ML20237K381

Accepts 870417 Revised Tech Spec Bases as Part of Util Restart Effort.Revised Bases Pages Encl

ML20237K381
Person / Time
Site:	Sequoyah
Issue date:	08/18/1987
From:	Zwolinski J NRC OFFICE OF SPECIAL PROJECTS
To:	White S TENNESSEE VALLEY AUTHORITY
References
TAC-00104, TAC-00105, TAC-104, TAC-105, NUDOCS 8708270165
Download: ML20237K381 (46)

Text

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August 18, 1987 Docket Nos. 50-327/328 Mr. S. A. White Manager of Nuclear Power Tennessee Valley Authority 6N 38A Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801

SUBJECT:

' CHANGES TO TECHNICAL SPECIFICATION BASES (TS 87-14) (TAC 00104, 00105)

Re: Sequoyah Nuclear Plant, Units 1 and 2 By letter dated April 17, 1987, Tennessee Valley Authority (TVA) submitted ,

revised bases pages for the Sequoyah Nuclear Plant, Units 1 and 2 Technical Specifications (TS). As a part of the Sequoyah restart effort, TVA has evaluated the accuracy and clarity of these bases. As a result of this evaluation, TVA has revised a number of the bases and requested that they be incorporated into the bases portion of the Sequoyah TS. The NRC staff has reviewed your request and concluded the revised bases are appropriate and remain consistent with the TS. Therefore, we agree that it would be appropriate to incorporate the revised {

bases pages into the controlled TS documents. ~ Copies of the revised bases are <

provided as an enclosure.

Please direct any questions you ma Rotella, Project Manager, at (301)y have concerning .this issue to Mr. Thomas S.

492-9043.

Sincerely, 1  %

3 on wb i istant Director for Projects I

TVA Projects Division Office of Special Projects

Enclosures:

_ DISTRIBUTION Revised bases pages ' Docket Filet JZwolinski OGC-BETH NRC PDR BDLiaw FMiraglia cc: See next page Local PDR GZech, RII. EJordan ,

JKeppler/JAxel rad SRConnelly JPartlow  ;

SEbneter 'CJamerson ACRS (10) j SRichardson TRote11a TVA-BETH j DHagan GPA/PA EButcher i TBarnhart (8) Wanda Jones LFMB l SQN Rdg Projects Rdg TVA:0SP/LA 3

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PDR ADOCK 05000327 l P PDR l 1

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g[ "%g',.g c UNITED STATES  !

t y NUCLEAR REGULATORY COMMISSION 5 z'g "ig.. ql WASHINGTON, D. C. 20555 g3 af f .

"%,[g# August 18, 1987 Docket Nos. 50-327/328  !

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l Mr. S. A. White Manager of Nuclear Power Tennessee Valley Authority ,

6N 38A Lookout Place  !

1101 Market Street i Chattanooga, Tennessee 37402-2801 i

SUBJECT:

CHANGES TO TECHNICAL SPECIFICATION BASES (TS 87-14) (TAC 00104, 00105)

Re: Sequoyah Nuclear Plant, Units 1 and 2 i

By letter dated April 17, 1987, Tennessee Valley Authority (TVA) submitted revised bases pages for the Sequoyah Nuclear Plant, Units 1 and 2 Technical Specifications (TS). As a part of the Sequoyah restart effort, TVA has evaluated the accuracy and clarity of these bases. As a result of this evaluation, TVA has revised a number of the bases and requested that they be incorporated into the bases portion of the Sequoyah TS. The NRC staff has reviewed your request and concluded the revised bases are appropriate and remain consistent with the TS. Therefore, we agree that it would be appropriate to incorporate the revised bases pages into the controlled TS documents. Copies of the revised bases are provided as an enclosure.

Please direct any questions you ma Rotella, Project Manager, at (301)y have concerning this issue to Mr. Thomas S.

492-9043.

Sincerely, d II JohnA.Zwolinski,Asd stad 1 rector

( i for Projects N TVA Projects Division Office of Special Projects

Enclosures:

Revised bases pages cc: See next page

I

' Mr. S. A. White Tennessee Valley Authority Sequoyah Nuclear Plant cc:

General Counsel Regional Administrator, Region II l Tennessee Valley Authority U.S. Nuclear Regulatory Commission 400 West Summit Hill Drive 101 Marietta Street, N.W. 1 E11 B33 Atlanta, Georgia 30323 Knoxville, Tennessee 37902 Resident Inspector /Sequoyah NP .

i Mr. R. L. Gridley c/o U.S. Nuclear Regulatory. Commission Tennessee Valley Authority 2600 Igou Ferry Road 5N 157B Lookout Place Soddy Daisy, Tennessee 37379 Chattanooga, Tennessee 37402-2801 Mr. . Richard King Mr. H. L. Abercrombie e/o U.S. GA0 Tennessee Valley Authority lill. North Shore Drive l Sequoyah Nuclear Plant Suite 225, Box 194 )

P.O. Box 2000 Knoxville, Tennessee 37919 l Soddy Daisy, Tennessee 37379 Tennessee Department of Mr. M. R. Harding Public Health Tennessee Valley Authority ATTN: Director, Bureau of l

Sequoyah Nuclear Plant Environmental Health Services P.O. Box 2000 Cordell Hull Building Soddy Daisy, Tennessee 37379 Nashville, Tennessee 37219 Mr. D. L. Williams Mr. Michael H. Mobley, Director lennessee Valley Authority Division of Radiological Health 400 West Summit Hill Drive T.E.R.R.A. Building W10 B85 150 9th Avenue North Knoxville, Tennessee 37902 Nashville, Tennessee 37203 County Judge Hamilton County Courthouse Chattanooga, Tennessee 37402

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FACILITY OPERATING LICENSE NO. DPR-77 )

