Text

Amend 155 to License DPR-77,replacing Certain Cycle Specific Parameter Limit Values in TS W/Refs to Core Operating Limits Rept,Per Generic Ltr 88-16
	ML20085M766
Person / Time
Site:	Sequoyah
Issue date:	10/23/1991
From:	Hebdon F; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20085M771	List: ML20085M766; ML20085M773;
References
	GL-88-16, NUDOCS 9111120157
	Download: ML20085M766 (36)
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%E NUCLEAR REGULATORY COMMISSION-i UNITED STATES -

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WASHINGTON. D.C. 20S55 l

TENNESSEE-VALLEY AllTHGRITY DOCKET NO. 50-327 SEQUOYAH NtlCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERfTING LICENSE Amendment No.155 License-No. DPR-77 1.

The Nuclear Regulatory Comission (the Comission)= ha. found that:

A.

The application-for arendment by Tennessee Valley Authority (the '

licensee) dated May 24, 1991 and amplified August 23,-1991, complies with the standards and requirements of the Atomic Energy Lact of 1954, as amended (the Act), and.the Comission's rules and ' regulations set forth in 10 CFR Chapter I; f;.

The facility will operate in conformity with_the applicationg-the-provisions of the Act, and the rules and-regulations of-the Comissien;-

C.

There is reasonable. assurance (i) that the activities authorized by l

this amendment car br conducted without endangering _the health and schty of. the public, ard (ii)-that such activities will be conducted in compliance with the Comission's regulations; D.

The-issuance of this amendment will'not be' inimical to the comon defense and security or. to the-health.and safety of the public; and E.

The issuance of this amendment'is in'accordance with 10 CFR Part of the Comission's regulations 'and all applicable' requirements. have been satisfied.

i 911112C157 911023 PDR ADCCK 05000327 p

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Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-77 is hereby amended to read as follows:

(2) Tecnnical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 155, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuence; to be.

implemented in conjunctio'1 with the Core Operating Limits Report within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION-p hw

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Frederick J. Hebdbn, Director:

Project Directorate II-4 Division of Reactor _ Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

i Changes to the Technical Specifications Date of Issuance:

October 23, 1991 l

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ATTACHMENT TO LICENSE AMEN 0 MENT NO.155 FACILITY OPERATING LICENSE NO. DPR-77 DOCKET NO. 50-327 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. Overleaf pages* are provided to r.aintain document completeness.

REMOVE INSERT I

I II II 1-?

1-2 1-3 1-3 1-4 1-4 1-5 1-5 1-6 1 1 -7 1-7 3/4 1-4 3/4 1-4 3/4 1-5 3/4 1-5 3/4 1-14 3/4 1-14 3/4 1-20 3/4 1-20 3/4 1-21 3/4 1-21 3/4 1-22 3/4 1-22 3/4 1-23 3/4 1-23 3/4 2-1 3/4 2-1 3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3/4 2-9 3/4 2-9 3/4 2-10 3/4 2-10 B 3/4 1-2 B 3/4 1-2 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B 3/4 2-4 B 3/4 2-4 6-21 6-21 6-21a 3 2-1 B 2-1 B 3/4 0-1 B 3/4 0-1 B 3/4 5-1 B 3/4 5-1 B 3/4-7-2 8 3/4-7-2 B 3/4 7-2a

INDEX DEFINITIONS SECTION 1.0 DEFINITIONS 1.1 ACTION...

1-1 1.2 AXIAL FLUX DIFFERENCE.......................................

1-1

1. 3 BYPASS LEAKAGE PATH.....................................

1-1 1.4 CHANNEL CALIBRATION............................................

1-1 1.5 CHANNEL CHECK................................................

1-1 1.6 CHANNEL FUNCTIONAL TEST...........................................

1-2

1. 7 CONTAINMENT INTEGRITY.............................................

1-2 1.8 CONTROLLED LEAKAGE..................

1-2 1.9 CORE ALTERATION........

1-2 1.10 CORE OPERATING LIMITS REP 0RT........................

1-2 1.11 DOSE EQUIVALENT I-131.......

1-3 1.12 l-AVERAGE DISINTEGRATION ENERGY....

1-3 1.13 ENGINEERED SAFETY FEATURE RESPONSE TIME...........................

1-3 1.14 FREQUENCY NOTATION..............................................

1-3 1.15 GASEOUS RADWASTE TREATMENT SYSTEM.....,........................

1-3 1.16 IDENTIFIED LEAKAGE................................................

1-3 1.17 MEMBERS OF THE PUBLIC...........

1-4 1.18 0FFSITE DOSE CALCilLATION MANUAL...................................

1-4 1.19 OPERABLE - OPERABILITY..........................

1-4 1.20 OPERATIONAL MODE - M0DE...........................................

1-4 1.21 PHYSICS TESTS.....................................................

1-4 1.22 PRESSURE BOUNDARY LEAKAGE.........................................

1-5 1.23 PROCESS CONTROL PROGRAM.............

1-5 SEQUOYAH - UNIT 1 I

Amendment No.

71, 155

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INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS (Continued) 1.24 PURGE-PURGING..

1-5 1.25 QUADRANT POWER TILT RATIO..........

1-5 1.26 RATED THERMAL POWER... 5 1.27 REACTOR TRIP SYSTEM RESPONSE TIME.................................

1-5 1.28 REPORTABLE EVENT..........

1-5 1.29 SHIELO BUILDING INTEGRITY..................

