ML20206Q963
| ML20206Q963 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 04/17/1987 |
| From: | Domer J TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8704220072 | |
| Download: ML20206Q963 (37) | |
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' TENNESSEE VALLEY ' AUTHORITY CHATTANOOr.A. TENNESSEE 37401 SN 157B Lookout Place APR 171987 TVA-SQN-TS-87-14 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:
In the Matter of ) Docket Nos. 50-327 Tennessee Valley Authority ) 50-328 SEQUOY4H NUCLEAR PLANT - TECHNICAL SPECIFICATION BASES CHANGE 87-14 As a part of the Sequoyah restart effort, TVA has evaluated the accuracy and clarity of the bases portion af the unit 1 and unit 2 technical specifications. As a result of this evaluation, TVA revised a number of the bases.. The revised bases are being provided for your information and for ,
inclusion in your copy of the Sequoyah unit 1 and unit 2 technical specifications.
The revised bases are identified in enclosure 1. The description and justification for these changes are provided in enclosure 2. A determination of no significant hazards is provided in enclosure 3.
Please direct questions you may have concerning this issue to M. J. Burzynski at (615) 870-6172.
Very truly yours, TENNESSEE VALLEY AUTHORITY I
. .M J. A. Domer, Assistant Director Nuclear Safety and Licensing Sworn g d subscr e b ore me
. th s - /7 day of 1987
)) h
' Notary Public ,
My Conunission Expires ,
Enclosures cc: see page 2 8704220072 870417 7 8 PDR ADOCK 0500 P
An Equal Opportunity Employer
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U.S. Nuclear Regulatory Conunission -
APR 171987 cc (Enclosures):
Mr. Michael H. Mobley, Director Division of Radiological Health T.E.R.R.A. Building 150 9th Avenue, N Nashville, TN 37203 Mr. G. G. Zech, Assistant Director for Inspection Programs Division of TVA Projects Office of Special Projects U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Sequoyah Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee 37379
l ENCLOSURE 1 ,
1 TECHNICAL SPECIFICATION BASES CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-87-14)
REVISION OF MISCELLANEOUS BASES ,
LIST OF AFFECTED PAGES Unit 1
.i B 2-2 B 3/4 1-3 B 3/4 6-4 B 2-3 B 3/4 3-3 B 3/4 6-Aa B 2-6 B 3/4 5-3 B 3/4 6-5 B 2-7 B 3/4 6-2 B 3/4 7-3 Unit 2 B 2-1 B 3/4 1-3 B 3/4 6-3 B 2-3 B 3/4 3-3 B-3/4 6-4 B 2-6 B 3/4 5-3 B 3/4 6-5 B 2-7 B 3/4 6-2 B 3/4 7-3 i
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SAFETY LIMITS BASES These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the I D el+o."T f) (Delta I) function of the Overtemperature4trip. When the axial power j imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature Delta T trips will reduce the setpoints to previde protection consistent with core safety limits.
2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Reactor Coolant System y) .-
piping, valves and fittings, are designed to ANS! B 31.1 1967 Edition, which permits a maximum transient pressure of 120% (2985 psig) of component design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.
The entire Reactor Coolant System is hydrotested at 3107 psig,125% of design pressure, to demonstrate integrity prior to initial operation.
2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit. The Trip Setpoi'nts have been selected to ensure that the reactor core and reactor coolant system are prevelted from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.
Operation with a trip set less conservative than its Trip Setpoint but within -
its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift' allowance assumed for each trip in the safety analyses, y M SEQUOYAH - UNIT 1 B 2-2 L l .
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. SAFETY LIMITS
'e BASES Manual Reactor Trip'
. -- The Manual Reactor Trip is a redundant channel to the automatic protective !
instrumentation channels and provides manual reactor trip capability.
Power Range, Neutron Flux . . .
h The Power Range, Neutron Flux channel high setpoint provides reactor core protection against reactivity excursions which are too' rapid to be protected by temperature and pressure protective circuitry. -The low set point provides i
- redundant protection in the power range for a power excursion beginning ,from ._ _
low power. The trip associated with the low setpoint may be manually bypassed when P-10 is active (two of_the four power range channels indicate a power I
'lo! . leven oT aoeve approximatelyIF percent of RATED THERMAL POWER) and is auto-matically .roinstated when P-10 becomes inactive (three of the four channels
- indicate a power level below approximately 9 percent of RATED THERMAL POWER).
Power Range Neutron Flux, High Rates .; _
- The P6wer Range Positive Rate trip provides protection against rapid flux-i increases which are characteristic of rod ejection events from any power level.
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- Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power.
'T The Power Range Negative Rate trip provides protection to ensure that the
- minimum DNBR is maintained above 1.30 for control rod dros accidents. At high 1 power a single or multiple rod drop accident could cause local- flux peaking which,
! when in conjunction with nuclear power being maintained equivalent to turbine power by action of the automatic rod control system, could cause an unconserva-tive' local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor for all single or multiple dropped rods.
Intermediate and Source Range, Nuclear Flux -
The' Intermediate and Source Range, Nuclear Flux trips provide reactor core protection during reactor startup. These trips provide redundant protec i .;. ' . '
. . tion to the low setpoint trip of the Power Range, Neutron Flux channels. The
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second'u'nless manually blocked when P-6 becomes active. The Intermediate r.y-y^. . . ' - ,,,:,
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BASES Steam Generator Water level The Steam Generator Water Level Low-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays of the auxiliary feedwater system.
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Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level
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The Steam /Feedwater Flow Mismatch in coincidence with a Steam Generator Low Water Level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified
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trip settings and thereby enhance the overall reliability of the Reactor Protection System. This trip is redundant to the Steam Generator Water Level Low-Low trip. The Steam /Feedwater Flow Mismatch portion of this trip is
- activatedwhenthgsteamflowexceedsthefeedwaterflowbygreaterthanor equal to 1.5 x 10 lbs/ hour. The Steam Generator low Water level portion of the trip is activated when the water level drops below 24 percent, as indicated G. ;
by the narrow range instrument. These trip values include sufficient allowance b"
-in excess of normal operating values to preclude spurious trips but will initiate a reactor trip before the steam generators are dry. Therefore, the required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient on the Reactor Coolant System and steam generators is minimized.