DOCKET NO. 50-327 Revise the Appendix A Technical Specifications Bases by removing the pages' identified below and inserting the encicsed pages. The revised pages contain marginal lines indicating the area of change. Overleaf and repaginated pages*

are provided to maintain document completeness.

l REMOVE ' INSERT B 2-1* I B 2-1 B 2-2 B 2 B'2-3 B 2 B 2-4 B 2-4*

B 2-5 B 2-5*

B 2 B 2-6' l 8 2-7 B 2-7 i B 3/4 1-3 B 3/4 1-3 l B 3/4 1-4 B 3/4 1-4* l B 3/4 3-3 B 3/4 3-3 l B 3/4 3-3a -- l B 3/4 3-4 B 3/4 3-4* l B 3/4 3-5* l' B 3/4 5-3 B 3/4 5-3 B 3/4 6-1 B 3/4 6-1*

B-3/4 6-2 B 3/4 6-2 B 3/4 6-3 B 3/4 6-3*

B 3/4 6-3a --

B 3/4 6-4 B 3/4 6-4 B 3/4 6-4a --

B 3/4 6-5 B 3/4 6-5 B 3/4 6-6 B 3/4 6-6 B 3/4 7-3 B 3/4 7-3 B 3/4 7 4 B 3/4 7-4*

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I , 2.1 SAFETY LIMITS  ;

l BASES

' 1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented ,

by restricting fuel operation to within the nucleate boiling regime where the l heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.  !

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer j coefficient. DNB is not a directly measurable parameter during operation and i therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been l related to DNB through the W-3 correlation. The W-3 DNB correlation has been l developed to predict the DNB flux and the location of DNB for axially uniform J and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR,  !

defined as the ratio of the heat flux that would cause DNB 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 I operational transients, and anticipated transients is limited to 1.30. This value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

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The curves of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL l POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

Thecurvesarebasedonanenthalpyhotchannelfactor,Fh,of1.55and a reference cosine with a peak of 1.55 for axial power shape. An allowance is included for an increase in F H at reduced power based on the expression:

F H = 1.55 [1+ 0.3 (1-P)]

where P is the fraction of RATED THERMAL POWER SEQUOYAH - UNIT 1 B 2-1 Amendment No. 19

SAFETY LIMITS c 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 f) (Delta I) function of the Overtemperature. Delta T trip. When the axial power imbalance is not within the' tolerance, the axial power imbalance effect on the Overtemperature Delta T trips will reduce the setpoints to provide protection consistent with core safety limits.

2.1.2 REACTOR C0OLANT 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 transier,t pressure of 110% (2735 psig) of design pressure. The Reactor Coolant System piping, valves and fittings, are designed to ANSI 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. l 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 Setpoints _

have been selected to ensure that the reactor core and reactor coolant system are prevented 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.

SEQUOYAH - UNIT 1 B 2-2 Revised- 08/18/87  ;

, SAFETY LIMITS 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 The Power Range, Neutnn 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 level of above approximately 10 percent of RATED THERMAL POWER) and is auto- l 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, High Rates 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 Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power.

The Power Range Negative Rate trip provides protection to ensure that the minimum 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 unconserva-tive local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping .he 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-tion to the low setpoint trip of the Power Range, Neutron Flux channels. The Source Range Channels will initiate a reactor trip at about 10+5 counts per second unless manually I, locked when P-6 becomes active. The Intermediate SEQUOYAH - UNIT 1 8 2-3 Revised 08/18/87

SAFETY LIMITS BASES Range Channels will initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active. No credit was taken for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System.

Overtemperature Delta T The Overtemperature Delta T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),

and pressure is within the range between the High and Low Pressure reactor trips. This setpoint includes corrections for axial power distribution, changes in density and heat capacity of water with temperature ard dynamic compensation for piping delays from the core to the loop temperature detec-tors. With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.

Operation with a reactor coolant loop out of service below the 4 loop P-8 setpoint does not require reactor protection system setpoint modification because the P-8 setpoint and associated trip will prevent DNB during 3 loop operation exclusive of the Overtemperature Delta T setpoint. Three loop operation above the 4 loop P-8 setpoint is permissible after resetting the K1, K2 and K3 inputs to the Overtemperature Delta T channels and raising the P-8 setpoint to its 3 loop value. In this mode of operation, the P-8 inter-lock and trip functions as a High Neutron Flux trip at the reduced power level.

Overpower Delta T The Overpower Delta T reactor trip provides assurance of fuel integrity, e.g., no melting, under all possible overpower conditions, limits the required range for Overtemperature Delta T protection, and provides a backup to the High Neutron Flux trip. The setpoint includes corrections for changes in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors. No credit was taken for operation of this trip in the accident SEQUOYAH - UNIT 1 B 2-4

SAFETY LIMITS BASES analyses; however, its functional capability at the specified trip setting is required by this specification to enhance thc overall reliability of the Reactor Protection System.

Pressurizer Pressure The Pressurizer High and Low Pressure trips are provided to limit the pressure range in whic' eactor operation is. permitted. The High Pressure trip is backed up by t;.e pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig). The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure.

Pressurizer Water Level i

l The Pressurizer High Water Level trip ensures protection against Reactor

! Coolant System overpressurization by limiting the water level to a volume l sufficient to retain a steam bubble and prevent water relief through the l pressurizer safety valves. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

1 Loss of Flow l The Loss of Flow trips provide core protection to prevent DNB in the i event of a loss of one or more reactor coolant pumps.

Above 11 percent of RATED THERMAL POWER, an automatic reactor trip will l occur if the flow in any two loops drop below 89% of nominal full loop flow.