1-6 1.30 SHUTDOWN MARGIN...................................................

1-6 1.31 SITE B0VNDARY...................

1-6 1.32 SOLIDIFICATION.....................

1-6 1.33 SOURCE CHECK.....................................................

1-6 1.34 STAGGERED TEST BASIS.........

1-6 1.35 THERMAL POWER....................................................

1-6 1.36 UNIDENTIFIED LEAKAGE..............................................

1-7 1.37 UNRESTRICTED AREA.................................................

1-)

1.38 VENTILATION EXHAUST TREATMENT SYSTEM..............................

1-7 1.39 VENTING...........................................................-

1-7 OPERATIONAL MODES (TABLE 1.1)............

1-8 FREQUENCY NOTATION (TABLE 1.2)......

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SEQUOYAH - UNIT 1 II Amendment No.

71, 155

DEFINITIONS CHANNEL FUNCTIONAL TEST

1. 6 A CHANNEL FUNCTIONAL TEST shall be:

Analog channels - the injection of a simulated signal into the a.

channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

b.

Bistable channels - the injection of a simulated signal into the sen3or to verify 0PERABILITY including alarm and/or trip functions.

Digital channels - the injection of a simulated signal into the c.

channel as close to the sensor ',nput to the process racks as practicable to verify OPERABILITY including alarm and/or trip functions.

CONTAINMENT INTEGRITY

1. 7 CONTAINMENT INTEGRITY shall exist when:

All penetrations required to be closed during accident conditions a.

are either:

1)

Capable of being closed by an OPERABLE containment autoinatic isolation valve system, or 2)

Closed by manual valves, blind flanges, or deact'ivated automatic valves secured in their closed positions, except as provided in-Table 3.6-2 of Specification 3.6.3.

b.

All equipment hatches are closed and sealed.

Each air lock is in compliance with the requirements of c

Specification 3.6.1.3, d.

The containmer.t leakage rates are within the limits of Specification 3.6.1.2, and The sealing mechansim associated with each penetration (e.g.,

e.

welds, bellows, or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE

1. 8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATION

1. 9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel-in the vessel.

Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMIT REPORT 1.10 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the cur ent operating reload cycle. These cycle-specific core operating limits shall be determined-for each reload cycle in accordance with Specification 6.9.1.14.

Unit operation within these operating limits is addressed in individual specifications.

SEQUOYAH - UNIT 1-1-2 Amendment No. 12, 71, 130, 141, 155 e

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DOSE EQUIVALENT I-131 1.11 0032 EQUIVALENT I-131 shall be that concentration of I-131 (microcurie /

gram) which alone would produce the sarne thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID '4844, " Calculation of Distance Factors for Power and Test Reactor Sites."

5 - AVERAGE DISINTEGRATION ENE?GY 1.12 E shall be the average (weighted in proportion to the concentration of l

each radionuclide in the reactor coolant at the time of sampling) of the tum of the average beta and gamma energies per disintegration (in MeV) for i',otopes, other than iodines, with half lives greater than 15 minutes, making up t.t least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval l

from when the monitored parameter exceeds its ESF actuation setpoint et the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.14 The FREQUENCY NOTATION specified for the performance of Surveillance l

Requirements shall correspond to the intervals defined.in Table 1.2.

i GASE0US RADWASTE TREATMENT SYSTEM 1.15 A GASEOUS RADWASTE TP.EATMENT SYSTEM is any system designed and installed j

to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of redu ing the total radioactivity prior to release to the ei,vironment.

IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAKAGE shall be:

Leakage (except CONTROLLED LEAKAGE) into closed systems, such a.

as rump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or a

SEQUOYA.. - UNIT 1 1-3 Amendment No. 12, 71, 155

b.

Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE COUNDARY LEAKAGE, or Reactor coolant system leakage through a steam generator to c.

the secondary system.

MEMBER (5) 0F THE PUBLIC 1.17 MEMBERS OF THE PUBLIC shall include all individuals who are not l

occupationally associated with the plant.

This category shall include non-employees of the licensee who are permitted to use portions of the site for recreational, occupational, or other purposes not associated with plant functions.

This category does not include non-employees such as vending machine servicemen or postmen who, as part of their. formal job function,_

occasionally enter an area that i controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

OFFSITE DOSE CALCULATION MANUAL (00CM) 1.18 The 0FFSITE DOSE CALCULATION MANUAL (0DCM) shall'contain the methodology l

and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program.

The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.5 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semiannual Radioactive Effluent Release Reports required by Specifications 6.9.1.6 and 6.9.1.8.

OPERABLE - OPERABILITY 1.19 A system, subsystem, train, or component or device shall be OPERABLE or I

have OPERABILITY when it is capable of-performing its specified function (s), and when all necessary attendant instrumentation, controls, a normal and an emergency I

electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or-device to perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive l

combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

PHYSICS TESTS

1. ?1 PHYSICS TESTS shall be those tests perf' id to measure the fundamental nuclear characteristics of the reactor core an.

elated instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

SEQUOYAH - UNIT 1 1-4 Amendment No. 12, 71, 148, 155 e -

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PRESSliRE B0UNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGL 3 hall be leakage (except steam generator tube i

leakage) throtto a non-isoiable fault in a Reactor Coolant System component body, pipe wa:

or vessel wall.

