ML051310280

Technical Specifications Change 04-06 - Relocation of Specifications in Accordance with Part 50.36 in Title 10 of the Code of Federal Regulations.

ML051310280
Person / Time
Site:	Sequoyah
Issue date:	04/27/2005
From:	Pace P Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TVA-SQN-TS-04-06
Download: ML051310280 (82)

Text

Tennessee Valley Authority, Post Office Box 2000, Soddy-Daisy, Tennessee 37384-2000 April 27, 2005 TVA-SQN-TS-04-06 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Gentlemen:

In the Matter of ) Docket Nos. 50-327 Tennessee Valley Authority 50-328 SEQUOYAH NUCLEAR PLANT (SQN) - UNITS 1 AND 2 - TECHNICAL SPECIFICATIONS (TS) CHANGE 04 "RELOCATION OF SPECIFICATIONS IN ACCORDANCE WITH PART 50.36 IN TITLE 10 OF THE CODE OF FEDERAL REGULATIONS (CFR)"

Pursuant to 10 CFR 50.90, Tennfessee Valley Authority (TVA) is submitting a request for a TS change (TS-04-06) to Licenses DPR-77 and DPR-79 for SQN Units 1 and 2. The proposed TS change will relocate the provisions for Rod Drop Time, Movable Incore Detectors, Meteorological Instrumentation, Reactor Coolant System Chemistry, Reactor Coolant System Head Vents, Steam Generator Pressure and Temperature Limitation, Sealed Source Contamination, Refueling Operations Communications, and Manipulator Crane.

These provisions are found in TS Sections 3.1.3.4, 3.3.3.2, 3.3.3.4, 3.4.7, 3.4.11, 3.7.2, 3.7.10, 3.9.5, and 3.9.6, respectively. These specifications do not meet the four criterion found in 10 CFR 50.36 and can be removed from the TSs. These specifications will be relocated to the Technical Requirements Manual without change to the current TS requirements. jaJe Pr.ted cn recKd paw

rI U.S. Nuclear Regulatory Commission Page 2 April 27, 2005 TVA has determined that there are no significant hazards considerations associated with the proposed change and that the TS change qualifies for categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and enclosures to the Tennessee State Department of Public Health.

TVA does not have specific scheduling needs for the proposed TS change and processing can be pursued at NRC's discretion.

TVA requests that the implementation of the revised TS be within 45 days of NRC approval.

There are no commitments contained in this submittal.

If you have any questions about this change, please contact me at 843-7170 or Jim Smith at 843-6672.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 27th day of April, 2005.

Sincerel P. L. Pace Manager, Site Licensing and Industry Affairs

Enclosures:

1. TVA Evaluation of the Proposed Changes

2. Proposed Technical Specifications Changes (mark-up)

3. Changes to Technical Specifications Bases Pages cc: See page 3

U.S. Nuclear Reguiatory Commission Page 3 April 27, 2005 Enclosures cc (Enclosures):

Framatome ANP, Inc.

P. 0. Box 10935 Lynchburg, Virginia 24506-0935 ATTN: Mr. Frank Masseth Mr. Lawrence E. Nanney, Director Division of Radiological Health Third Floor L&C Annex 401 Church Street Nashville, Tennessee 37243-1532 Mr. Douglas V. Pickett, Senior Project Manager U.S. Nuclear Regulatory Commission Mail Stop 07-A15 One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2739

ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY (TVA)

SEQUOYAH NUCLEAR PLANT (SQN)

UNITS 1 AND 2

1.0 DESCRIPTION

This letter is a request to amend Operating Licenses DPR-77 and DPR-79 for SQN Units 1 and 2. The proposed changes would remove the requirements for Rod Drop Time, Movable Incore Detectors, Meteorological Instrumentation, Reactor Coolant System (RCS) Chemistry, RCS Head Vents, Steam Generator Pressure and Temperature Limitation, Sealed Source Contamination, Refueling Operations Communications, and the Manipulator Crane. These specifications do not meet the four criterion found in Part 50.36 of Title 10 in the Code of Federal Regulations (10 CFR 50.36) and can be removed from the technical specifications (TSs). These specifications will be relocated to the Technical Requirements Manual (TRM) without change to the current TS requirements. The proposed change will place these requirements in an appropriate document consistent with the safety significance for these functions. As part of the TRM, changes to these specifications will be performed in accordance with 10 CFR 50.59.