Undervoltace and Underfrecuency - Reactor Coolant Pumo Busses The Undervoltage and Underfrequency Reactor Coolant Pump bus trips provide
- reactor core protection against DNB as a result of loss of voltage or under-frequency to more than one reactor coolant pump. The specified set points '
.-reached.
assure a reactor trip signal is generated before the low flow trip set point is
- Time delays are incorporated in the underfrequency and undervoltage
- - trips to prevent spurious reactor trips from momentary electrical power transients.
.For undervoltage, the delay.is set so that the time required for a signal to reach the reactor trip breakers following the simultaneous trip _of two or more f g, g reactor coolant pump bus circuit breakers shall not exceed.O T seconds. For
( underfrequency,.the delay is set so that the time. required for a signal to reach the reactor trip breakers 'after the underfrequency trip 'set point is reached
,'shall not exceed Off seconds.
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BASES ' . Turbine Trip. 9 1 ATurbineTripcausesadirectreactortripwhenoperatingaboveP-I Each the ensuing of the transient. turbine trips provide turbine protection and reduce the severity o . tion of these trips. No credit was taken in the accident analyses for opera- { is required to enhance the overall reliability of the Reactor ProtectiTh on System. Safety Infection Input from ESF instrumentation, the ESF automatic actuation logic channels r reactor trip upon any signal which initiates ~a safety injection. This trip is provided to protect the core in the event of a LOCA. channels which initiate a safety injection signal are shown in Table 3.3-3The ESF instr .
' Reactor Trip System Interlocks increasing power:The Reactor Trip System Interlocks perform the following functions on c P-6 Enables the manual block of the source range reactor trip (i.e.,
prevents premature block of source range trip). P-7 Defeats the automatic block of reactor trip on: ; Low flow in more P-13 than one primary coolant loop, reactor coolant pump undervoltage { and underfrequency, pressurizer high level.turbine trip, pressurizer low pressure, and *I
,I P-8 t Defeats in a single the automatic block of reactor trip on low RCS coolant flow loop. L I.
P-9 k Defeats the automatic block of Reactor Trip on Turbine Trip i g31 ,
.y P-10 Enables the manual block of reactor trip on power range (low setpoint) ,
s diate functions). range rod stops (i.e., prevents premature - block W. of th y% g On decreasing power, the opposite function is performed at reset setpoints ~
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O.:, .b P-4 on _ ' Reactor T tripped - Actuates turbine trip, closes' main feedwater valves
' valvDU below setpoint, prevents the opening of the main feedwater bl' water level signal, allows manual block of the automatic reactu i' of safety injection. n '
[;t W c. Reactor notoftripped - defeats manual block preventing automatic
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SEQUOYAH - UNIT 1 . Araendment 7 4 B 2-7 6/26/81 yl
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4 REACTIVITY CONTROL SYSTEMS i-BASES 5408 gallons of 20,000 ppe borated water from the boric acid storage tanks or i 64,160 gallons of 2000 ppe borated water from the refueling water storage tank. With the RCS temperature below 200*F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE j ~ ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable. , 4
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The boron capability required below 200*F, is sufficient to provide a SHUTDOWN MARGIN of 1% delta k/k after xenon decay and cooldown from 200*F to 140'F. This condition' requires either 835 gallons of 20,000 ppe borated water from the boric acid storage tanks or 9,690 gallons of 2000 ppe borated water from the refueling water storage tank. g The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.
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The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between.erS*and M 1 for the solution recirculated # withincontainmentafteraLOCA.fThispH and minimizes the evolution of iodine and minimizes the effect oN:bloride and caustic stress corrosion on
. . mechanical systems and components. '
L. 9T ^ i The OPERASILITY of one boron inj.1.rection system during REFUELING ensures that this system is available for reactivity control while in MODE 6. 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distri-bution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod misalignment on associated accident i analyses. OPERABILITY of the control rod position indicators is required to -
- ' determine control rod positions and thereby ensure compliance with the control 4
rod alignment and insertion limits. l .~ SEQUOYAH - UNIT 1 , B 3/4 1-3 l
1 1 m INSTRUMENTATION 1 BASES l design basis for the facility to determine if plant shutdown is required pursuant to Appendix "A" of 10 CFR Part 100. This instrumentation is consistent with the recommendations of Regulatory Guide 1.12, " Instrumentation for Earthquakes," April 1974. 3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION
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The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need
* ~ ~ for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972.
3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION
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TheOPERABILITYoftheremoteshutdownin[trumentationensuresthat sufficient capability is available to permit shutdown and maintenance of HOT _Q STANDBY of the facilit from locations outside of the control room. This W capability is' required n the event control room habitability is lost and is consistent with Genera Design Criteria 19 of 10 CFR 50. 4 Ar pded; \ apd'.ld) fan 5 d 5E P I 3/4.3.3.6 CHLORINE DETECTION SYSTEMS Call 5k41** The OPERABILITY of the chlorine detection system ensures that sufficient capibility is available to promptly detect and initiate protective action in the event of an accidental chlorine release. This capability is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release," February 1975. , 3/4.3.3.7 ACCIDENT MONITORING INSTRUMENTATION il t , The OPERABILITY of the accident monitoring instrum::ntation ensures,th'at U sufficient information is available on selected plant' parameters to monitor and assess these variables following an accident. This cap ~ ability is ' consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light + Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Follcwing d an Accident," December 1975. l
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1 3/4.5.5 REFUCLING WATER STORAGE TANK The OPERABILITY of the RWST as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWST minimum volume and baron concentration ensure that
- 1) sufficient water is available within containment to permit recirculation cooling flow to the core, and 2) the reactor will remain suberitical in the cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses.
The contained water volume limit includes an allowance for water not usuable because of tank discharge line location or other physical characteristics. 7.s 95 The limits on contained wate olume and boron concentration of the RWST 8210 ensure a'pH value of between and . for the solution recirculated si;$1n containment after a LOCA. This pH band minimizes the evolution of icGene and cechanical minimizes systems the effect af chloride and caustic stress corrosion on and components. 5 wh 4 i . ,s SEQUOYAH - UNIT 1 - B 3/4 5-3
CONTAlfe8ENT SYSTEMS BASES i
,% A \s idene d containment peak pressure does not exceed the dee+ge pressure of 12 psig during LOCA conditions. A 7: L i v p :k ;rt::ur; c;;;;ted t: M cht:in d 'r '"Cf :;::: ':
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3/4.6.1.5 AIR TEMPERATURE The linitations on containment average air temoerature ensure that 1) the containmnt air nass is limited to an initial nass sufficiently low to prevent W **"'Th exceedino thefde+++a cressure durino LOCA conditions and 2) the ambient air
*bbj temperature does not exceed that temperature allowable for the continuous duty ratino specified for ecuipment and instrumentation located within containment.