PROCESS CONTROL PROGRAM (PCP) 1.23 The PRCCESS CONTROL PROGRAM shall cont >.

the current formulas, sampling, l

analyses, tests, and determinations to be made to ensure that the processinn and packaging of solid radioactive wastes based on demonstrated processing of tc;ual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with Id CFR i' arts 20, 61, and 71; State regulations; and other requirements goverr.ing the disposal of solid radioactive wastes.

PURGE - PURGING 1.24 PURGE or PURGING is the controlled process of discharging air or gas

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from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

Qll8 M ANT POWER IILT RATIO 1.25 QUADRANT DOWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

With one '

ore detector inoperable, the remaining three detectors shall be used for puting the average.

RATED THERMAL POWER (RTP) 1.26 RATED THERMA! F0WER (RTP) shall be a total reactor core heat transfer l

rate to the reactor coolant of 3411 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.27 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor unti loss of stationary gripper ca" voltap REPORTABLE EVE _NT 1.28 A REPORTABLE EVENT shall be any

,e conditions specified in l

Section 59.73 to 10 CFR Part 50.

t Su, 'AH - UNIT 1 1-5 Amendment No. 12, 71, 141, 148, 155

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SHIELD BUILDING INTEGRITY

1. 2P SHIELD BUILDING INTEGRITY shall exist when:

The door in each access opening is closed except when the a.

access opening is being used for normal transit entry and exit.

b.

The emergency gas treatment system is OPERABLE.

L The sealing mechanism ar - " ated with cach penetration (e.g.,

c.

welds, bellows or 0 ringt s OPERABLE.

SHUTDOWN MARGIN 1.30 SHUTDOWN MARGIN shall be toe instantaneous amount of reactivity by which the reactor is subtritical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY 1.31 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee (see Figure 5.1-1).

SOLIDIFICATION 1.32 Deleted l

SOURCE CHECK 1.33 Deleted j

STAGGERED TEST BASIS 1.34 A STAGGERED TEST BASIS hall consist of:

A test schedule for n systems, subsystems, trains or other designated a.

components obtained by dividing the specified test interval into n equal subintervals, b.

The testing of one system, subsystem, train or other designated I

component at the beginning of each subinterval.

THERMAL POWER 1.35 THERMAL POWER shall be the total reactor core heat transfer rate to the veactor coolant.

SEQUOYAH - UNIT 1 1-6 Amendment No. 12, 71, 148, 155

N UNIDENTIFIED LEAKAGE 1.36 UNIDENTIFIED LEAKAGE shall be al' leakage which is not IDENTIFIED LEAKAGE l

or CONTROLLED LEAKAGE.

UNRESTRICTED AREA 1.37 An UNRESTRICTED AREA shall be any area, at or beyond the site boundary l

to which access is not contrclied by the licensee for purposes of protection of individuals f rom exposure to radiation and radioactive materials or any area within the site boundary used for residential quarters or industrial, commerical, institutional, and/or recreational purposes.

VENTILATION EXHAUST TREATMENT SYSTEM 1.38 a VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radiciodine or radioactive material in particulate l

form in effluents by passing ventilation or vent exhaust gases through charcoal adsorberc and/or HEPA filters for the purpose of removing iodines or parti-culates from the gaseous exhaust stream prier to the release to the environment (such a system is not considered to have any effect on noble gas effluents).

Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING 1.39 VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or l

other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING.

Vent, used in system names, does not imply a VENTING process.

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t SEQUOYAH - UNIT 1 1-7 Amendment No. aa, 71, 155

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1. 3 The moderator temperature coefficient (MTC) shall be within the limits specified in the COLR.

The maximum upper limit shall be less then 0 delta k/k/ F.

APPLICABILITY:

Beginning of cycle life (BOL) limit - MODES 1 and 2* only#

End of cycle life (E0L) limit - MODES 1, 2 and 3 only#

ACTION:

With the MTC more positive than the BOL limit specified in the COLR a.

operation in MODES 1 and 2 may proceed provided:

1.

Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than the BOL limit specified in the COLR within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6.

2.

The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods 4

withdrawn condition.

3.

In lieu of any other report required by Specification 6.6.1, a Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the mea >ured MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

b.

With the MTC more negative than the EOL limit specified in the COLR, l

be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With Keff greater than or equal to 1.0

See Special Test Exception 3.10.3 SEQUOYAH - UNIT 1 3/4 1-4 Amendment No. 36, 155

REACTIVITY CONTROL SYSTEMS S,URVEILLANCE REQUIREMENTS 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows:

The MTC shall be measured and compared to the BOL limit specified in a.

the COLR prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading, b.

The MTC shall be measured at any THERMAL POWER and compared to the 300 ppm surveillance limit specified in the COLR (all rods withdrawn, l

RATED THERMAL POWER condition) within 7 EFPD after reaching an equi-librium boron concentration of 300 ppm.

In the event this comparison indicates that MTC is more negative than the 300 ppm surveillance limit specified in the COLR, the MTC shall be remeasured and compared to the E0L MTC limit specified in the COLR at least once per 14 EFP0 during the remainder of the fuel cycle.

SEQUOYAH - UNIT 1 3/4 1-5 Amendment No.155

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REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All fui: length (shutdown ai.d control) rods shall be OPERABLE and positioned within i 12 steps (indicated position) of their group step counter demand position.

APPLICABILITY-MODES 1* and 2*

ACTION:

With one or more full length rods inoperable due to being immovable a.

as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.