2.0 PROPOSED CHANGE

The proposed change will relocate several current TSs to the TRM. These relocations are consistent with the criterion found in 10 CFR 50.36 for functions that must be included in the TSs. These specifications do not meet the four criterion of 10 CFR 50.36 and are proposed to be relocated to the TRM.

The requirements affected are Specifications 3.1.3.4 (Rod Drop Time), 3.3.2 (Movable Incore Detectors), 3.3.3.4 (Meteorological Instrumentation), 3.4.7 (RCS Chemistry),

3.4.11 (RCS Head Vents), 3.7.2 (Steam Generator Pressure and Temperature Limitation), 3.7.10 (Sealed Source Contamination), 3.9.5 (Refueling Operations Communications),

and 3.9.6 (Manipulator Crane). The proposed change is consistent with the content of the latest version of the standard TSs for Westinghouse Electric Company units (NUREG-1431, Revision 3).

These specifications will be incorporated into the TRM without change to the current TS requirements. Future El-i

changes to these specifications in the TRM would be performed in accordance with 10 CFR 50.59 and applicable site procedures.

In summary, TVA proposes to relocate several TS requirements to the TRM for those features that do not meet the criteria for inclusion in the TSs. These relocations will retain current requirements in the TRM. Modifications to these provisions in the TRM will be in accordance with 10 CFR 50.59.

3.0 BACKGROUND

The specifications proposed for relocation provide a variety of functions. Below is a brief description of the function provided by each specification.

Rod Drop Time This specification ensures that the safety analysis assumption on control rod insertion time is preserved.

The intended safety function of the control rods (reactor trip) is to place the reactor in a subcritical condition when a trip setpoint is exceeded.

The negative reactivity insertion following a reactor trip is a function of the position of the rod control cluster assemblies, and the variation in rod worth as a function of rod position. The critical parameter as used in the safety analysis is the time of insertion up to the dashpot entry.

Movable Incore Detectors This specification ensures the operability of the movable incore detector instrumentation when required to monitor the flux distribution within the core. The movable incore detector system is used for periodic surveillance of the power distribution and calibration of the excore detectors. This surveillance verifies that the peaking factors are within their design envelope. The system is not used continuously and does not initiate any automatic protection action.

Meteorological Instrumentation The meteorological instrumentation is used to record meteorological data for use in evaluating the effect of an E1-2

accidental radioactive release from the plant. 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. The meteorological instrumentation is not used to mitigate a design basis accident (DBA) or transient.

Reactor Coolant System Chemistry This specification places limits on the oxygen, chloride, and fluoride content in the RCS to minimize corrosion.

The limitations on the RCS chemistry, provided by this requirement, ensure that corrosion of the RCS is minimized and reduces the potential for RCS leakage or failure due to corrosion. Maintaining the chemistry within the steady-state limits provides adequate corrosion protection to ensure the structural integrity of the RCS over the life of the plant.

The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

Reactor Coolant System Head Vents The RCS head vents are provided to exhaust non-condensable gases and/or steam from the RCS which could inhibit natural circulation core cooling following any event involving a loss of offsite power and requiring long-term cooling, such as a loss of coolant accident (LOCA). Their function, capabilities, and testing requirements are consistent with the requirements of Item II.B.1 of NRC's Nuclear Regulation (NUREG) 0737, "Clarification of TMI Action Plan Requirements," however, the operation of the RCS vents is not assumed in the safety analysis. This is because the operation of the vents is not part of the primary success path. The operation of these vents is an operator action after the event has occurred and is only required when there is indication that natural circulation is not occurring.

Steam Generator Pressure and Temperature Limitation This specification places limits on the steam generator pressure and temperature to ensure that the pressure induced stresses are within the maximum allowable fracture E1-3

toughness stress limits. The pressure and temperature limits are based on a steam generator reference temperature sufficient to prevent brittle fracture.

Sealed Source Contamination This specification ensures that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values. The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR Part 70.39(a)(3) limits for plutonium.

Refueling Operations Communications This specification requires communication between the control room and the refueling station to ensure that any abnormal change in the facility status observed on the control room instrumentation can be communicated to the refueling station personnel.

Manipulator Crane This specification ensures that the lifting device on the manipulator crane has adequate capacity to lift the weight of a fuel assembly and a rod control cluster assembly, and that an automatic load limiting device is available to prevent damage to the fuel assembly during fuel movement.