The containant pressure transient is sensitive to the initially contained air mass during a LOCA. The contained air mass increases with decreasing temperature. The lower temperature limits of 100*F for the lower conoartnent, P5"F for the uoper comoartnent, and 60*F when less than or eoual to 5% of PATED THEPf'AL POWEP will limit the peak pressure top' .C ;;i; "ir
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s ture limit influences the peak accident tercerature slichtly durino a LOCA; however, this limit is hased primarily upon eouiDrent protection and anticipated operating conditions. Both the upper and lower tercerature limits are consistent with the parameters used in the accident analyses. 3/4.6.1.6 CONTAINPENT VESSEL STPUCTttpAL INTEGRITY This liritation ensures that the structural intecrity of the containment steel vessel will be maintained comparable to the original design standards for the life of the facility. Structural inteority is recuired to ensure that the vessel will withstand the maxirnun pressure of H ra losic in the event of a 1f LOCA. A visual insoection in conjunction with Type a leakace tests is sufficient to dermnstrate this capability. 3/4.4.1.7 SHIELP PUILDING eTottCTitPAL INTECDITY
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This limitation ensures that the structural intecrity of the containrent shield buildina will be raintained comparable to the oricinal desian standards for the li'e of the facility. Structural intecrity is recuired to orovide li orotection for the steel vessel from external rnissiles, 2I radiation shieldino in the event of a LOCA, and 3) and annulus surroundino the steel vessel that can be raintained at a neoative pressure durino accident conditions, m CEPI'OYAH - UNIT 1 A 3/4 6-2 L
CONTAINMENT SYSTEMS BASES - 3/4.6.4 o COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to
,c:C.'Ub6 maintain limit during thepost-LOCA hydrogen concentration conditions. within containment below its flammable -( np 4' g;g Qgmgy as wa$ gia capable of controlling the expected hydrogen generation associated p
zirconium-water mactions, 2) radiolytic decomposition of-water and 3) corro-slon of metals within containment. s=; 4 geb t Of These hydrogen control systems are d es,p d 4 =, _ Regulatory-Guid6 1.7, " Control 4of d {gsts es sa C o.c.S d7 d u I I 4_ombustible Gas Concentrations in Containment Following"---2 a LOCA", rewW ea 'Z.
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- The hydrogen mixing systems are provided to ensure adequate mixing of the flo.eth containment atmosphere following a LOCA. This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit. gyg g 3/4.6.5 ICE CONDENSER E AJEeT' g g.L The requirements associated with each of the components of'the ice con-denser pressureensure suppression that the overall system capability to will be available to provide sufficient sient to less than 12 psig during LOCA conditions. limit the containment peak pressure tran- . -
3/4.6.5.1 ICE BED -
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M-The OPERABILITY of the ice bed ensures that the requihed ice inventory will 1) be distributed evenly through the containment bays, 2) contain suffi-cient boron to preclude dilution of the containment sump following the LOCA and 3) contain suffic,ient heat mmoval capability to condense the reactor system volume mleased during a LOCA. the assumptions used in the accident analyses.These conditions are consistent with The minimum weight figure of 1200 pounds of ice per basket contains a 10% R7 conservative allowance for ice loss through sublimation which is a factor of . 10 higher than assumed for the ice condenser design. The minimum weight figure of 2,333,100 pounds of ice also contains an additional.1% conservative allowance to account for systematic error in weighing instruments. .In the . R7 event that observed sublimation rates are equal to or lower than design predic-tions after adjusted downward. three years of operation, the minimum ice baskets weight may be In addition,' the number of ice baskets' required to be weighed each 9 months may be reduced after 3 years of. operation if'such a reduction is supported by. observed sublimation data. i 3/4.6.5.2 ICE BED TEMPERATURE MONITORING SYSTEM The OPERABILITY of the ice bed temperature monitoring system ensures that
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the capability is available for. monitoring- the' ice temperature. In the event the monitoring system is inoperable, the ACTION requirements provide assurance that the time limits. ice bed heat removal'eapacity will be retained within the sp . e cified SEQUOYAH - UNIT 1 . B 3/4 6-4 Amendment 4 3/6/81 - .- -
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INSERT to section 3/4.6.4 The OPERABILITY of at least 66 of 68 ignitors in the hydrogen mitigation system will maintain an effective coverage throughout the containment. This system of ignitors will initiate combustion of any significant amount of hydrogen released after a degraded core accident. This system is to ensure burning in a controlled manner as the hydrogen is released instead of allowing it to be ignited at high concentrations by a random ignition source.
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p CONTAINMENT SYSTEMS U BASES M se b 3/4.6.5.3 ICE CONDENSER DOORS sSY TL_ nnenASTt TTV .s &L. f. .._J_.,.. Jan., ..A eka +ks6 6kau
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"^d Cl^C^d 0" Cur"C th2t thO rO20 tor COOlO"t Oy-tC; fluid r;l;;;;i dur' ; : LCC." u 5: di;;rt:d thr ;h th: i:: ;r.d;n;;r b;y; f;r h;;t .;;;.;
a..d t?.;; ex;;;;iV; ;;bli;; tier. Of th; i;; b;d Will r.;t ;- ;r b^;;; ; ;f ;;r-2i" # 7tru^* 7. 3/4.6.5.4 INLET 000R POSITION MONITORING SYSTEM The OPERABILITY of the inlet door position monitoring system ensures that the capability is available for monitoring the individual inlet door position. In the event the monitoring system is inoperable, the ACTION requirements provide assurance that the ice bed heat removal capacity will be retained within the specified time limits. 3/4.6.5.5 DIVIDER BARRIER PERSONNEL ACCESS DOORS AND EQUIPMENT HATCHES (} A _/ s The requirements for the divider barrier personnsi dccess doors and equipment hatches being closed and OPERABLE ensure that a minimum bypass steam flow will occur from the lower to the upper containment compartments during a LOCA. This condition ensures a diversion of the steam through the ice condenser bays that is consistent with the LOCA analyses. 3/4.6.5.6 CONTAINMENT AIR RETURN FANS The OPERABILITY of the containment air return fans ensures that following a LOCA 1) the containment atmosphere is circulated for cooling by the spray system and 2) the accumulation of hydrogen in localized portions of the contain-ment structure is minimized. 3/4.6.5.7 and 3/4.6.5.8 FLOOR AND REFUELING CANAL DRAINS The OPERABILITY of the ice condenser floor and refueling canal drains ~ ensures that following a LOCA, the water from the melted ice and containment spray system has access for drainage back to the containment lower compartment and subsequently to the sump. This condition ensures the availability of the water for long term cooling of the reactor during the post accident phase. SEQUOYAH - UNIT 1 B 3/4 6-5
INSERT to section 3/4.6.5.3 The OPERABILITY of the ice condenser doors ensures that these doors will open because of the differential pressure between upper and lower containment resulting from the blowdown of reactor coolant during a LOCA and that the blowdown will be diverted through the ice condenser bays for heat removal and thus containment pressure control. The requirement that the doort be maintained closed during normal operation ensures that excessive sublimation of the ice will not occur because of warm air intrusion from the lower containment. l
.D
'w ySLANT SYSTEMS 4
BASES 3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of also This dose 10 CFR Part 100 limits in the event of a steam line rupture. includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the accident analyses. 3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES ~ The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator,will blowdown.in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure' rise within containment in the event the steam line l rupture occurs within containment. The OPERABILITY of the main steam isolation \ vr.lves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses. 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70 F and 200 psig are based on a steam generator RTNDT G F and are sufficient to prevent brittle fracture. of[,g I 3/4.7.3 COMPONENT COOLING WATER SYSTEM t The OPERABILITY of the component cooling water system ensures that sufficient cooling capacity is available for continued operation of safety related equipment
.during normal and accident conditions. The redundant cooling capacity of this ,
- system, assuming a single failure, is consistent with the assumptions used in the accident analyses.