With more than one full length rod inoperable or misaligned from the group step counter demand position by more than i 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Witt one full length rod inoperable due to causes other than addressed c.

by ACTION a, above, or misaligned from its group step counter demand height by more than i 12 steps (indicated position), POWER OPERATION may continue provided that within or.e hour either:

1.

The rod is restored to OPERABLE status within the above alignment requirements, 2.

The remainder of the rods in the group with the inoperable rod are aligned to withir.

12 steps of thr inoperable rod within one hour while maintaining the rod sequence and insertion limit of specification 3.1.3.6.

The THERMAL POWER level shall be l

restricted pursuant to Specification 3.1.3.6 during subsequent operation, or 3.

The rod is-declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied.

POWER OPERATION may then continue provided that:

See Special Test Exceptions 3.10.2 and 3.10.3.

SEQUOYAH - UNIT 1 3/4 1-14 Amendment No. 114, 155'

E-REACTIVITY CONTROL SYSTEMS SHUTOOWN R0D INSERTION LIMIT LIMITING CONDITION FOR OPERATION s

3.1. 3. 5 All shutdown rods shali be limited in physical insertion as specified in the COLR.

i APPLICABILITY:

MODES 1* and 2*#

ACTION:

With a maximum of one shutdown rod inserted beyond the insertion limit speci-tied in the COLR, except for surveillance testing pursuant to Specification 4.1.3.1. 2, within one hour eit kar:

a.

Restore the rod to e the insertion limit specified in the CCLR, or b.

Declare the rod to be inoperable and apply Specification 3.1.3.1.

SURVEILLANCE REQUIREMENTS 4.1.3.5 EacS shutdown rod shall be determined to be within the insertion limit specified in the COLR:

Within 15 minutes prior to withdrawal of any rods in control a.

banks A, B, C or D during an approach to reactor criticality, and b.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter, k

See Special Test Exceptions 3.10.2 and 3.10.3.

With K,7f greater than or equal to 1.0.

SEQUOYAH - UNIT 1

.3/4 1-20 Amendment No. 108,-155

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REACTIVITY CONTROL SYSTEMS CONTROL R00 INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1. 3. 6 The control banks shall be limited in physical insertion as specified in the COLR.

APPLICABILITY:

MODES 1* and 2*#.

ACTION:

With the control banks inserted beyond the insertion limits, except for sur-l veillance testing pursuant to Specification 4.1.3.1.2, either:

Restore the control banks to within the limits within a.

two hours, or b.

Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THEPMAL POWER which is allowed by the group position using the insertion limits specified in the COLR, or I

i c.

Be in HOT STAND 8Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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See Special Test Exceptions 3.10.2 and 3.10.3.

With K,7f greater than or equal to 1.0.

SEQUOYAH - UNIT 1 3/4 1-21 Amendment No. 41, 114, 155 i

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. 3/4 1-22 Amendment No. 108, 155 t

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- SEQUOYAH - UNIT 1 3/4 1-23 Amendment No. 41, 108, 155

3/4.2 POWER OISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) w LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FL'X DIFFERENCE (AFD) shall be maintained within J

the limits specified in the COLR.

l APPLICABILITY:

MODE 1 above 50% RATED THERMAL P0WER*

ACTION:

With the indicated AXIAL FLbX DIFFERENCE outside of the limits speci-a.

fied in the COLR; 1.

Either restore the indicated AFD to within the limits within i

15 minutes, or 3

2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 55 percent of RATED THERMAL F0WER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is withite the limits specified in the l

COLR.

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SEQUOYAH -- UNIT 1 3/4'2-1 Amendment No. 19, 155'

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l SEQUOYAH - UNIT 1 3/4 2-4 Amendment No.19,155

POWER DISTRIBUTION LIMITS 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION 3.2.2 F (z) shall be limited by the following relationships:

9 F (z) 1 [F

] [K(z)] for P > 0.5 9

P F (z) 1 [F TP] [K(z)] for P $ 0.5 9

0.5 whereFhTP

= the F limit at RATED THERMAL POWER (RTP) 9 specified in the COLR, THERMAL POWER

, and p,

RATED THERMAL POWER K(z) = the normalized F (z) as a function of core height speci-fied in the COLR.

9

)

APP'ICABILITY:

MODE 1 ACTION:

With F (z) exceeding its limit:

9 Reduce THERMAL POWER at least 1% for each 1% F (z) exceeds the limit a.

q within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower Delta T Trip Setpoints (value of K ) have been reduced at least 1% (in AT span) for each 1% F (z) 4 q

exceeds tne lim't.

b.

Ider.tify and correct the cause of'the out of limit condition prior to increasing THERMAL POWEh; THERMAL POWER may then be increased provided F (z) is demonstrated thrcugh incnre mapping to be within its limit.9 1

SURVEILLANCE REQ!!IREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

SEQUOYAH - UNIT 1 3/4 2-5 Amendment No. 19, 95, 140, 155

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.2.2 F (z) shall be evaluated to determine if F (2) is within its q

limit by:

q a.

Using the movable incore detectors o'otain a power distribu-tion map at any THERMAL POWER greater aan 5% of RATED THERMAL.

POWER.

b.

Increasing the measured F (z) component of the power distribution q

map by 3 percent to account for manufacturing tolerances and further increasing the value by 5% to account for mewsrament uncertainties.

c.