This specification also ensures that the auxiliary hoist on the manipulator crane has adequate capacity for latching and unlatching control rod drive shafts.

4.0 TECHNICAL ANALYSIS

The proposed change will relocate the identified specifications to the TRM consistent with the 10 CFR 50.36 requirements. In each case, the four criteria of 10 CFR 50.36 did not apply to these functions. These criteria are as follows:

1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

2. A process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

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3. A structure, system, or component that is a part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The following is a discussion for each proposed relocation and the basis for not satisfying each of these criterions.

These discussions are consistent with evaluations performed during the development of the improved standard TSs for Westinghouse plants as documented in Westinghouse Commercial Atomic Power (WCAP) 11618, "Methodically Engineered, Restructured, and Improved Technical Specifications, MERITS Program - Phase II Task 5, Criteria Application," dated November 1987.

Rod Drop Time Criterion 1: The rod drop time is not installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2: The rod drop time is a variable that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. However, the rod drop time is not a process variable that is displayed on a permanently installed readout device in the control room, nor can it be controlled by the operator.

Criterion 3: The rod drop time is not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4: Rod drop times are not explicitly modeled in probabilistic risk assessments. Several probabilistic risk assessment modeling assumptions may, however, be affected if the post-trip integral core power is higher.

Potential impacted areas are manual shutdown following an E1-5

anticipated transient without scram (ATWS), auxiliary feedwater capacity and consequential LOCAs. If reactor power were delayed, additional heat must be removed via the main and auxiliary feedwater systems or via bleed and feed techniques.

For severe accidents, the Westinghouse pressuried water reactor is relatively insensitive to the potential increase in rod drop time due to large initial steam generator masses associated with the Westinghouse design and the fact that the main feedwater is available for the vast majority of transient initiators. ATWS studies show that reactivity insertion via manual reactor trip at two minutes or shutdown, via boration after 10 minutes (much longer than any potential deviations in rod drop time),

would prevent a core meltdown event and minimize the potential for public risk. The rod drop time is not of prime importance to dominate risk sequences for shutdown following ATWS and post-trip heat removal.

Probabilistic risk assessments also model the impact of consequential LOCAs and steam generator valve failures.

The risks from consequential LOCAs and steam generator valve failures have typically been low relative to the spurious LOCA (open drains, seals, and leaks) sequences.

However, a review of the basis for pressurizer and steam generator valve challenge rates and potential for failure indicates that small trip delays would not significantly increase the potential for steam generator valve challenges or subsequent potential failure. Spurious LOCAs would still dominate core damage risk.

It is concluded the rod drop time is not an important constraint in a dominate risk sequence.

Movable Incore Detectors Criterion 1: The movable incore detectors are not installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2: The movable incore detectors are not a process variable that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

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Criterion 3: The movable incore detectors are not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4: Probabilistic risk assessment studies are not sensitive to the peaking factors and reactor coolant flow present at the beginning of severe accidents. This is because in probabilistic risk assessments, success is defined as preventing significant fuel melting following a severe accident by maintaining adequate gross core cooling using available safety systems. Adequacy of core cooling is determined through deterministic systems modeling, which includes assumptions on flow rates, peaking factors, and so forth, during the accident. Information provided by the movable incore detectors would be of little or no use in mitigating the consequences of a severe accident and such information is not modeled in probabilistic risk assessments. Local hot spots, at which departure from nucleate boiling may occur or which may experience other localized adverse effects and initial RCS flow are not significant to risk.

This specification does not contain constraints of prime importance in limiting the likelihood or severity of the accident sequences that are commonly found to dominate risk.

Meteorological Instrumentation Criterion 1: The meteorological instrumentation is not installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2: The meteorological instrumentation is not a process variable that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3: The meteorological instrumentation is not assumed to function in the safety analysis. The meteorological instrumentation is not a structure, system, E1-7

or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4: Offsite dose calculations in probabilistic risk assessment studies for large accidental releases of radioactive materials rely on conservative meteorological and evacuation assumptions and do not take credit for the meteorological instruments cited in this TS to guide emergency measures to protect the public. Routine releases of radioactive materials are not risk significant.

This TS is not of prime importance in risk dominant sequences.