l 3/4.7.4 ESSENTIAL RAW COOLING WATER SYSTEM P 16 ,, The OPERABILITY of the essential raw cooling water system ensures that sufficient cooling capacity is available for continued operation of safety The redundant cooling related equipment during normal and accident conditions. i l capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident conditions within acceptable limits. l V MAR 251982 l g B 3/4 7-3 Amendment No. 12 - SEQUOYAH - UNIT 1 i L-
E 2.1 SAFETY LIMITS BASE 5 - 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overneating of the fuel and possiole clacding perforation wnich would result in the release of fission prooucts to the reactor coolant. Overneating of the fuel claccing in prevented-by restricting fuel operation to within the nucleate boiling regime wnere the heat transfer coefficient is large and the claading surface temperature is slightly above the coolant saturation temperature. Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatu.res because of the onset of departure from nucleate boiling (DNS) and the resultant sharp reduction 1, heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB througn the W-3 correlation. The W-3 ONS correlation has been developed to predict the DNB flux and'the location of DNB for axially uniform and non-uniform neat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause ONB at a particular core location to the local heat flux, is indicative of the margin to DNB. The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.30. This value corresponds to a 95 percent probability at a 95 percent conficence level
#. ' that DNB will not occur and is chosen as an appropriate margin ~to DNB for all O'- operating conditions.
The curves of Figures 2.1-1 anc 2.1-2 show the loci of points of THERMAL. POWER, Reactor Coolant System pressure and average temperature for which tne minimum DNBR is no less than 1.30, or the average enthalpy at tne vessel exit is equal to the enthalpy of saturated liquid. These curves are based on an enthalpy hot channel factor, F g, of 1.55 and a reference cosine with a peak of 1.55 for axial power shape.- An allowance 9 isincludedforanincreaseinFyatreducedpowerbasecontheexpressiorr. FSH=1.55[1+0.3(1-P)] wnere r is the fraction of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated 'for the range of all control rocs fully witharawn to the maximum allowaole control ..
. rod insertion assuming the axial power imbalance is within the limits of the f3 (celta I) function of the Overtemperatu trip. Wnen the axial power ,
imbalance is not within the tolerance, the ial power imbalance effect on the Overtemoerature delta T trips will reduce he setpoints to provide protection consistent with core safety limits. p 9s\%.T
.- SEQUOYAH - UNIT 2 B 2-1 Amendment No. 21
*SEP 2 91993 L
2.2 LIMITING SAFETY SYSTEM SETTINGS BASES ,, Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protective instrumentatinn channels and provides manual reactor trip capability. Power Range, Neutron Flux .
- The Power Range, Neutron Flux channel high setpoint provides reactor core protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective, circuitry. The low set point provides redundant protection in the power range for a power excursion beginning from -
low power. The trip associated with the low setpoint may be manually bypassed-when P-10 is active (two of the four power range channels indicate a power 10 Clevel of above approximatetyg percent of RATED THERMAL POWER) and is auto- [ matically reinstated when P-10 becomes inactive (three of the four channels indicate a power level below approximately 9 percent of RATED THERMAL POWER). Power Range, Neutron Flux, Hich Rates t' The Power Range Positive Rate trip provides protection against rapid flux
~
. increases which are characteristic of rod ejection events from any power' level. Specifically, this. trip complements the Pawer Range Neutron Flux .
., c High and Low trips to ensure that the criteria are met for rod efection from
~
partial power. The Power Range Negative Rate trip provides protectirn to ensure that the miriimum DNBR is maintained above 1.30 for control rod drop accidents. At high power a single or multiple rod drop accident could cause local flux peaking
, which, when in conjunction with nuclear power being maintained equivalent to turbine power by action of the automatic rod control system, could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor for all single or multiple dropped rods.
Intermediate and Source Ranoe, Nuclear Flux The Intermediate and Source Range, Nuclear Flux trips provide reactor 1. ez; ., _- core protection during reactor startup. These trips provide redundant protec-
^ p' : ' - tiontothelowsetpointtripofthePowerRange,NeutronFlux4hannels.- The' '
~
Source Range Channels will initiate a reactor trip at about 10 counts per second unless manually blocked when P-6 becomes active. The Intermediate
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. - 3
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SEQUOYAH - UNIT 2 B 2-3
LIMITING SAFETY SYSTEM SETTINGS q Q_U) m ' BASES ' Steam Generator Water Level The Steam Generator Water Level Low-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume reouired for adequat^ heat removal capacity. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to allow
- for starting delays of the auxiliary feedwater system.
Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level The Steam /Feedwater Flow Mismatch in coincidence with a Steam Generator Low Water Level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the Reactor , Protection System. This trip is redundant to the Steam Generator Water Level Low-Low trip. The Steam /Feedwater Flow Mismatch portion of this trip is , activated when the steam flow exceeds the feedwater flow by greater than or equal to 1.5 x 10' lbs/ hour. The Steam Generator Low Water level portion of the trip is activated when the water level drops below 24 percent, as indicated by the narrow range instrument. These trip values, include sufficient allowance in excess of normal operating values.to preclude spurious trips but will . p-Qyq initiate a reactor trip before the steam generators are dry. Therefore, the ~ required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient on the Reactor Coolant System and steam generators is minimized. Undervoltace and Underfrecuency - Reactor Coolant pumo Busses The Undervoltag'e and Underfrequency Reactor Coolant Pump bus trips provide reactor core protection against DNB as a result of loss of voltage or underfrequency to more than one reactor coolant pump. The specified set points assure a reactor trip signal is generated before the low flow trip set point is reached. Time delays are incorporated in the underfrequency and , undervoltage trips to prevent spurious. reactor trips from momentary electrical power transients. For undervoltage, the delay is set so that the time required for a signal to reach the reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers snall not exceed
'f,'2, . M ' seconds. For cnderfrequency, the delay is set so that the time required.. P for a signal to reach the reactor trip brea.uts i m 7 . :derfrequency trip ~~
set point is reached shall not exceed A f seconcs. - 0;to < u.
!D SEQUOYAH - UNIT 2 B 2-6
LIMITING SAFETY SYSTEM SETTINGS
- ,t BASES 4
Turbine Trip 9 ATurbineTripcausesadirectreactortripwhenoperatingabove.P-\ Each of the transient. turbine trips provide turbine protection and reduce the severity of [ the ensuing No credit was taken in the accident analyses for opera-tion of these trips. Their functional capability at the specified trip settings is required to enhance the overall reliability of the Reactor Protection System. Safety Injection Input from ESF If a reactor trip has not already been generated by the reactor protective instrumentation, the ESF automatic actuation logic channels will initiate a - - reactor trip upon any signal which initiates a safety injection. This trip is provided to protect the core in the event of a LOCA. The ESF instrumentation
. channels which initiate a safety injection signal are shown in Table 3.3-3.
Reactor Trip System Interlocks The Reactor Trip System Interlocks perform the following functions on
, _ increasing power:
J ' P-6 Enables the manual block of the source range reactor trip ". (i.e., prevents; premature block of source range trip).
- P-7 Defeats the automatic block of reactor trip on: Low flow in more .
P-13 than one. primary coolant loop, reactor coolant pump undervoltage and underfrequency, pressurizer low pressure, and pressurizer high level. P-8 Defeats the automatic block of reactor trip on low RCS coolant flow in a single loop. P-9 Defeats the automatic block of reactor trip on turbine trip. P-10 Enables the manual block of reactor trip on power range (low setpoint), intermediate range, as a backup block for source range, and intermediate range rod stops (i.e. , prevents premature block of the noted functions). On decreasing power, the opposite function is performed at reset setpoints. P-4 Reactor-tripped - Actuates turbine trip, closes main feedwater valves'
.' on T ' below setpoint, prevents the opening of the main feedwater valve 9which were~ closed by a safety injection or high steam generator water level signal, allows manual block of the automatic reactuation of safety injection. ,h! Reactor not tripped - defeats manual block preventing automatic reactuation of safety injection.
SEQUOYAH - UNIT 2 B 2-7
~
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, _ _ - . _ _ _ . _ _ . " * * ~ - ' r - " "*-"v
' - - - - - - - - ' ' =
REACTIVITY CCNTROL SYSTEMS A
- U. 7
'~ f BASES BORATION SYSTEMS (Continued) provide a SHUTDOWN MARGIN from expected operating conditions of 1.6% delta k/k after xenon decay and cooldown to 200*F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 5408 gallons of.20,000 ppe borated water from the boric acid storage tanks or 64,160 gallons of 2000 ppm borated water from the refueling water storage tank. With the RCS temperature below 200*F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable. The boron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN of 1% delta k/k after xenon decay and cooldown from 200*F to 140*F. This condition requires either 835 gallons of 20,000 ppm borated water from the boric acid storage tanks or 9,690 gallons of 2000 ppm borated water from the refueling water storage tank. The contained water volume limits include allowance for water not g available because of discharge line location and other physical g characteristics. 9 The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between and LLdr for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6. 3/4.1.3 MOVABLE CONTROL ASSEMBLIES , The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is main-tained, and (3) limit the potential effects of rod misalignment on associated accident analyses. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with ~ the control rod alignment and insertion limits. L <C3 D SEQUOYAH - UNIT 2 B 3/4 1-3
.-- -- .- . _ . . . .- - .. _-.- -=
INSTRUMENTATION N
')
BASES - 4 4 3/4.3.3.4~ METEOROLOGICAL' INSTRUMENTATION i 4 The OPERABILITY of the meteorological instrumentation ensures that 1 sufficient meteorological data fa available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere.- This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972. - j 3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION 4 The OPERABILITY of the remote 'shutd'own instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT i STANDBY of the facility from locations outside of the control room. This capability is required n the event control room habitability is lost and is
- consistent'with Genera
% A _h p.4mh\ c g.b" DesignCriteria19of10CFR50.l;%h.=5-(5'P'A 3/4.3.3.6 CHLORINE DETECTION SYSTEMS gQ sbuth,.n The OPERABILITY of the chlorine detection system ensures that sufficient capability is available to promptly detect and initiate protective action in
_d;Q - the event of an accidental chlorine release. This capability is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95, " Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release," February 1975. 4 3/4.3.3.7 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and i Following an Accident," December 1975. Sequoyah has four separate methods of determining safety valve position (i.e., l open or closed).
- a. Acoustic flow monitors mounted on each safety valve line (one per valve). ..
- A flow indicating module in the main control room is calibrated to detect failure of a valve to reclose. An alarm in the main control room will
. actuate when any valve is not fully closed. L i b. Temperature sensors downstream of each safety valve (one per valve). Tem-3' perature indication and alarm are provided in the main control room.