Above 36% (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 89% of nominal full loop flow. This latter trip will prevent the minimum value of the DNBR from going below 1.30 during normal operational transients and anticipated transients when 3 loops are in operation and the Overtemperature Delta T trip set point is adjusted to-the value specified for all loops in operation. With the Overtsmperature Delta T trip set point adjusted to the value specified for 3 loop operation, the P-8 trip at 76% RATED THERMAL POWER will prevent the minimum value uf the DNBR from going below 1.30 during normal operational transients and anticipated transients-with 3 loops in operation.

SEQUOYAH - UNIT 1 B 2-E

l SAFETY LIMITS .

BASES I

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.

1 Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level I

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 J included in Table 2.2-1 to ensure the functional capability of the specified trip settin'gs and thereby enhance the overall reliability of the Reactor i Protection System. This trip is redundant to the Steam Generator Water Level 1 Low-Low trip. The Steam /Feedwater Flow Mismatch portion of this trip is j activated when the steam flow exceeds the feedwater flow by greater than or  ;

equal to 1.5 x 106 lbs/ hour. The Steam Generator Low Water level portion of i 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 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 4

System and steam generators is minimized.

Undervoltage and Underfrequency - Reactor Coolant Pump 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 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. 4 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 shall not exceed 1.2 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 0.6 seconds.

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SEQUOYAH - UNIT 1 B 2-6 Revised 08/18/87 j

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SAFETY LIMITS BASES

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Turbine Trip -l A Turbine Trip causes a direct reactor trip when operating above P-9.

Each of the turbine trips provide turbine protection and reduce the severity of the ensuing transient. No credit was taken in the accident analyses for opera-tion of these trips. Their functional capability at the specified trip settings I is required to enhance the overall reliability of-the Reactor Protection System.

Safety Injection Input from ESF t

If a reactor trip has not already been generated by the reactor protective  !

instrumentation, the ESF automatic actuation logic channel 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:

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. I 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), l intermediate range, as a backup block for source range, and intermediate range rod steps (i.e., prevents premature block of the noted functions). l 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.

Reactor not tripped - defeats manual block preventing automatic '

reactuation of safety injection.

Amendment No.'7 SEQUOYAH - UNIT 1 B 2-7 Revised 08/18/87

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, REACTIVITY CONTROL SYSTEMS BASES I i

i 5408 gallons of 20,000 ppm 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 change 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 available ,

because of discharge line location and other physical characteristics. I l The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.5 and 9.5 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  ;

l mechanical systems and components.

The OPERABILITY of one boron injection system during REFUELING ensures 'l 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 4

analyses. OPERABILITY of the control rod position indicators is required to l determine control rod positions and thereby ensure compliance with the control i l rod alignment and insertion limits. ]

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l SEQUOYAH - UNIT 1 B 3/4 1-3 Revised 08/18/87 s

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REACTIVITY CONTROL SYSTEMS ,

BASES The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of peaking factors and a restriction in THERMAL POWER. These restrictions provide assurance of fuel rod integrity during continued operation. In addition, those accident analyses affected by a misalignment rod are'reevalu-ated to confirm that the results remain valid during future operation.

The maximum rod drop time restriction'is consistent with the assumed rod drop time used in the accident analyses. Measurement with T avg greater than or equal to 541 F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

' Control rod positions and OPERABILITY.of the rod position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent verifications required if an automatic monitoring channel is inoperable.

These verification frequencies are adequate for assuring that the applicable LCO's are satisfied.

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SEQUOYAH - UNIT 1 B 3/4 1-4 j i

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INSTRUMENTATION BASES 1

1 design basis for the facility to determine if plant shutdown is required j 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.

l l 3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION l

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 l for initiating protective measures to protect the health and safety of the I

public and is consistent with the recommendations of Regulatory Guide 1.23,  !

l "0nsite Meteorological Programs," February 1972.

3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of H0T STANDBY of the facility and the potential capability for subsequent cold shut-down from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR 50.

3/4.3.3.6 CHLORINE DETECTION SYSTEMS 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 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 Following an Accident," December 1975.

SEQUOYAH - UNIT 1 B 3/4 3-3 Revised 08/18/87 1

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i INSTRUMENTATION

.l BASES Sequcyah has four separate methods of determining safety valve position (i.e.,

open or closed).

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a. Acoustic flow monitors mounted on each safety valve line (one per valve).

A flow indicating module in the main control rcom is calibrated to detect failure of a valve to reclose. An alarm in the main control room will j actuate when any valve is not fully closed. )

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b. Temperature sensors downstream of each safety valve (one per valve). Tem-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,

d. Pressurizer pressure indication and alarm in the main control room.

Although all the above position indicators for the pressurizer safety valves and the PORVs are acceptable as one of the channels, the acoustic monitors must be one of the two required operable chanels. In addition to the four methods described above, the PORVs use an electromagnetic " reed"-switch to determine valve position. The stem mounted switches are no longer in use since the PORVs were changed.

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. x SEQUOYAH - UNIT 1 B 3/4 3-4 Amendment No. 43 i Revised 08/18/87 l

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INSTRUMENTATION BASES 3/4.3.3.8 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages. Prompt detection of fires will reduce the potential for damage to safety related equipment and is an integral element in the overall facility fire protection program.

In the event that a portion of the fire detection instrumentation is inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

l 3/4.3.3.9 RADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicC;1e, the releases of radioactive materials in liquid effuents during actual or potential releases of liquid effluents. The alarm /

trip setpoints for there instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

3/4.3.3.10 RADI0 ACTIVE GASE0US EFFLUENT INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring the concentrations of potentially explosive gas mixtures in the waste gas holdup system. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

SEQUOYAH - UNIT 1 B 3/4 3-5 Revised 08/18/87

EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.5 REFUELING 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 boron concentration ensure that

1) sufficient water is available within containment to permit recirculation cooling flow to the care, and 2) the reactor will remain subtritical in the l cold condition following mixing of the RWST and the RCS water volumes with all l control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses.

