintained at least 10*F below saturation temperature.
3.4.2.3 The number of required means for decay heat removal as specified by 3.4.2.1 may be reduced to one provided that one of the following conditions is satisfied:
- a. The Reactor is in a Refueling Shutdown condition with the Fuel Transfer Canal water level greater than 23 feet above the reactor vessel flange.
- b. The BWST level is greater than 44 feet with an associated flow path through the RCS OPERABLE such that the core outlet temperature can be maintained at least 10*F below saturation temperature for at least 7 days.
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3-26 Amendment No. 4, 78
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- c. Equipment Maintenance on one of the means for decay heat removal '
specified by 3.4.2.1 is required and the equipment outage does
. not exceed 7 days.
3.4.2.4 Specification 3.4.2.1 does not apply when either of the following conditions exist:
- a. Decay heat generation is less than 188 KW with the RCS full.
- b. Decay heat generation is less than 100 KW with the RCS drained down for maintenance.
3.4.2.5 With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible.
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3-26a
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j Bases j A reactor shutdown following power operation requires removal of core decay i
heat. Normal decay heat removal is by the steam generators with the steam
, dump to the condenser when RCS temperature is above 250*F and by the decay
! heat removal system below 250'F. Core decay heat can be continuously
- dissipated up to 15 percent of full power via the steam bypass to the
! condenser as feedwater in the steam generator is converted to steam by heat
! absorption.- Normally, the capability to return feedwater flow to the steam l generators is provided by the main feedwater system.
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! The main steam safety valves will be able to relieve to atmosphere the total I steam flow if necessary. If Main Steam Safety Valves are inoperable, the
! power level must be reduced, as stated in Technical Specification 3.4.1.2 such
} that the remaining safety valves can prevent overpressure on a turbine trip.
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In the unlikely event of complete loss of off-site electrical power to the i station, decay heat rerreal is by either the steam-driven emergency feedwater i
pump, or two half-sized motor-driven pumps. Steam discharge is to the
- atmosphere via the main steam safety valves and controlled atmospheric relief
, valves, and in the case of the turbine driven pump, from the turbine exhaust.(1) i' 1
Both motor-driven pumps are required initially to remove decay heat with one eventually sufficing. The minimum amount of water in the condensate storage tanks, contained in Technical Specification 3.4.1.1, will allow cooldown to l
l 250*F with steam being discharged to the atmosphere. After cooling to 250'F.
the decay heat removal system is used to achieve further cooling.
l When the RCS is below 250'F, a single DHR string, or single OTSG and its
! associated main or erergency feedwater flowpath is sufficient to provide removal of decay heat at all times following the cooldown to 250*F. Tne
- requirement to maintain two OPERABLE means of decay heat removal ensures that
- a single failure does not result in a complete loss of decay heat removal i capability. The requirement to keep a system in operation as necessary to I
maintain a 10*F subcooling margin at the core outlet provides the guidance to ensure that steam conditions which.could inhibit core cooling do not occur.
} Limited reduction in redundancy is allowed for preventive or corrective j maintenance on 'the primary means for decay heat removal to ensure that j maintenance necessary to assure the continued reliability of the systems may l be accomplished.
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! As decay heat loads are reduced through decay time or fuel off loading, i alternate flow paths will providc cdequate cooling for a time sufficient to
! take compensatory action if the normal means of heat removal is lost. At such i times, removal of the redundant or diverse cooling system is permitted.
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.M 3-26b m
Following extensive outages or major core off loading, the decay heat generation being removed from the Reactor Vessel is so low that ambient losses are sufficient to maintain core cooling and no other means of heat removal is required. The system is passive and requires no redundant or diverse backup system. Decay heat geheration is calculated in accordance with ANSI 5.1-1979 to determine when this situation exists.