- c. Pressurizer relief tank temperature, pressure and level indication, and alarm in main control room.
ff d. Pressurizer pressure indication and alarm in the main control room. SEQUOYAH - UNIT 2 - B 3/4 3-3 Amendment No. 35 January 29, 1986
- ( ) EMERGENCY CORE COOLING SYSTEMS BASES REFUELING WATER STORAGE TANK (Continued)
RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses. The contained water volume limit includes an allowance for water not usable because of tank discharge line lodation or other physical characteristics. 1.v. - q .T The limits on contained wate volume and boron concentration of the RWST also ensure a pH value of between and rlJed for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. 1 N SEQUOYAH - UNIT 2 B 3/4 5-3
CONTAINMENT SYSTEMS BASES INTERNAL PRESSURE (Continued) g\wMS N g', v " containment peak pressure does not exceed the g d::!;n pressure of 12 psig during LOCA conditions. h e - 9 ur ;29 ; :::u :rperted t: bc tt:i d 'r:: : ' CC? :::nt ::
.9 p;ig '!:F i ! :: th r the d::ign p :::;r :nd ,; ::n;i;t;nt itt 1
- !d nt :n;!y::: 'i:F 'nc! d:: : -itic! ;;:it!;: :: ici :: t pr:::;r: :'
O.2 ;;ig. 3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that 1) the M **",
- _
containment air mass is limited to an initial mass sufficiently low to prevent exceeding tneAdeu p pressure during LOCA conditions and 2) the ambient air M e " O temperature does not exceed that temperature allowable for the continuous duty hke.d rating specified for equipment and instrumentation located within containment. The containment pressure transient is sensitive to the initially contained air mass during a LOCA. The contained air mass increases with decreasing temperature. The lower temperature limits of 100*F for the lower compartment, 85*F for the upper compartment, and 60 F when less than or eoual
':to !:::
5%t':n of RATED th ::rt:fTHERMAL POWER
- nt d::ign pr:::;r:will
- ' 'limit
- ; ;the
. ;,.peak pressure to f!.0 p;i; _nic " M 4 Q ,'
The upper tempera- g ture limit influences the peak accident temperature slightly during a LOCA; * ' however, this limit is based primarily upon equipment protection and anticipated operating conditions. Both the upper and lower temperature limits are consistent with the parameters used in the accident analyses. 3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment steel vessel will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that , the vessel will withstand the maximum pressure of .JJpsig in the event of a ~ (1 LOCA. A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability. 3/4.6.1.7 SHIELD BUIt. DING STRU:TURAL INTEGRITY This limitation ensures that the structural integrity of the containment shield building will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to provide 1) protection for the steel vessel from external missiles, 2) radiation shielding in the event of a LOCA, and 3) and annulus surrounding the steel vessel that can be maintained at a negative pressure during accident conditions.
- SEQUOYAH - UNIT 2 8 3/4 6-2
l CONTAINMENT SYSTEMS BASES 3/4.6.1.8 EMERGENCY GAS TREATMENT SYSTEM (EGTS) The OPERABILITY of the EGTS cleanup subsystem ensures that during LOCA conditions, containment vessel leakage into the annulus will be filtered through the HEPA filters and charcoal adsorber trains prior to discharge to the atmosphere. This requirement is necessary to meet the assumptions used in the accident analyses and limit the site boundary radiation doses to within the limits of 10 CFR 100 during LOCA ccnditions. Cumulative operation of the system with the heaters on for 10 hours.over a 31 day period is sufficient to reduce the buildup of moisture on the assorbers and HEPA filters. ANSI N510-1975 will be used as a procedural guide for surveillance testing. 3/4.6.1.9 CONTAINMENT VENTILATION SYSTEM Use of the cont'ainment' purge lines is restricted to only one pair (one supply line and one exhaust line) of purge system lines at a time to ensure that the site boundary dose guidelines of 10 CFR Part 100 would not be exceeded in the event of a loss of coolant accident during purging operations. The analysis of this accident assumed purging through the largest pair of lines (a 24 inch inlet line and a 24 inch outlet line), a pre-existing iodine spike in the reactor coolant and four second valve closure times.
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1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the containment spray system ensures that containment
- depressurization and cooling capability will be available in the event of a LOCA. The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the accident analyses.
l 3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the l containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment. Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a - LOCA. 3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable , q} limit during post-LOCA conditions. Either recombiner unit or the puage system 4 SEQUOYAH - UNIT 2 8 3/4 6-3 ea m%,3 4on
~ CONTAINMENT SYSTEMS BASES pone:#.3 8 68 %A'*5 # COMBUSTIBLE GAS CONTROL (Continued) ip:+rns pra. i
,,,.+>
4 s capable of controlling the expected hydrogen generation associated with
- 1) zirconium-water reactions, 2) radiolytic decomposition of water and gp y 3) corrosion,of metals within containment. These hydrogen control systems are d6P d b-
% d, rah ::-rittea' m th th: ::: m :nd:ti: : c' Regulatory Guide 1.7, " Control of h h P"". ,( o oc,,,;&g ombustible Gas Concentrations in Containment Following a LOCA",l rm% 2. Mcl Ibe4 9'"4"*'"' p'i The hydrogen mixing systems are provided to ensure adequate mixing of the ivia containment atmosphere following a LOCA. This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit.
The operability of at least 66 of 68 ignitors in the hydrogen control distributed ionition system will maintain an effective coverage throughout the R 21 containment. ~This system of ignitors will initiate combustion of any signifi-cant amount of hydrogen released after a degraded core accident. This system is to ensure burning'in a controlled manner as the hydrogen is released instead of allowing it to be ignited at high concentrations by a random ignition source. 3/4.6.5 ICE CONDENSER The requirements associated with each of the components of the ice condenser .O ensure that the overall system will be available to provide sufficient pressure suppression cacability to limit the containment peak pressure transient to s b'/ less than 12 psig during LOCA conditions. 3/4.6.5.1 ICE BED The OPERABILITY of the ice bed ensures that the required ice inventory will 1) be distributed evenly through the containment bays, 2) contain suffi-cient boron to preclude dilution of the containment sump following the LOCA and 3) contain sufficient heat removal capability to condense the reactor systen volume released during a LOCA. These conditions are consistent with , the assumptions used in the accident analyses. The minimum weight figure of 1200 pounds of ice per basket contains a 10% conservative allowance for ice loss through sublimation which is a factor of 10 higher than assumed for the ice condenser design. The minimun weight fioure of 2,333,100 pounds of ice also contains an additional 1% conservative , a.llowance to account for systematic error in weighing instruments. In the event that observed sublimation rates are equal to or lower than design predictions after three years of operation, the minimum ice baskets weight may be adjusted downward. In addition, the number of ice baskets required to be weighed each 9 months may be reduced after 3 years of operation if such a reduction is supported by observed sublimation data.