Satisfying the following relationship:

F[(z)1F RTP x K(z) for P > 0.5 P x W(z)

M RTP Fq (z)

F x v.( z )

for P $ 0.5 W(z) x 0.5 M

where F (z) is the measured F (z) increased by the allowances for q

q manufacturing tolerances and measurement uncertainty, FRTP

.,3 q

the F limit, .(d is the normalized F (z) as a function of core q

q height, P is the relative THERMAL POWER,and W(z) is the cycle dependent function that accoa ts for power distribution transients TP encountered during normal operation.

F X(z), and W(z) are speci-fied in the COLR as per Specification 6.9.1.14.

Measuring F[(z) according to the following schedule:

d.

1.

Upon achieving equilibrium conditions after exceeding by 10 peicent or more of RATED THERMAL POWER, the THERMAL POWER at which F (z) was last determined,* or q

2.

At least once per 31 effective full power days, whichever occurc first.

l

0uring power escalation at the heginning of each cycle, power level may be increased until a po.ver level for extended operation has been achieved and a power distribution map obtained, l

SEQUOYAH - UNIT 1 3/4 2-6 Amendment No. 19, 95, 140, 155

I' POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (C.3ntic.ued)

With measurements indicating e.

F (z) maximum over z K(z) has increased since the previous determinatin of F N(z) either of the following actions shall be taken:

0 1.

F (z) shall be ircreased by 2 percent over that specified in q

4.2.2.2.c, or

"(z) shall be measured at least once per 7 effective full 2.

Fq power days until 2 successive maps-indicate that maximum (Z

is not increasing.

over z K(z) f.

With the relationships specified in 4.2.2.2.c above not being satisfied:

1.

Calculate the percent F (z) exceeds its limit by the following q

expression:

r 3

N Fg (z) x W(2) maximum

-1 100 for P > 0.5 p:RTP

/

1 over z

~

x 4

)

F "(z) x W(z maximum g

-1 x 100

-for P < 0.5 over z RTP p

x K(z)

I

0. 5 -

J-e 2.

Either of the follosing actions shall 'be taken; Place the core in an equilibrium condition where the a.

limit'in 4.2.2.2.c is satisfied.

Power level may. then be increased provided the AFD limits of Specification.3 l'

~

are reduced 1% AFD for each percent F (7) exceeded its limit, or 0

'Comhly_with-therequirementsofSpecification3.2.2for-b.

F (z) exceeding.its limit by the percent calculated c.bove.

q J

SEQUOYAfi - UNIT 1 -

3/4 2-7 Amendment No. 19,'95, 140, 155

POWER DISTRIBUTION LIMITS i

This page intentionally deleted.

SEQUOYAH - UNIT 1 3/4 2-9 Amendment No. 12.-140, 155 1

- - - - =

POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FArTOR LIMITING CONDITION FOR OPERATION 3.2.3 The Nuclear Enthalpy Hot Channel Factor, F shall be limited by the following relationship:

H F

i H

[1. 0 + UaH (1.0- W H

THERMAL POWER where P = RATED THERMAL POWER '

RTP F

= The F limit at RATED THERMAL POWER (RTP) specified in the H

COLR, and N

PF3g = The power factor multiplier for F specified fit the COLR.

H E

APPLICABILITY:

MODE 1 ACTION:

With F exceeding its limit:

3g Reduce THERMAL POWER to less than 50% of RATED 1HERMAL POWER within a.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to 5 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

(

b.

Demonstrate thru in-core mapping that F is within its limit g

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and Identify and correct the cause of the out of limit condition prior c.

to increasing THERMAL POWER above the reduced limit required by a.

or b. above; subsequent POWER OPERATION nay proceed provided that F H is demonstrated through in-core mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior-to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior _to exceeding this THERMAL POWER and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95% or greater RATED THERMAL POWER.

SEQUOYAH - UNIT 1 3/4 2-10 Amendment No. 19, 138, iS5

REACTIVITY CONTROL SYSTEMS BASES mndition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with ter.oerature at RATED THERMAL POWER conditions.

This value of the MDC was then transformed into the limiting end of cycle life (E0L) MTC value.

The 300 ppm surveillance limit MTC value represents a conservative value (with corrections ter burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting E0L l

MTC value.

The su.veillance requirements for measurement of the MTC at the beginning and near the end of each fuel cycle are ac'tquate to confirm that the MTC remains within its limits since this coefficient changes slowlv aue principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATIIRE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Re' actor Coolant System average temperature less than 541 F.

This limitation is required to ensure 1) the moderator temperature coefficient is withir. its analyzed temperature range, 2) the protective instrumentation is within its normai operating range, 3) the P-12 interlock is above its setpoint,

4) the pressurizer is capable of being in a OPERABLE status with a steam bubble, and 5) the reactor pressure vessel is above its minimum RT temperature.

NDT 3/4.1.2 BORATION SYSTEMS

)

The boron injection system ensures that negative reactivity control is available during each mode of facility operation.

The components required to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow pathr. 4) boric acid transfer pumps, 5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above 200 F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoperable.

The boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.6% delta k/k after xenon decay and cooldown to 200 F.

The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires SEQUOYAH - UNIT 1 B 3/4 1-2 Amendment 155

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:

(a) caintaining the calculated DNBR in the core at or above design during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within atsumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not exceeded.

The definitions of certain hot channel e seaking factors as used in these specifications are as follows:

F (:)

Heat Flux Hot Channel Factor, is defined as the maximum local n

heat flux on the surface of c fuel rod at coro elevation z divided by the average fuel rod heat flux, allowing for menufacturing tolerances on fuel pellets and rods.