1 The contained water volume limit includes an allowance for water not usable be 4se of tank discharge line location or other phys cal characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.5 and 9.5 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 systenis and components, l

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3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1 PRIMARY CONTAINMENT Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the  ;

site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.

3/4.6.1.2 CONTAINMENT LEAKAGE l The limitations on containment leakage rates ensure that the total ,

containment leakage volume will not exceed the value assumed in the accident I analyses at the peak accident pressure, Pa. As an added conservatism, the measured overall integrated leakage rate is further limited to less than or i equal to 0.75 La during performance of the periodic tests to acccunt for )

possible degradation of the containment leakage barriers between leakage l tests. I l

The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix "J" of 10 CFR 50. I 3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals provide assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.

3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal ptessure ensure that 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the annulus atmosphere of 0.5 psig ar:d 2) the SEQUOYAH - UNIT 1 B 3/4 6-1

CONTAINMENT SYSTEMS BASES containment peak pressure does not exceed the maximum allowable internal pressure of 12 psig during LOCA conditions.

3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that 1) the containment air mass is limited to an initial mass sufficiently low to prevent exceeding the maximum allowable internal pressure during LOCA conditions and

2) the ambient air temperature does not exceed that temperature allowable for the continuous duty rating specified for equipment and instrumentation located within containment.

The containment pressure transient is sensitive to the initially contained air mass during c 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 equal to 5% of RATED THERMAL POWER will limit the peak pressure to an acceptable value. The upper temperature 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 aesign standards for the life of the facility. Structural integrity is required to ensure that the vessel will withstand the maximum pressure of 12 psig in the event of a LOCA. A visual inspection in conjur: tion with Type A leakage tests is sufficient to demonstrate this capability.

3/4.6.1.7 SHIELD BUILDING STRUCTURAL INTEGRITY This limitation ensures that the structoral 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.

I l

SEQUOYAH - UNIT 1 B 3/4 6-2 Revised 08/18/87

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 HLOA filters and charcoal adsorber trains prior to discharge to l the atmosphe e. This requirement is necessary to meet the assumptions used in l the accident analyses and limit the site boundary radiation doses to within the limits of 10 CFR 100 during LOCA conditions. Cumulative operation of the system with the heaters on for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> over a 31 day period is sufficient to reduce the buildup of moisture on the absorbers and HEPA filters. ANSI N510-1975 ,

will be used as a procedural guide for surveillance testing, 3/4.6.1.9 CONTAINMENT VENTILATION SYSTEM i 1

Use of the containment purge lines is restricted to only one pair (one i supply line and one exhaust line) of purge system lines at a time to ensure j 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 oper tions. The analysis of this accident assumed purging through the largest pair of lines (a l 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.

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM  ;

1 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.

3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the 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. By letters dated March 3, 1981, and April 2, 1981, TVA will submit a report on t!a operating experience of the plant no later than startup after the first refueling. This information will be used to provide a basis to re-evaluate the adequacy of the purge and vent time limits.

SEQUOYAH - UNIT 1 B 3/4 6-3 Amendment No. 5

CONTAINMENT SYSTEMS BASES 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 limit during post-LOCA conditions. Either recombiner unit or the hydrogen mitigation system, consisting of 68 hydrogen ignitions per unit, is capable of controlling the expected hydrogen generation associated with 1) zirconium-water reactions, 2) radiolytic decomposition of water and 3) corrosion of metals within containment. These hydrogen control systefas are designed to mitigate the effects of an accident as described in Regulatory Guide 1.7,

" Control of Combustible Gas Concentrations in Containment Following a LOCA",

revision 2 dated November 1978.

The hydrogen mixing systems are provided to ensure adequate mixing of the 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 mitigation system will maintain an ef fective 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 t) 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 con-denser ensure that the overall system will be available to provide sufficient pressure suppression capability to limit the containment peak pressure tran-sient to 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 system 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 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 Amendment 4, 5 SEQUOYAH - UNIT 1 B 3/4 6-4 Revised 08/18/87

C_0f' Y4ENT SYSTEMS BAU '

event that observed sublimation rates are equal to or lower than design predic-tions 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.

3/4.6.5.2 ICE BED TEMPERATURE MONITORING SYSTEM The OPERABILITY of the ice bed temperature monitoring system ensures that 3 the capability is available for monitoring the ice temperature. 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 limit.s.

3/4.6.5.3 ICE CONDENSER D0 ORS 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 blow-down 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.

3/4.6.5.4 INLETD00RPOSITJpNMONITORINGSYSTEM The OPERABILITY of the inlet door position monitoring system ensures that l 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 The requirements for the divider Larrier 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 This condition ensures a diversion of the steam thrcugh the ice condenser

~

LOCA.

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.

SEQUOYAH - UNIT 1 B 3/4 6-5 Revised' 08/18/87

CONTAINMENT SYSTEMS BASES 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.

3/4.6.5.9 DIVIDER BARRIER SEAL The requirement for the divider barrier seel to be OPERABLE ensures 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 steam through the ice condenser bays that is consistent with the LOCA analyses.

3/4.6.6 VACUUM RELIEF VALVES The OPERABILITY of the primary containment to atmosphere vacuum relief valves ensures that the containment internal pressure does not become more negative than 0.1 psid. This condition is necessary to prevent exceeding the containment design limit for internal vacuum of 0.5 psid.

1 SEQUOYAH - UNIT 1 B 3/4 6-6 Revised 08/18/87

PLANT SYSTEMS  !