- f~'s
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- SEQUOYAH - UNIT 2 B 3/4 6-4 Amendment No. 21 Q SEP 2 91983
/
CONTAINMENT SYSTEMS BASES 3/4.6.5.2 ICE BED TEMPERATURE MONITORIWG SYSTEM The OPERABILITY of the ice bed temperature monitoring system ensures that the capability is available for monitoring the ice temperature. In the event the monitoring system is inoperable, the ACTION requirements provide assurance that time the ice bed heat removal capacity will be retained within the specified 1imits. . 3/4.6.5.3 ICE CONDENSER 000RS D.- h Th CICAC:LITY cf the.it: c:nden::r decr: nd th C';uire :nt th:t they be -.ci* ained !;::d en:ure; that th r ::ter :: !:nt :y:t= fluid r:!::::d d ain; ; LOCA will b; di;= t d thr; ugh th: it: :nd:n:Or bay: for h::t r:-:;;'-
;nd that =::::i : :tb!':: tier of the it: bed "' n:t : cur b::: :: O f c r-
-cir .tru:icr 3/4.6.5.4 INLET 000R POSITION MONITORING SYSTEM The OPERABILITY of the inlet door position monitoring system ensures that the capability is available for monitoring the individual inlet door position.
In the event the monitoring system is inoperable, the ACTION requirements provide assurance that the ice bed heat removal capacity will be retained within the specified time limits. 3/4.6.5.5 DIVIDER BARRIER PER5ONNEL ACCESS DOORS AND EQUIPMENT H The requirements for the divider barrier personnel access doors and equipment hatches being closed and OPERABLE ensure that a minimum bypass steam flow will occur from the lower to the upper containment compartments during a LOCA. This condition ensures a diversion of the steam through the ice condenser bays that is consistent with the LOCA analyses. 3/4.6.5.6 CONTAINMENT AIR RETURN FANS The OPERABILITY of the containment air return fans ensures that following a LOCA 1) the containment atmosphere is circulated for cooling by the spray system and 2) the accumulation of hydrogen in localized portions of the contain-ment structure is minimized. " 3/4.6.5.7 and 3/4.6.5.8 FLOOR AND REFUELING CANAL ORAINS The OPERABILITY of the ice condenser floor and refueling canal drains ensures that following a LOCA, the water from the melted ice and containment spray system has access for drainage back to the containment lower compartment and subsequently to the sump. This condition ensures the availability of the water for long term cooling of the reactor during the post accident phase. SEQUOYAH . UNIT 2 B 3/4 6-5
INSERT to section 3/4.5.s.3 The OPERABILITY of the ice condenser doors ensures that these doors will open because of the differential pressure between upper and lower containment resulting from the blowdown of reactor coolant during a LOCA and that the blowdown will be diverted through the ice condenser bays for heat removal and thus containment pressure control. The requirement that the doors be maintained closed during normal operation ensures that excessive sublimation of the ice will not occur because of warm air intrusion from the lower containment. T 6 ee L ~ -- - _ _ . _ _ . _ . _ _ _ _ _ _ _ _ _ _ _
-PLANT SYSTEMS V BASES 3/4.7.1.5 DIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with-the assumptions used in the accident analyses.
3/4.7.2 STEAM GENERATOR PRESSUPE/ TEMPERATURE LIMITATION The. limitation on steam generator pressure and temperature ensures that
- the pressure induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70*F and 200 psig are based on a steam generator RTNDT f AC'F and are sufficient to prevent l brittle fracture. (
3/4.7.3 COMPONENT COOLING WATER SYSTEM D
'._j The OPERABILITY of the component cooling water system ensures that
~
sufficient cooling capacity is available for continued operation of safety related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses. 3/4.7.4 ESSENTIAL RAW COOLING WATER SYSTEM The OPERABILITY of the essential ra.t cooling water system ensures that sufficient cooling capacity is available for continued operation of safety related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the* assumptions used in the accident conditions within acceptable limits. SEQUOYAH - UNIT 2 B 3/4 7-3
ENCLOSURE 2 TECHNICAL SPECIFICATION BASES CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-87-14) DESCRIPTION AND JUSTIFICATION FOR REVISION OF MISCELIANEOUS BASES f I I
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The following changes only involve the bases section in Sequoyah units 1 and 2 technical specifications. This information is being provided to NRC for information purposes. Because of the high number and diversity of
. these changes, they will be given a unique identifier and presented in a tabular format for clarity. , Description of Change i Change Bases Page Number Number i Section l Unit 1 l Unit 2 'l Description l l l l The phrase " Delta T" is inserted 1 l 2.1.1 l B 2-2 l B 2-1 l between the words "Overtemperature" l l l l and " trip."
. l l l 1 l l l l The stated percent of Rated Thermal 1 2 l 2.2.1 l B 2-3 l B 2-3 l Power at which P-10 becomes active l l is changed from 9 to 10. l l l 1 1 I l l l l The stated maximum time delay for 3 l 2.2.1 l B 2-6 l B 2-6 l an undervoltage or underfrequency l l l l l signal to reach the reactor trip 1 l l l breakers is changed from 0.9 l l + l l l second to 1.2 seconds and 0.3 l l l second to 0.6 second, respectively. I I I ' I l l l l The stated permissive which causes 4 l 2.2.1 l B 2-7 l B 2-7 l a direct reactor trip on a turbine l l l l trip is changed from P-7 to P-9. l l l 1 l l l l The stated pH range for the solution 5 l 3/4.1.2 l B 3/4 1-3 l B 3/4 1-3 l recirculated within the containment l l l l after a LOCA is changed fro.n a value l l l l l between 8.5 and 11.0 to a value l l l between 7.5 and 9.5. l 1 l l l l Adds the phrase "and the potential l l 6 l 3/4.3.3.5 l B 3/4 3-3 l B 3/4 3-3 l capability for subsequent cold shut-l l l l down" to the description of the l l l l l function of the Remote Shutdown l l l Instrumentation. I l l 1 7 l 3/4.5.5 l B 3/4 5-3 l B 3/4 5-3 l Same as change number 5 above. ' l l I 1 l ' l l l Replaces the word " design" with 8 l 3/4.6.1.4 l B 3/4 6-2 l B 3/4 6-2 l " maximum allowable internal" when -- l l l l discussing the limiting containment l l l l pressure value. l l l l Also, the last paragraph in this l l l l section is deleted. 4
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_-,,_.,._m-.-m,_-m_m--