Reactor Coolant System Chemistry Criterion 1: The RCS chemistry is not installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2: The RCS chemistry is not a process variable that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3: The RCS chemistry is not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4: Primary system corrosion is a slow process which could be detected by inservice inspections or small leakages before it caused a rupture. Undetected corrosion would not be expected to have a significant affect on LOCA frequencies.

Primary system chemistry is not evaluated in probabilistic risk assessment studies and is not considered to be a significant contributor to risk.

Reactor Coolant System Head Vents Criterion 1: The RCS head vents are not installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary.

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Criterion 2: The RCS head vents are not a process variable that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3: The RCS head vents may be used to assist in creating conditions conducive to natural circulation, but are not components that are part of the primary success path and which function to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4: The design of Westinghouse pressurized water reactors is such that a buildup of non-condensable gases or steam within the primary system, which is sufficient to inhibit natural circulation core cooling, is unlikely.

Inadvertent opening of an RCS head vent would result in a loss of primary coolant. Small break LOCA initiating event frequencies, as used in probabilistic risk assessments, are developed from a large base of pressurized water reactor operating experience. These initiating event frequencies include failures of all types of components that would result in a small LOCA. The RCS head vent contribution to the overall initiating event frequency is very small and is not a primary contributor to risk.

This specification does not contain constraints of prime importance to dominant risk sequence.

Steam Generator Pressure and Temperature Limitation Criterion 1: The steam generator pressure and temperature limits are not installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2: The steam generator pressure and temperature limits are not a process variable that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3: The steam generator pressure and temperature limits are not a structure, system, or component that is part of the primary success path and which functions or E1-9

actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4: This limit is intended to prevent the brittle fracture of the steam generator. Brittle fractures of steam generators are not specifically modeled in probabilistic risk assessments. During operation in Modes 1, 2, and 3, which are the operating modes typically evaluated in probabilistic risk assessments, violation of these pressure and temperature limits would be unlikely.

Violation of the pressure and temperature limits while in other modes could increase the likelihood of subsequent brittle failure at higher pressures. This could lead to a somewhat higher probability of steam generator tube rupture or steam generator secondary side failure (equivalent to an inside containment secondary high energy line break). Steam generator tube ruptures and inside secondary high energy line (steam or feedwater line) breaks are not dominant contributors to risk in current probabilistic risk assessment studies.

The requirements of this TS are not of prime importance in limiting the consequences of risk dominant sequences.

Sealed Source Contamination Criterion 1: The sealed source contamination is not installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2: The sealed source contamination is not a process variable that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3: The sealed source contamination is not a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4: The requirements addressed in this TS controls removable low level contamination on sealed sources. This is done to limit the leakage to allowable El-10

levels. This requirement is not addressed in probabilistic risk assessments nor important to probabilistic risk assessment conclusions. Sealed sources are used for calibration and other purposes which have no impact on sequences that will degrade the reactor core.

Refueling Operations Communications Criterion 1: This refueling communication specification does not contain requirements for installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2: This refueling communication specification is not a process variable that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3: This refueling communication specification does not contain requirements for a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4: This specification helps assure direct communications between control room and refueling station personnel during refueling, which would help to preclude inadvertent criticality.

This event is not addressed in probabilistic risk assessment studies and would not be important in accident sequences that are commonly found to dominate risk.

Manipulator Crane Criterion 1: This manipulator crane specification does not contain requirements for installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2: This manipulator crane specification does not contain requirements for a process variable that is an initial condition of a DBA or transient analysis that El-1l

either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3: This manipulator crane specification does not contain requirements for a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4: The strength and load limits of the manipulator cranes and the crane travel could affect the chance and severity of fuel degradation resulting from damaging a fuel assembly during fuel handling.

These are design basis type accidents that have not been significant to risk when analyzed in environmental reports.

The above evaluations for the applicability of 10 CFR 50.36 were performed in support of the effort to develop the latest version of the improved TSs. Westinghouse proposed that these specifications, along with several others, not be included in the standard TS based on the applicability to 10 CFR 50.36. NRC agreed with this assessment and issued NUREG-1431 for Westinghouse plants without the inclusion of the specifications proposed for relocation. The latest version of NUREG-1431 (Revision 3) retains this position by not including these specifications. TVA has evaluated the above basis for 10 CFR 50.36 applicability and has concluded that they are applicable to the SQN design and operation.

Therefore, the proposed relocation to the TRM is acceptable.