BASES 3/4.7.1.4 ACTIVITY i

l The limitations on secondary system specific activity ensure that the >

resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also 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 i with the assumptions used in the accident analyses.

l 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 i.. the event of a steam line j rupture. This restriction is required to 1) minimize the positive reactivity I effects of the Reactor Coolant System cooldown associated with the blowdown, j 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  ;

l valves within the closure times of the surveillance requirements are consistent j with the assumptions used in the accident analyses.  ;

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION l

! 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 RT f 25 F and are sufficient to NDT prevent brittle fracture j l

3/4.7.3 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the component cooling water system ensures that sufficient  !

cooling capacity is available for continued operation of safety related equipment l during normal and accident conditions. The redundant cooling capacity of this i system, assuming a single failure, is consistent with the assumptions used in j the accident analyses.

3/4.7.4 ESSENTIAL RAW COOLING WATER SYSTEM j The OPERABILITY of the essential raw 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 i capacity of this system, assuming a single failure, is consistent with the {

assumptions used in the accident conditions within acceptable limith I

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SEQU0YAH - UNIT 3 B 3/4 7-3 Amendment No. 12 Revised 08/18/87

4 PLANT SYSTEMS BASES 3/4.7.5 ULTIMATE HEAT SINK The limitations on temperature ensure that sufficient cooling capacity is l available to either 1) provide normal cooldown of the facility, or 2) to mitigate the effects of accident conditions within acceptable limits.

The limitations on the maxmum temperature are based on providing a 30 day l cooling water supply to safety related equipment without exceeding their design basis temperature and is consi. stent with the recommendations of Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Plants", March 1974.

3/4.7.6 FLOOD PROTECTION The requirements for flood protection ensures that facility protective actions will be taken and operation will be terminated in the event of flood conditions. A Stage 1 flood warning is issued when the water in the forebay is predicted to exceed 697 feet Mean Sea Level USGS datum during October 1 through April 15, or 703 Feet Mean Sea Level USGS datum during April 15 through September 30. A Stage II flood warning is issued when the water in the forebay is predicted to exceed 703 feet Mean Sea Level USGS datum. A maximum allowed water level of 703 Mean Sea Level USGS datum provides sufficient margin to ensure waves due to high winds cannot disrupt the flood mode preparation. A Stage I or Stage II flood warning requires the imple-mentation of procedures which include plant shutdown. Further, in the event of a loss of communications simultaneous with a critical combination flood, headwaters, and/or seismically induced dam failure the plant will be shutdown and flood protection measures implemented.

3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM The OPERABILITY of the control room ventilation system ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room will remain habitable for operations personnel during and following all credible accident conditions. The OPERABILITY of .

this system in conjunction with control room design provisions is based on  !

limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A", 10 CFR 50.

ANSI N510-1975 will be used as a procedural guide for surveillance testing.

SEQUOYAH - UNIT 1 B 3/4 7-4 Amendment No. 8

FACILITY OPERATING LICENSE NO. DPR-79 I

DOCKET NO. 50-328 i  !

i Revise the Appendix A Technical Specifications Bases by removing the pages I identified below and inserting the enclosed pages. The revised pages contain )

marginal lines indicating the area of change. Overleaf and repaginated pages* ,

are provided to maintain document completeness.

I REMOVE INSERT l B 2-1 B 2. 1 l

B 2-2 B 2-2* l l B 2-3 B 2-3

! B 2-4 B 2-4* i

! B 2-5 B 2-5* I B 2-6 B 2-6  !

B 2-7 B 2-7 I B 3/4 1-3 B 3/4 1-3 i B 3/4 1-4 B 3/4 1-4* i B 3/4 3-3 B 3/4 3-3  !

B 3/4 3-3a -- l B 3/4 3-4 B 3/4 3-4* .

B 3/4 5-3 B 3/4 5-3  !

B 3/4 6-1 B 3/4 6-1"  !

B 3/4 6-2 B 3/4 6-2 l B 3/4 6-3 8 3/4 6-3 l B 3/4 6-4 B 3/4 6-4 i B 3/4 6-5 B 3/4 6-5  !

B 3/4 6-Sa -- j B 3/4 6-6* l B 3/4 7-3 B 3/4 7-3 I B 3/4 7-4 B 3/4 7-4* l 1

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l 2.1 SAFETY LIMITS i BASES  !

i 2.1.1 REACTOR CORE t The restrictions of this Safety Limit prevent overheating of the fuel and  ;

possible cladding prforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. )

l Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been i related to DNB througn the W-3 correlation. The W-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The ininimum 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 confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curves of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

These curves are based on an enthalpy hot channel factor, F f 1.55 and a reference cosine with a pegk of 1.55 for axial power shape H,An allowance is included for an increase in F AH at reduced power based on the expression:  ;

F H = 1.55 [1+ 0.3 (1-P)]

where P is the fraction of RATED THERMAL POWER 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 f (delta I) function of the Overtemperature Delta T trip. When the axial power  !

ikbalance is not within the tolerance, the axial power imbalance effect on the  !

Overtemperature delta T trips will reduce the setpoints to provide protection consistent with core safety limits.

l SEQUOYAH - UNIT 2 B 2-1 Amendment No. 21 Revised 08/.18/87 1

_ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ w

SAFETY LIMITS .

BASES 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 piping, valves and fittings, are designec + 7 ANSI B 31.11967 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 func+ ;r il unit. The Trip Setpoints have been selected to ensure that the reacto ra a"d reactor coolant system are prevented from exceeding their safety li s- ung normal operation and design basis anticipated operational occurrences and tc assist the Engineered Safety Features Actuation System in mitigati.ig 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.

l l SEQUOYAH - UNIT 2 B 2-2

a f

l 1

1 1

2. 2 LIMITING SAFETY SYS',EM SETTINGS 1

BASES i Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

'l Power Range, Neutron Flux l 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 l redundant protection in the power range for a power excursion beginning from  !

l 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 l level of above appr;oximately 10 percent of RATED THERMAL POWER) and is auto-matically reinstated when P-10 becomes inactive (tnree of the four channels 0 indicate a power level below approximately 9 percent of RATED THERMAL POWER). l

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)

Power Range, Neutron Flux, High Rates i 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 Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power.