N F

Nuclear Ent!tipy Rise Hot Channel Factor is defined as the ratio of the g

integral of linecr power along the rod with the -highest integrated power to the average rod power.

3/4.2.1 AXIAL FLUX DIFFERENCE (AFC)

The limits on AXIAL FLUX DIFFERENCE assure that the F (z) upper bound q

envelope of the F limit specified in the COLR times the normalized axis 1 peak-l q

irig factor is not exceeJed during either ncemal operation or in the event of xenon redistribution following power changes.

Provisions for monitoring the AFD on an automatic basis are deriveri from the plant process computer through the AFD Monitor Alarm-The computer deter-mines the one.ninute average of each of the OPERABLE excore detector-outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the allowed AI-Power operating space and the THERMAL POWER is greater than 50 percent of RATED THERMAL POWER._

3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ANTHALPY HOT ~ CHANNEL FACTORS The limits on the heat. flux hot channel factor and the nuclear enthalpy rise hot channel factor ensure that 1) the:Jesign limits on peak local power. density and minimum DNBR are not exceeded =and.2) in the event-of a LOCA the peak fuel clad temperature will not exceed-the 2200 F ECCS acceptance criteria limit.

SEQUOYAH - UNIT I

.B 3/4 2 Amendment Non 19,1138,140, :155 A

_,-ersn. #srme_ g te gs -m,wv.ra-ts' yww-w.er w-resp.',gywe-w w w.w a e* + 9 e + s wwry'w sy,

' amp wart-e

.N.ee-e,"g.

l POWER DISTRIBUTION LIMITS j

BASES I

Each of these hot channel factors is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to insure that the limits are maintained provided:

a.

Control rods in a single group move together with no individual rod insertion differing by more than + 13 steps from the group demand position.

b.

Control rod groups are sequenced with overlappit,g groups as described in Specification 3.1.3.6.

The control rod insertion limits of Specifications 3.1.3.5 and c.

3.1.3.6 are maintained.

d.

The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

The F limit as a function of THERMAL POWER allows changes in the radial H

power shape for all permissible rod insertion limits.

Ffg will be maintained within its limits provided conditions a thru d above, at maintained.

When an F tolerance mustqbe allowed for. measurement is taken, both experimental error and manufactu The 5% is the appropriate d lowance for a full core map taken with the incore detector flux mapping system and 3% is the appropriate allowance for manufacturing tolerance.

When an F is measured, experimental error must be allowed for and 4% is H

the appropriate allowance for a full core map taken with the incore rietection ThespecifiedlimitforF$H also contains an 8% allowance for system.

uncertainties which mean that normal operation will result in FN RTP

< p

/1.08.

The 8% allowance is based on the following considerations.

~

abnormal perturbations in the radial power shape, such as from rod a.

misalignment, effect F more directly than F.

H q

b.

although rod movement has a direct influence upon liaiting F to q

within its limit, such control is not readily available to limit Ffg,and errors in prediction for control power shape detected during startup c.

phycics test can be compensated for in F by restricting axial flux q

distribution.

This compensation for F is less readily.available.

H SEQUOYAH'- UNIT 1 B 3/4 2-2 Amendment No. 19, 138, 155

P_0WER DISTRIBUTION LIMITS BASES Fuel rod bowing reduces the value of DNB ratio. Margin has been retained between the DNBR value used in the safety analysis (1.38) and the WRB-1 correlation limit (1.17) to completely offset the rod bow penalty.

The applicable value of rod bow penalty is referenced in the FSAR.

Margin in excess of the rod bow penalty is available for plant design flexibility.

N

)

The hot channel ' actor Fq (z) is measured periodically and increased by a cycle and height dependent power factor, W(z), to provide assurance that the limit on the hot channel factor, F (z), is met.

W(z) accounts for the effects q

of normal operation transients and was determined from expected power control maneuvers over tha full range of burnup conditions in the core.

The W(z) function is specified in the COLR.

3/4.2.4 QUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis.

Radial power distribution measurements are made during startup testing and periodically during power operation.

-The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and cor-rection of a dropped or misaligned rod.

In the event such acticn does not correct the tilt, the margin for uncertainty on F is reinstated by reducing n

the power by 3 percent from RATED THERMAL POWER fDr each percent of tilt in excess of 1.0.

3/4.2.5 DNB PARAMETERS

\\

The limits on the DNB related parameters assure that each of the para-meters are maintained within the normal steady state envelope of operation assumed in the transient and accident analysos.

The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adeouate to maintain a minimum DNBR of greater than or equal to the safety analysis DNBR limit throughout each analyzed transient.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

SEQUOYAH - UNIT 1 B 3/4 2-4 Amendment No. 19, 138, 155 l

ADMINISTRATIVE CONTROLS MONTHLY REACTOR OPERATING REPORT 6.9.1.10 Routine reports of opcating statistics and shutdown experience, including documentation of all challenges to the PORVs or Safety Valves, shall be submitted on a monthly basis no later than the 15th of each month following

{

the calendar month covered by the report.

CORE OPERATING LIMITS REPORT i

I 6.9.1.14 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

s 1.

Moderator Temperature Coefficient BOL and E0L limits and 300 ppm 4

surveillance limit for Specification 3/4.1.1.3, 2.

Shutdown Bank Insertion Limit for Specification 3/4.1.3.5, 3.

Control Bank Insertion Limits for Specification 3/4.1.3.6, 4.