The relocation of these specifications to the TRM will place these requirements in a controlled document that requires an evaluation for changes in accordance with 10 CFR 50.59. TVA controls the TRM through administrative procedures and the revision process includes onsite review and approval requirements equivalent to the processing of a change to the TSs. These provisions ensure that the requirements found in the TRM will be properly controlled to ensure that safety functions are not adversely affected. These specifications are being .added to the TRM without changing the current requirements; therefore, the level of safety currently found in the TSs will continue to apply. Changes to these requirements will be processed appropriately as described above. For these reasons and the evaluations for 10 CFR 50.36 applicability, the proposed relocation of these specifications to the TRM is acceptable.

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5.0 REGULATORY SAFETY ANALYSIS The proposed change is a request to amend Operating Licenses DPR-77 and DPR-79 for SQN Units 1 and 2. The proposed changes would remove the requirements for Rod Drop Time, Movable Incore Detectors, Meteorological Instrumentation, Reactor Coolant System (RCS) Chemistry, RCS Head Vents, Steam Generator Pressure and Temperature Limitation, Sealed Source Contamination, Refueling Operations Communications, and the Manipulator Crane. These specifications do not meet the four criterion found in Part 50.36 of Title 10 in the Code of Federal Regulations (10 CFR 50.36) and can be removed from the technical specifications (TSs). These specifications will be relocated to the Technical Requirements Manual (TRM) without change to the current TS requirements. The proposed change will place these requirements in an appropriate document consistent with the safety significance for these functions. As part of the TRM, changes to these specifications will be performed in accordance with 10 CFR 50.59.

5.1 No Significant Hazards Consideration TVA has evaluated whether or not a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change only relocates requirements to TRM that are not required to be included in the TSs in accordance with 10 CFR 50.36. Changes to the TRM require evaluations and reviews in accordance with 10 CFR 50.59 to ensure that the health and safety of the public is not adversely affected. The proposed relocation retains the current TS requirements and only alters the location of these provisions. This relocation cannot affect the probability or consequences of an accident as this is only an administrative revision that will not alter any plant equipment or processes. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

Since the proposed change only relocates the current TS requirements without change, there is not a potential for a change in the accident generation potential. This change will not alter plant components, systems, or operating practices.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change relocates specifications that do not meet the threshold for inclusion in the TSs as defined in 10 CFR 50.36. This change will not alter the requirements for these functions or plant setpoints or functions that maintain the margins of safety. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, TVA concludes that the proposed amendment(s) present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The Commission's regulatory requirements related to the content of the TS are contained in Title 10, Code of Federal Regulations (10 CFR), Section 50.36. The TS requirements in 10 CFR 50.36 include the following categories: (1) safety limits, limiting safety systems settings and control settings, (2) limiting conditions for operation (LCO), (3) surveillance requirements (SRs), (4) design features, and (5) administrative controls. The specifications proposed for relocation El-14

are included in the TSs but do not meet the applicability requirements in 10 CFR 50.36, "Technical Specifications." NRC noticed in the Federal Register on July 22, 1993, "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," that included the position that specifications that do not meet the 10 CFR 50.36 applicability criteria may be proposed for relocation to licensee-controlled documents such as the Final Safety Analysis Report (FSAR). TVA has relocated several other specifications since this time to the TRM, which includes 10 CFR 50.59 review requirements equivalent to the FSAR. The TRM meets the NRC expectations for a licensee-controlled document for the purpose of relocated specifications that do not meet 10 CFR 50.36 criteria as evident by previous NRC approvals to relocate TSs. Therefore, the proposed relocation of these specifications to the TRM is acceptable.

NUREG-1431, Revision 3, "Standard Technical Specifications Westinghouse Plants," provides generic recommendations for requirements associated with the operation of Westinghouse Electric Company designed nuclear power plants. NUREG-1431 does not contain requirements for the specifications proposed for relocation. Therefore, the removal of these specifications from the TSs and incorporation into the TRM is consistent with the NUREG.