The Power Range Negative Rate trip provides protection to ensure that the  !

minimum 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 maintainca equivalent to turbine power by action of the automatic rod control system, could cause 1 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 i multiple dropped rods.

1 Intermediate and Source Range, Nuclear Flux The Intermediate and Source Range, Nuclear Flux tripu g ovide reactor core protection during reactor startup. These trips provide redundant protec-tion to the low setpoint trip of the Power Range, Neutron Flux 4 hannels. 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 SEQUOYAH - UNIT 2 B 2-3 Revised 08/18/87  !

I

- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - J

LIMITING SAFETY SYSTEM SETTINGS -

BASES Intermediate and Source Range, Nuclear Flux (Continued)

Range Channels will initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active. No credit was taken for operation of the trips associ-ated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System.

Overtemperature AT The Overtemperature delta T trip provides core protection to prevent DN8 for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),

and pressure is within the range between the High and Low Pressure reactor trips, This setpoint includes corrections for axial power distribution, changes in density and heat capacity of water with temperature and dynamic compensation for piping delays'from the core to the loop temperature detectors.

With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.

Operation with a reactor coolant loop out of service below the 4 loop P-8 setpoint does not require reactor protection system set point modification because the P-8 setpoint and associated trip will prevent DNB during 3 loop operation exclusive of the Overtemperature delta T setpoint. Three loop operation above the 4 loop P-8 setpoint is permissible after resetting the K1, K2, and K3 inputs to the Overtemperature delta T channels and raising the P-8 setpoint to its 3 loop value. In this mode of operation, the P-8 interlock and trip functions as a High Neutron Flux trip at the reduced power level.

Overpower AT The Overpower delta T reactor trip provides assurance of fuel integrity, e.g., no melting, under all possible overpower conditions, limits the required range for Overtemperature delta T protection, and provides a backup to the High Neutron Flux trip. The setpoint includes corrections for changes in density and heat capacity of water with temperature, and dynamic compensation for piping delays fro,n the core to the loop temperature detectors. No credit l was taken for operation of this trip in the accident analyses; however, its i functional capability at tt.c specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

8 SEQUOYAH - UNIT 2 B 2-4

i LIMITING SAFETY SYSTEM SETTINGS 1

BASES Pressurizer Pressure The Pressurizer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted. The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig). The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure.

Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves. No credit was taken'for operation of this trip 'in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

1 Loss of Flow

.l The Loss of Flow trips provide core protection to prevent DNB in the event of a loss of one or more reactor coolant pumps.

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Above 11 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drop below 89% of nominal full loop flow.

Above 36% (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 83% of nominal full loop flow. This latter trip will prevent the minimum value of the DNBR from going below 1.30 during normal operational transients and anticipated transients when 3 loops are in operation and the Overtemperature delta T trip set point is adjusted to the value specified for all loops in operation. With the Overtemperature delta T trip set point adjusted to the value specified for 3 loop operation, the P-8 trip at 76% RATED THERMAL POWER will prevent the minimum value of the DNBR from going below 1.30 during normal operational transients and anticipated transients with 3 loops in operation.

SEQUOYAH - UNIT 2 B 2-5

j LIMITING SAFETY SYSTEM SETTINGS .

BASES l

Steam Generator Water Level The Steam Generator Water Level Low-Low trip provides core protection by pieventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity. The specified setpoint l provides allowance that tOere will be sufficient water inventory in the steam  !

generators at the time of trip to allow for starting delays of the auxiliary I 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 i equal to 1.5 x 106 lbs/ hour. The Steam Generator Low Water level portion of l l the trip is activated when the water level drops below 24 percent, as indicated l by the narrow range instrument. These trip values include sufficient allowance 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.

Undervoltage and Underfrequency - Reactor Coolant Pump Busses The Undervoltage 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 j 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 shall not exceed 1.2 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 0.6 seconds.

t i

i SEQUOYAH - UNIT 2 B 2-6 Revised - 08/18/87

LIMITING SAFETY SYSTEM SETTINGS BASES Turbine Trip A Turbine Trip causes a direct reactor trip when operating above P-9.

Each of the turbine trips provide turbine protection and reduce the severity of the ensuing transient. 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 l l

The Reactor Trip System Interlocks perform the following functions on increasing power:

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 ferdwater valves on T below setnoint, prevents the opening o-T 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.

Reactor not tripped - defeats manual block preventing automatic reactuation of safety injection.

SEQUOYAH - UNIT 2 B 2-7 Revised 08/18/87-

REACTIVITY CONTROL SYSTEMS BASES l

l BORATION SYSTEMS (Continued) i 1

provide a SHUTDOWN MARGIN from expected operating conditions of 1.6% delta k/k after xenon decay and couldown to 200 F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon l conditions and requires 5408 gallons of 20,000 ppm 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 available because of discharge line location and other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.5 and 9.5 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 associaLed 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.

SEQUOYAH - UNIT 2 B 3/4 1-3 Revised 08/18/87

REACTIVITY CONTROL SYSTEMS BASES MOVEABLE CONTROL ASSEMBLIES (Continued)

The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of peaking factors and a restriction in THERMAL POWER. These restrictions provide assurance of fuel rod integrity during continued operation. In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.

The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with T greater than or avg equal to 541 F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LC0's are satisfied.

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SEQUOYAH - UNIT 2 B 3/4 1-4

INS'IRUMENTATION BASES 3/4.3.3.3 SEISMIC INSTRUMENTATION (Continued) recommendations of Regulatory Guide 1.12, " Instrumentation for Earthquakes,"

t Ap'ril 1974.