Axial Flux Difference Limits for Specification 3/4.2.1, 5.

Heat Flux Hot Channel Factor, K(z), and W(z) for Specification 3/4.2.2, and 6.

Nuclear Enthalpy Hot Channel Factor and P uer Factor Multiplier for Specification 3/4.2.3.

6.9.1.14.a The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in:

1.

WCAP-9272-P-A, "WESTINCHOUSE RELOAD SAFETY EVALUATION METHODOLOGY", July 1985 (W Proprietary).

(Methodology for SpeciTications 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 -

Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Hot Channel Factor.)

4 2.

WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL 0FFSET CONTROL F q

SURVEILLANCE TECHNICAL SPECIFICATION", JUNE 1983 (W Proprietcry).

(Methodology for Specification 3.2.1 - Axial Fliix Difference (Relaxed Axial Offset Control) and 3.2.2 - Heat Flux Hot Channel Factor (W(z) surveiM ance requirements for F Methodology).)

9 3.

WCAP-10266-P-A Rev. 2, "THE 1981 REVISION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987, (W Proprietary).

(Methodology for Specification 3.2.2 - Heat FTux Hot Channel Factor).

1 SEQUOYAH - UNIT 1 6-21 Amendment Nos. 52, 58, 72, 74, 117,-155

CORE OPERATING LIMITS REPORT (continued) 6.9.1.14.b The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

6.9.1.14.c The CORE OPEMTING LIMITS REPORT shall be provided within 30 days after cycle start up (Mode 2) for each relcad cycle or within 30 days of issuance of any midcycle revision to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

SPECIAL REPORTS 6.9.2.1 Special reports shall be submitted within the time period specified for each report, in accordance with 10 CFR 50.4.

6.9.2.2 Diesel Generater Reliability Improvement Program As a minimum the Reliability Improvement Program report for NRC audit, required by LCO 3.8.1.1, Table 4.8-1, shall include:

(a) a summary of all tests (valid and invalid) that occurred within the time period over which the last J0/100 valid tests were performed (b) analysis of failures and determination of root causes of' failures (c) evaluation of each of the recommendations of NUREG/CR-0660, " Enhancement of Onsite Emergency Diesel Generator Reliability in Operating Resctors,"

with respect to their application to the Plant (d) identification of all actions taken or to be taken to 1) correct the root causes of failures defined in b) above and 2) achieve a general improvement of diesel generator reliability (e) the schedule for implementation of each action from d) above (f) an assessment of the existing reliability of electric power to engineered-safety-feature equipment SEQUOYAH - UNIT 1 6-21a Amendment Nos. 52, 58, 72, s

-74,t117, 155.

J

\\

2.1 SAFETY LIMITS BASES 2.1.1 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 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 coefficient.

DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the WRB-1 correlation and the W-3 correlation for conditions outside the range of WRB-1 correlation.

The DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.

The local DNB haat 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 DNB design basis is a:, follows:

there must be at least a 95 percent probability that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 or W-3 correlation in this application).

The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with-95 pe-cent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit.

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

The curves are based on an enthalpy hot channel factor, Ffg, specified in the Core Operating Limit Report (COLR) and a reference cosine with a peak of 1.55 for axial power shape.

An allowance is included for an increase in F

at reduced power based on the expression:

g Ffg=Ffg[1+PF g (1-P)]

P where P =

THERMAL POWER RATED THERMAL POWER FhP=theFfg limit at RATED-THERMAL POWER (RTP) specified in the COLR, and PFaH = the power factor multiplier for F H specified in the COLR.

SEQUOYAH - UNIT 1 B 2-1 Amendment No. 19, 114, 138, 155

3/4.0 APPLICABILITY BASES The specifications of this section provide the general requirements applicable to each of the Limiting Conditions for Operation and Surveillance Requirements within Section3/4.

3.0.1 This specification defines the applicability of each specification in terms of defined OPERATIONAL MODES or other specified conditions and is provided to delineate specifically when each specification is applicable.

3.0.2 This specification defines those conditions necessary to constitute compliance with the terms of an individual Limiting Condition for Operation and associated ACTION requirement.

3.0.3 This specification delineates the ACTION to be taken for circumstances not directly provided for in the ACTION statements and whose occurrence would violate the intent of the specification.

For example, Specification 3.5.2 re-quires two independent ECCS subsystems to be OPERABLE and provides explicit ACTION requirements if one ECCS subsystem is inoperable.

Under the requirements of Specification 3.0.3, if both of the required ECCS subsystems are inoperable, within one hour measures must be initiated to place the unit in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in at least HOT SHUTOOWN within the fol-lowing 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

As a further example, Specification 3.6.2.1 requires two con-tainment spray subsystems to be OPERABLE and provides explicit ACTION require-ments if one spray subsystem is inoperable.

Under-the requirements of Specifi-cation 3.0.3, if both of the required containment spray subsystems are inoper-able, within one hour measures must be. initiated to place the unit in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , in at least HOT SHUTOOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , s

3.0.4 This specification provides that entry into an OPERATIONAL MODE or other specified applicability condition must be made with (a) the full complement of required systems, equipment or components OPERABLE and (b) all other parameters as specified in the Limiting Conditions for Operation being met-witnout regard

-for allowable deviations and out of service provisions contained in the ACTION statements.

The intent of this provision is to insure that facility operation is not initiated with either required equipment or systems inoperat'le or other specified limits being exceeded.

Exceptions to this provision have been provided-for a limited number of specifications when startup with inoperable equipment would not affect plar,t

~

safety.