Based on the current requirements for the specifications proposed for relocation being retained without change, other regulatory requirements and criteria will continue to be satisfied. The application of these requirements as part of the TRM will be identical to those applicable to TSs. The only difference is that changes can be processed without prior NRC review and approval provided 10 CFR 50.59 criteria is satisfied. Therefore, applicable regulatory requirements and criteria will be satisfied as a result of the proposed relocation.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

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6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 50.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

7.0 REFERENCES

1. NUREG-1431, Revision 3, "Standard Technical Specifications Westinghouse Plants," June 2004

2. Federal Register, dated July 22, 1993, "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors"

3. Title 10 of the Code of Federal Regulations, Part 50.36, "Technical Specifications"

4. Westinghouse Commercial Atomic Power (WCAP) 11618, "Methodically Engineered, Restructured, and Improved Technical Specifications, MERITS Program - Phase II Task 5, Criteria Application," dated November 1987 El-16

ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQN)

UNITS 1 AND 2 Proposed Technical Specification Changes (mark-up)

I. AFFECTED PAGE LIST Unit 1 Unit 2 Index Page IV Index Page IV Index Page V Index Page V Index Page VI Index Page VI Index Page VIII Index Page VIII Index Page Ix Index Page Ix Index Page X Index Page X Index Page XIII Index Page XIII Index Page XIV Index Page XIV 3/4 1-19 3/4 1-19 3/4 3-43 3/4 3-44 3/4 3-47 3/4 3-48 3/4 3-48 3/4 3-49 3/4 3-49 3/4 3-50 3/4 4-16 3/4 4-21 3/4 4-17 3/4 4-22 3/4 4-18 3/4 4-23 3/4 4-28 3/4 4-33 3/4 7-11 3/4 7-11 3/4 7-29 3/4 7-41 3/4 7-30 3/4 7-42 3/4 9-5 3/4 9-6 3/4 9-6 3/4 9-7 II. MARKED PAGES See attached.

E2-1

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION 314.0 APPLICABILITY ............................................ 3/4 0-1 314.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - T,,g greater than 200OF ..................................... 314 1-1 Shutdown Margin - Tavg less than or equal to 2000 F .................................... 3/4 1-3 Moderator Temperature Coefficient .................................... 3/4 1-4 Minimum Temperature For Criticality .................................... 3/4 1-6 3/4.1.2 BORATION SYSTEMS Flow Paths - Shutdown (Deleted) .................................... 3/4 1-7 Flow Paths - Operating (Deleted) .................................... 3/4 1-7 Charging Pump - Shutdown (Deleted) .................................... 3/4 1-7 Charging Pumps - Operating (Deleted) .................................... 3/4 1-7 Borated Water Sources - Shutdown (Deleted) .................................... 3/4 1-7 Borated Water Sources - Operating (Deleted) .................................... 3/4 1-7 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height .................................... 3/4 1-14 Position Indication Systems - Operating .................................... 314 1-17 Position Indication System - Shutdown (Deleted) .................................... 3/4 1-18 S utdown Rod Insertion Limit 3/ 4 1-20 Control Rod Insertion Limits .3/4 1-21 December 18, 2000 SEQUOYAH - UNIT 1 IV Amendment No. 264 E2-2

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) ....................................... 3/4 2-1 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR-FQ (Z)................................................................. 3/4 2-5 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR ....................................... 3/4 2-10 3/4.2.4 QUADRANT POWER TILT RATIO ....................................... 3/4 2-12 3/4.2.5 DNB PARAMETERS ....................................... 3/4 2-15 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION .... 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION .... 3/4 3-14 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION .. . . 3/4 3-39 OVABLE INCORE DETECTORS '(Deleted) ................................ . 314 3-43 SEISMIC INSTRUMENTATION (DELETED) .................................. . 3/4 3-44 MTOOOIA NTUN(Deleted) ........................... 314 347 REMOTE SHUTDOWN INSTRUMENTATION .. . . 3/4 3-50 CHLORINE DETECTION SYSTEMS (DELETED) .. . . 3/4 3-54 ACCIDENT MONITORING INSTRUMENTATION .... 3/4 3-55 FIRE DETECTION INSTRUMENTATION (DELETED) .... 3/4 3-58 DELETED . 314 3-70 EXPLOSIVE GAS MONITORING INSTRUMENTATION .314 3-71 September 7, 1999 SEQUOYAH - UNIT 1 V Amendment No. 62, 138, 148, 227, 245 E2-3