3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteoro bgical instrumentation ensures that sufficient meteorological data ia r.svailable for estimating potential radiation  !

doses to the public as a result of routine or accidental release of radioactive i materials to the atmosphere. This capability is required to evaluate the need  !

for initiating protective measures to protect the health and safety of the i 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 I 1

The OPERABILITY of the remote shutdown instrumentation ensures that suf-ficient capability is available to permit shutdown and maintenance of HOT ,

SiANDBY of the facility and the potential capability for subsequent cold shut-l down from locations outside of the control room. This capability is required ,

i in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR 50.

3/4.3.3.6 CHLORINE DETECTION SYSTEMS l The OPERABILITY of the chlorine detection system ensures that sufficient capability is available to promptly detect and initiate protective action in l the event of an accidental chlorine release. This capability.is required to protect control room personnel and is consistent with the recommendations of l

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 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 Following an Accident," December 1975.

Sequoyah has four separate methods of determining safety valve position (i.e.,

open or closed).

a. Acoustic flow monicors 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.

SEQUOYAH - UNIT 2- B 3/4 3-3 AmendmentNo.)$,46 Revised 08/18/87 L _

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l INSTRUMENTATION

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1 3/4.3.3.7 ACCIDENT MONITORING INSTRUMENTATION (continued) i

b. Temperature sensors downstream of each safety valve (one per valve). Tem-perature indication and alarm are provided in the main control room. {;

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Pressurizer relief tank temperature, pressure and level indication, and alarm in main control room. i l

d. Pressurizer pressure indication and alarm in the main control room. i Although all the above position indicators for the pressurizer safety valves and the PORVs are acceptable as one of the channels, the acoustic monitors must be one of the two required operable chanels. In addition to the four methods i; described above, the PORVs use an electromagnetic " reed"-switch to determine i valve position. The stem mounted switches are no longer in use since the PORVs I were changed. ,

l 3/4.3.3.8 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages. Prompt detection of fires will reduce the potential for damage to safety related equipment and is an integral element in the overa'il facility fire protection program.

In the event that a portion of the fire detection instrumentation is inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

3/4.3.3.9 RADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATION The radioactive liquid cffluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effuents during actual or potential releases of liquid effluents. The alarm /

trip setpoints for these instruments shall be calculated in accordance with l the procedures in the ODCM to ensure that the alarm / trip will occur prior to l

exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

3/4.3.3.10 RADI0 ACTIVE GASE0US EFFLUENT INSTRUMENTATION-The radioactive gaseous effluent instrumentation is provided to monitor l

and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments shall be calculated in accordance with the procedures in the ODCM to ensure that the alarm /tr'ip will occur prior SEQUOYAH - UNIT 2 B 3/4 3-4 AmendmentNo.)$,46 l

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l EMERGENCY CORE COOLING SYSTEMS I I

BASES REFUELING WATER STORAGE TANK (Continued)

RWST and the RCS water volumes with all control rods inserted except for the l 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 location or other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.5 and 9.5 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.

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SEQUOYAH - UNIT 2 B 3/4 5-3 Revised 08/18/87

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1 3/4.6 CONTAINMENT SYSTEMS l I

BASES l

l 3/4.6.1 PRIMARY CONTAINMENT 1

1 3/4.6.1 PRIMARY CONTAINMENT i 1

Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage ,

paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during I accident conditions.

3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the accident

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analyses at the peak accident pressure, Pa . As an added conservatism, the j measured overall integrated leakage rate is further limited to less than or equal to 0.75 La during performance of the periodic tests to account for i possible degradation of the c.ontainment leakage barriers tetween leakage tests.

l The surveillance testing for measuring leakage rates are consistent with the requirements of Appendix "J" of 10 CFR 50.

1 3/4.6.1.3 CONTAINMENT AIR LOCKS  !

1 The limitations on closure and leak rate for the containment air locks  !

are required to meet the restrictions on CONTAINMENT INTEGRITY and containment j leak rate. Surveillance testing of the air lock seals provide assurance that I the overall air lock leakage will not become excessive due to seal damage  !

during the intervals between air lock leakage tests. l l

3/4.6.1.4 INTERNAL PRES $URE The limitations on containment internal pressure ensure that 1) the I containment structure is prevented from exceeding its design negative pressure differential with respect to the annulus atmosphere of 0.5 psig and 2) the 1

SEQU0YAH - UNIT 2 B 3/4 6-1 j I

'j CONTAINMENT SYSTEMS BASES q

l INTERNAL PRESSURE (Continued) containment peak pressure does not exceed the maximum allowable internal pressure of 12 psig during LOCA conditions.

1 3/4.6.1.5 AIR TEMPERATURE The limitations on containment average. air temperature ensure that 1) the containment air mass is limited to an initial mass sufficiently low to prevent  :

exceeding the maximum allowable internal pressure during LOCA conditions and 1' 2).the ambient air temperature does not exceed that temperature allowable for the continuous duty 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 equal to'5% of RATED THERMAL POWER will limit the peak pressure to an acceptable value. The upper temperature 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  !

l This limitation ensures that the structural integrity of the containment I steel vessel will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that i the vessel will withstand the maximum pressure of 12 psig in the event of a LOCA.

A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.

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3/4.6.1.7 SHIELD BUILDING STRUCTURAL 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 rf the facility. Structural integrity is required to provide '

l 1)' protection for the steel vessel from external missiles, 2) radiation shield-

! ing in the esent of a LOCA, and 3) an annulus surrounding the steel vessel that can be ;naintained at a negative pressure during accident conditions.