Change' Bases Page Number Number l Section l Unit 1 l Unit 2 l Description l l l 1 l l l l Replaces the word " design" with 9 l 3/4.6.1.5 l B 3/4 6-2 l B 3/4 6-2 l " maximum allowable internal" when l l l l discussing the limiting containment l l l l pressure value. l l l l Also, the specific values referenced l l l l as the peak pressure and design l l l l pressure are replaced with "an l l l l acceptable value." l 1 1 I l l l l The referenced maximum pressure 10 l 3/4.6.1.6 l B 3/4 6-2 l B 3/4 6-2 l is revised from 11.8 psig to 12 psig. l l l l l l l l Several revisions to section 3/4.6.4 11 l 3/4.6.4 l B 3/4 6-4 l B 3/4 6-3&4 l are made to clarify the purpose and l and i and l l design of the combustible gas control l 3/4.6.4.3 l B 3/4 6-4a l N/A l systems. Section 3/4.6.4.3 on the l l l l old interim hydrogen mitigation l l l l system in unit 1 is deleted, and a l l l l paragraph is added to section 3/4.6.4 l l l l to describe the permanent hydrogen l l l l mitigation system. Section 3/4.6.4.3 l l l l does not exist in unit 2 technical l l l l specifications. I I I I l l l l This section is rewritten to provide 12 l 3/4.6.5.3 l B 3/4 6-5 l B 3/4 6-5 l a clearer description of the ice l l l l condenser doors' function. l l 1 I l l l l The stated reference temperature for 13 l 3/4.7.2 l B 3/4 7-3 l B 3/4 7-3 l nill ductility transition RT NDT l l is changed from 600F to 250F. I l ] _ , - _ - _ . _ . _ _ _ _ _ , __ . , _ . . _ _ , , _ _ . . _ . _ _ _ _ , _ __ _ _ . , _ _ _ , , _ _ _ . . - - , _ _ . _ _ _ _ _ ~ , _ _ _ . . - _ , . _ , - - _ , _ . _ _ _ _ , _ . _ , _ _ , _ _ . . -
Reason and Justification Change Number l Reason and Justification i 1 l The reactor trip discussed in this section is the Overtemperature Delta T l trip. An "Overtemperature trip" does not exist at Sequoyah. I 2 l On power ascension, P-10 allows blocking of the Source, Intermediate, and l Power Range (low setpoint) Reactor trips at a setpoint of 2 10% rated l thermal power. The value of 10%, not 9%, is the P-10 setpoint as shown in l table 2.2-1 of the technical specifications. I 3 l The limiting undervoltage and underfrequency delays are given in j table 3.3-2. The values shown in that table should be used in basis 2.2.1. 1 4 l P-9, not P-7, is the permissive which blocks a direct reactor trip on a l turbine trip at thermal power levels s 50%. The function of P-9 is l discussed in table 2.2-1 of the technical specifications. The reference j to P-7 is apparently a typographical error. I 5 l The recirculation solution pH range of 8.5 to 11.0 currently specified in l the basis is not consistent with the range specified in chapter 6 of the FSAR, l nor is it consistent with current NRC and industry recommendations. The l equilibrium sump recirculation solution pH specified in section 6.2.2.2 of l the FSAR is 8.0 to 8.2. NRC recommends a recirculation solution pH of 4 l 7.0 to 9.5 to minimize the occurrence of chloride-induced stress corrosion l cracking of stainless steel (branch technical position MTEB 6-1). The ANS l recommends a long-term recirculation solution pH of 8.5 to 9.5 to maximize l iodine retention and minimize chloride-induced stress corrosion cracking s l (ANSI /ANS-56.5-1979). l Based on minimizing chloride-induced stress corrosion cracking, the minimum l pH could be as low as 7.0. However, because of measurement and calculational l errors, the minimum pH should be set at 7.5 to ensure that the actual pH l will not fall below 7.0. l l Based on the considerations stated above, the recirculation solution pH l specified in the basis should be 7.5 to 9.5. I 6 l The remote shutdown instrumentation at Sequoyah is designed to have the l capabilities set forth in General Design Criterion (GDC) 19 of 10 CFR 50, l Appendix A. GDC 19 delineates two capabilities for remote instrumentation l and controls. The phrase added to section 3/4.3.3.5 is the second capability l described in GDC 19. I 7 l Same as number 5 above. I b___.___--..__ - -
Reason and Justification Change Number l Reason and Justification 1 8 l The design pressure of Sequoyah's primary containment vessel is 10.8 psig. l The maximum allowable internal pressure is 12 psig. This difference is in l accordance with ASME, Boiler and Pressure Vessel Code, Section III, 1971 l Edition. This change corrects the terminologies used in the bases to be l consistent with the version of the ASME Code to which the vessel was l designed. l l Also, references made to a specific peak pressure as determined by the FSAR l accident analyses are deleted. This is prudent because the peak pressure l value is subject to change as the FSAR analyses are updated and revised l periodically. I 9 l Same as number 8 above. I 10 l Same as number 8 above. I 11 l The combustible gas control systems at Sequoyah were enhanced by l adding the hydrogen mitigation system. This system consists of 68 hydrogen l ignitors per unit to ensure burning of hydrogen in controlled manner before a l dangerous quantity could accumulate. This change deletes references to an l interim system which was temporarily installed in unit 1 and clarifies the j design of the current combustible gas control system installed in both l units 1 and 2. I 12 l The wording in this section, 3/4.6.5.3, was not very clear, and it implied ' l that the ice condenser doors stayed closed during a LOCA, which is l incorrect. Therefore, this section was rewritten to more accurately describe l the purpose and function of the ice condenser doors. I 13 l Vendor documentation on the steam generators states that the RTNDT is l 250F, not 60 F as shown in the basis. \
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l ENCLOSURE 3 TECHNICAL SPECIFICATION BASES CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-87-14) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS REVISION OF MISCELLANEOUS BASES I et f i l I l l
c , i SIGNIFICANT HAZARDS EVALUATION
- 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
No. The changes only involve updating, clarifying, and correcting the bases given for existing technical specification requirements. The changes do not affect the plant's design, method of operation, procedures, or testing.
- 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
4 No. The changes do not affect the plant's design, method of operation, procedures, or testing.
- 3. Does the proposed amendment involve a significant reduction in a margin of safety?
. No. The changes do not affect the plant's design, method of operation, procedures, or testing. Thus,the existing margins of safety are not affected. L J 4 J l __ _ . . _ . . . _ ___ . . _ . . .-.m ._.. . , _ _ _ _ _ . . _ _ _ . - - _ . , . . , . . , . _ _ , _ , _ . _ _ _ . . _ _ . . _ _ _ _ . . _ _ _ _ . , - - . . _ . _ . _ _ . . _ _ . _ ,}}