These exceptions are stated in the ACTION statements of the appropriate specifict.tions.

SEQUOYAH - UNIT 1-B 3/4 0-1 Amendment No.155

3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each cold leg injection accumulator ensures that a sufficient volume of borated water will be imme @.tely forced into the reactor core in the event that the RCS pressure falb below the specified pressure of the accumulators.

For the cold leg injection accumulators, this condition occurs in the event of a large or small rupture.

The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met.

The limits in the specification for accumulator volume and nitrogen cover pressure are analysis limits and do not include instrument uncertainty.

The cover pressure limits were determined by Westinghouse to be 615 psia and 697.5 psia.

Since the instrument read-outs in the control room are in psig, the TS valves have been converted to psig and rounded to the nearest whole numbers.

The actual nitrogen cover pressure safety limits in SQN's design documents are 600.3 psig and 682.8 psig.

The minimum boron concentration ensures that the reactor core will remain subcritical-during the post-LOCA (loss of coolant accident) recirculation phase based upcn the cold leg accumu-lators' contribution to the post-LOCA sump mixture concentration, The accumulator

" operating bypasses" power operated isolation valves are considered to be in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not mat.

In addition, as these accumulator isolation valves

{

fail to meet single failure criteria, removal of power to the valves is requireu.

~

The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures.

If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required.

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be availabl in the event of a LOCA assuming the loss of one subsystem through any sing e failure consideration.

Eithersubsystemoperatinginconjunctionwiththeaccumulatorsiscapableof supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging frcm the double ended break of the largest RCS cold leg pipe downward.

In addition, each ECCS subsystem provides long term core cooling capability in the recircu-lation mode during the accident recovery period.

SEQUOYAH - UNIT 1 B 3/4 5-1 Amendment No.155

. = _ -

s PLANT SYSTEMS BASES X =

Total relieving capacity of all safety valves per steam line in lbs/ hour, 4. 75 x 108 lbs/ hour at 1170 psig.

Y =

Maximum relieving capacity of any one safety valve in lbs/ hour, 950,000 lbs/ hour at 1170 psig.

3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY-of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350 F from normal operating conditions in the event of a total loss of off-site power.

The steam driven auxiliary feedwater pump is capable of delivering 880 gpm (total feedwater flow) and each of the electric driven auxiliary feedwater pumps are capable of delivering 440 gpm (total feedwater flow) to the entrance of the steam generators at steam generator pressures of 1100 psia.

At 1100 psia the open steam generator safety valve (s) are capable of relieving at least 11% of nominal steam flow.

A total feedwater flow of 440 gpm at pressures of 1100 psia is sufficient to ensure that adequate feedwater flow is available to remove de::ay heat and reduce the Reactor Coolant System temperature to less than 30CF where the Residual Heat Removal System may be placed into operatica.

The surveillance test values ensure that each pump will provide at leas'. MO gpm plus pump recirculation flow against a steam generator pressure of 1100 psia.

Each motor-driven auxiliary feedwater pump (one Train A and one Train B) supplies flow paths to two steam generators.

Each flow path contains an automatic air-operated level control valve (LCV).

The LCVs have the same train designation as the associated pump and are provided trained air.

The turbine-driven auxiliary feedwater pump supplies flow paths to all four steam generators.

Each of these flow paths contains an automatic air-operated LCV, two of which are designated as Train A, receive A-train air, and provide flow to the same steam generators that are supplied by the B-train motor-driven auxiliary feedwater pump.

The remaining two LCVs are designated as Train 8, receive B-train air, and provide flow to the same steam generators that are supplied by the A-train motor-driven pump.

This design provioes the required redundancy to ensure that at least two steam generators receive the necessary flow assuming any single failure.

It can be seen from the description provided above that the loss of a single train of air (A or B) will not prevent the auxiliary feedwater system from performing its intended safety function and is no more severe than the loss. of a single auxiliary feedwater Therefore, the loss ~ of a single train of auxiliary air only affects the pump.

capability of a single motor-driven auxiliary feedwater pump because the turbine-driven pump is still capable of providing flow to two steam generators that are separate from the other motor-driven pump.

Two redundant steam sources are required to be operable to ensure that at least one source is available for the steam-driven auxiliary feedwater (AFW) pump operation following a feedwater or main steam line break. This require-ment ensures that the plant remains within its design basis (i.e., AFW to two intact steam generators) given the event of a loss of the No. 1 steam generator

. SEQUOYAH - UNIT 1

.B 3/4 7-2 Amendment No. 115 155

4 PLANT SYSTEMS BASES because of a main steam line or feedwater line break and a single failure of the B-train motor driven AFW pump.

The two redundant sources must be aligned such that No.1 steam generator source is open and operable and the No. 4 steam generator source is closed and operable.

For instances where one train of emergency raw cooling water (ERCW) is declared inoperable in accordance with technical specifications, the AFW turbine-driven pump is con.cidered operable since it is supplied by both trains of ERCW.

This position is consistent with American National Standards Insti-tute/ANS 58.9 requirements (i.e., postulation of the failure of the opposite train is not required while relying on the TS limiting condition for operation).

3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months with steam discharge to the atmosphere concurrent with total loss of off-site power.

The contained water volume limit includes an allowance for water not useable because of tank discharge line location or other physical characteristics.

l SEQUOYAH - UNIT 1 B 3/4 7-2a Amendment No. 115, 155

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