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l SEQUOYAH - UNIT 2 B 3/4 6-2 Revised 08/18/87

a CONTAINMENT SYSTEMS BASES 3/4.6.1.8 EMERGENCY GAS TREATMENT SYSTEM (EGTS)

The OPERABILITY of the EGTS cleanup subsystem ensures that during LUCA 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 conditions. Cumulative 'peration of the system with the heaters on for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> over a 31 day period is sufficient to reduce the buildup of moisture on the absorbers 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 containment purge lines is restricted to only one pair (one l 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.

3/4.6.2 DEPRESSURIZATION AND C0OLING 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.

3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the 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 = trial to the environment will be consistent with the assumptions used in the anaij us for a LOCA.

i 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 limit during post-LOCA conditions. Either recombiner unit or the hydrogen SEQUOYAH - UNIT 2 B 3/4 6-3 Revised 08/18/87

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CONTAINMENT SYSTEMS BASES COMBUSTIBLE GAS CONTROL (Continued) mitigation system, consisting of 68 hydrogen igniter! per unit, is capable of controlling the expected hydrogen generation associated with 1) zirconium-water reactions, 2) radiolytic decomposition of water and 5) corrosion of metals within containment. These hydrogen control systems are designed to mitigate the effects of an accident as described in Gegulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a LOCA," Revision 2, dated November 1978.

The hydrogen mixing systems are provided to enwre adequate mixing of the containment atmosphere following a LOCA. This mixing action will prevcnt localized accumulations of hydrogen from exceeding the flammable limit.

The operability of at least 66 of 68 igniters in the hydrogen control distributed ignition system will maintain an effective coverage throughout the 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 ensure that the overall system will be available to provide sufficient pressure suppression capability to limit the containment peak pressure transient to 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 system volume released during a LOCA. These conditions are consiste.nt 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 minimum weight figure of 2,333,100 pounds of ice also contains an additional 1% conservative i allowance 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.

SEQUOYAH - UNIT 2 B 3/4 6-4 Amendment No. 21 ,

Revised 08/18/87

CONTAINMENT SYSTEMS BASES l

3/4.6.5.2 ICE BED TEMPERATURE MONITORING 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 the ice bed heat remcial capacity will be retained within the specified time limits. 3 i

1 3/4.6.5.3 ICE CONDENSER D0 ORS The OPERABILITY of the ice condenser doors ensures that these doors will open because of the differential pressure between upper and lower containment I resulting from the blowdown of reactor coolant during a LOCA and that the blow-down will be diverted through the ice condenser bays for heat removal and thus-containment pressure control. The requirement that the doors be maintained l closed during normal operation ensures that excessive sublimation of the ice  ;

will not occur because of warm air intrusion from the lower containment.

l 3/4.6.5.4 INLET DOOR 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 l 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 j system and 2) the accumulation of hydrogen in localized p6rtions of the contain-l ment structure is minimized.

l 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 l SEQUOYAH - UNIT 2 B 3/4 6-5 Revised 08/18/87 l

CONTAINMENT SYSTEMS .

BASES I

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, 3/4.6.5.9 DIVIDER BARRIER SEAL j The requirement fia the divider barrier seal to be OPERABLE ensures 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 steam through the ice condenser bays that is consistent with the LOCA analyses.  ;

3/4.6.6 VACUUM RELIEF VALVES i

The OPERABILITY of the primary containment to atmosphere vacuum q relief valves ensures that the containment internal pressure does not become j more negative than 0.1 psid. This condition is necessary to prevent exceeding {

the containment design limit for internal vacuum of 0.5 psid, '

l SEQUOYAH - UNIT 2 B 3/4 6-6 Revised 08/18/87

a PLANT SYSTEMS BASES 3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES 1

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 i valves within the closure times of the surveillance requirements are consistent l with the assuinptions used in the accident analyses. I I

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION 1

The limitation on steam generator pressure and temperature ensures that l 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 RT NDT f 25 F and are sufficient to prevent ,

brittle fracture. l l

3/4.7.3 COMPONENT COOLING WATER SYSTEM l l

The OPERABILITY of the component cooling water system ensures that j suf ficient 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 raw 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 Revised 08/18/87

PLANT SYSTEMS .'

BASES 3/4.7.5 ULTIMATE HEAT SINK The limitations on the ultimate heat sink temperature ensures that sufficient cooling capacity is available to either 1) provide normal cooldown of the facility, or 2) to mitigate the effects 'af accident conditions within acceptable limits.

The limitation on maximum temperature is based on providing a 30 day  !

cooling water supply to safety related equipment without exceeding their design basis temperature and is consistent with the recommendations of Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Plants", March 1974.

3/4.7.6 FLOOD PROTECTION The requirements for flood protection ensures that facility protective actions will be taken and operation will be terminated in the event of flood conditions. A Stage 1 flood warning is issued when the water in the forebay is predicted to exceed 697 feet Mean Sea Level USGS datum during October 1 through April 15, or 703 Feet Mean Sea Level USGS datum during April 15 through September 30. A Stage II flood warning is issued when the water in the forebay is predicted to exceed 703 feet Mean Sea Level USGS datum. A maximum allowed water level of 703 Mean Sea Level USGS datum provides sufficient margin to ensure waves due to high winds cannot disrupt the flood mode preparation. A Stage I or Stage 11 flood warning requires the implementation of procedures which include plant shutocwn. Further, in the event of a loss of communications simultaneous with a critical combination flood, headwaters, and/or seismically induced dam failure the plant will be shutdown and flood protection measures l implemented.

l 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM The OPERABILITY of the control room ventilation system ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this l system and 2) the control room will remain habitable for operations personnel during and following all credible accident conditions. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A", 10 CFR 50.

ANSI N510-1975 will be used as a procedural guide for surveillance testing.

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SEQUOYAH - UNIT 2 B 3/4 7-4 l