An unlimited emergency feedwater supply is available from the river via either of the two motor-driven reactor building emergency cooling water pumps for an indefinite period of time.
i The requirements of Technical Specification 3.4.1.1 assure that before the reactor is heated to above 250'F, adequate auxiliary feedwater capability is available. One turbine driven pump full capacity (920 gpm) and the two half-capacity motor-driven pumps (460 gpm each) are specified. However, only one half-capacity motor-driven pump is necessary to supply auxiliary feedwater flow to the steam generators in the onset of a small break loss-of-coolant accident.
! REFERENCES (1) FSAR Section 10.2.1.3.
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I 3-26c Amendment No. 78 1
1 4.9 DECAY HEAT REMOVAL CAPABILITY - PERIODIC TESTING Applicability Applies to the periodic testing of systems or components which function to remove decay heat.
Objective ~
- To verify that systems / components required for decay heat removal are capable of performing their design function.
Specification 4.9.1 Emergency Feedwater System - Periodic Testing (Reactor Coolant System Temperature greater than 250*F.)
4.9.1.1 Whenever the Reactor Coolant System temperature is greater than 250*F, the EFW pumps shall be tested in the recirculation mode in accordance with the requirements and acceptance criteria of ASME Section XI Article IWP-3210. The test frequency shall be at least every 31 days of plant operation at Reactor Coolant Temperature above ,
250*F.
! 4.9.1.2 During testing of the EFW System when the reactor is in STARTUP, HOT STANDBY or POWER OPERATION, if one steam generator flow path
- is made inoperable, a dedicated qualified individual who is in communication with the control room shall be continuously stationed at the EFW local manual valves (See Table 4.9-1). On instruction from the Contt01 Room Operator, the individual shall realign the valves from the test mode to their operational alignment.
4.9.1.3 At least once per 31 days each valve listed in Table 4.9-1 shall be verified to be in the status specified in Table 4.9-1, when required to be OPERABLE.
4.9.1.4 On a quarterly basis, verify that the manual control (HIC-849/850) valve station functions properly.
4.9.1.5 On a quarterly basis, EFY-30A and B shall be checked for proper j operation by cycling each valve over its full stroke.
1 4.9.1.6 Prior to start-up, following a refueling shutdown or a cold shutdown greater than 30 days, conduct a test to demonstrate that the motor driven EFW pumps can pump water from the condensate tanks to the Steam Generators.
- For the purpose of this requirement, an OPERABLE flow path shall mean an unobstructed path from the water source to the pump and from the pump to a Steam Generator.
Amendment No. 78 l 1
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j 4.9.1.7 Acceptance Criteria 3
l I These tests shall be considered satisfactory if control board i
! indication and visual observation of the equipment demonstrates that j all components have operated properly except for the tests required
- by Specification 4.9.1.1.
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4.9.2 Decay Heat Removal Capability - Periodic Testing (Reactor Coolant
! System Temperature 250'F or less).*
i-4.9.2.1 On a daily basis, verify operability of the means for decay heat l removal required by specification 3.4.2 by observation of console
,. status indication.
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l *These requirements supplement the requirements of 4.5.2.2 and 4.5.4. l I
Bases The 31-day testing frequency will be sufficient to verify that the turbine i driven and two motor-driven EFW pumps are operable and that the associated j valves are in the correct alignment. ASME Section XI. Article IWP-3210 :
j specifies requirements and acceptance standards for the testing of nuclear !
safety related pumps. Compliance with the normal acceptance criteria assures 4
! that the EFW pumps are operating as expected. The test frequency of 31 days
- (nominal) has been demonstrated by the B&W Emergency Feedwater Reliability i Study to assure an appropriate level of reliability. In the case of the EFW -
- System flow, the flow shall be considered acceptable if under the worst case single pump failure, a minimum of 500 gpm can be delivered when steam ,
generator pressure is 1050 psig and one steam generator is isolated. A flow j of 500 gpm, at 1050 psig head, ensures that sufficient flow can be delivered :
- to either Steam Generator. The surveillance requirements ensure that the overall EFW System functional capability is maintained.
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4 Daily verification of the operability of the required means for decay heat
! removal ensures that sufficient decay heat removal capability will be j maintained. ;
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l 4-52a j Amendment No. 78 l
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