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup And Power Operation ............................................... 3/4 4-1 Hot Standby ............................................... 3/4 4-1a Shutdown ............................................... 3/4 4-2 3/4.4.2 SAFETY VALVES - SHUTDOWN (DELETED) ............................................... 3/4 4-3 3/4.4.3 SAFETY AND RELIEF VALVES - OPERATING Safety Valves - Operating ............................................... 3/4 4-4 Relief Valves - Operating ............................................... 3/4 4-4a 3/4.4.4 PRESSURIZER ............................................... 3/4 4-5 3/4.4.5 STEAM GENERATORS ............................................... 3/4 4-6 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Instrumentation ............................................... 3/4 4-13 Operational Leakage ............................................... 3/4 4-14 Reactor Coolant System Pressure Isolation Valve Leakage. ......................... 3/4 4-15 314.4.7 CHEMISTRY (Deleted) ...................................................................... 3/4 4-16 . I 3/4.4.8 SPECIFIC ACTIVITY ......................................................................... 3/4 4-19 ..

3/4.4.9 RCS PRESSURE AND TEMPERATURE (P/T) LIMITS Reactor Coolant System ./........................................... 3/4 4-23 Pressurizer (Deleted) ........................................... 3/4 4-26 3/4.4.10 DELETED ........................................... 3/4 4-27 3/4.4.11 REACTOR COOLANT SYSTEM HEAD VENTS (Deleted)........................................... 3/4 4-28 3/4.4.12 LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEM ... . 3/4 4-29 March 9, 2005 SEQUOYAH - UNIT 1 VI Amendment No. 116,133,157, 208, 259, 294, 297, 299 E2-4

This Page affected by TS Change 03-14 l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.6.3 CONTAINMENT ISOLATION VALVES .......................................................... 3/4 6-17 3/4.6.4 COMBUSTIBLE GAS CONTROL HYDROGEN ANALYZERS (Deleted) .. 3/4 6-24 ELECTRIC HYDROGEN RECOMBINERS - W (Deleted) . .... 3/4 6-25 HYDROGEN MITIGATION SYSTEM ..... 3/4 6-25a 3/4.6.5 ICE CONDENSER ICE BED .............................................. . 3/4 6-26 ICE BED TEMPERATURE MONITORING SYSTEM (Deleted) . ........................3/4 6-28 ICE CONDENSER DOORS ............................................... 3/4 6-29 INLET DOOR POSITION MONITORING SYSTEM (Deleted) . .........................3/4 6-31 DIVIDER BARRIER PERSONNEL ACCESS DOORS AND EQUIPMENT HATCHES ............................................... 3/4 6-32 CONTAINMENT AIR RETURN FANS .............................................. 3/4 6-33 FLOOR DRAINS .............................................. 3/4 6-34 REFUELING CANAL DRAINS ............................................... 3/4 6-35 DIVIDER BARRIER SEAL ............................................... 3/4 6-36 3/4.6.6 VACUUM RELIEF LINES .............................................. 3/4 6-38 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES .......................................................... 3/4 7-1 AUXILIARY FEEDWATER SYSTEM .......................................................... 3/4 7-5 CONDENSATE STORAGE TANK .......................................................... 3/4 7-7 ACTIVITY .......................................................... 3/4 7-8 MAIN STEAM LINE ISOLATION VALVES .......................................................... 3/4 7-10 MAIN FEEDWATER If0lA1101slREGI A1NG, AND RYS VAL VES ............... 3/4 7-10a

.72SEA EERATOR PRESSURE/TEMPERATURE LIMITATION (Deleted)...............................

3/4.7.3 CMPONENT COOLING WVA I LK bYb: I LM ............................................................. 3/4 7-12 3/4.7.4 ESSENTIAL RAW COOLING WATER SYSTEM .......................................................... 3/4 7-13 September 20, 2004 SEQUOYAH - UNIT I Vill Amendment No. 116,197, 213, 232, 277, 296 E2-5

This Page affected by TS Change 00-06 l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7.5 ULTIMATE HEAT SINK ............................................................. 3/4 7-14 3/4.7.6 FLOOD PROTECTION (DELETED) ............................................................. 3/4 7-15 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM .......................................... 3/4 7-17 3/4.7.8 AUXILIARY BUILDING GAS TREATMENT SYSTEM .................................................... 3/4 7-19 314.7.9 SNUBB 'S (DELETED)... I.....................3/4 7-21 3/..0SAE ORCE CONTAMINATION (Deleted) .......................................................... 3472 3/4.7.11 FIRE SUPPRESSION SYSTEMS (DELETED) ............................................. 3/4 7-31 3/4.7.12 FIRE BARRIER PENETRATIONS (DELETED) ............................................. 3/4 7-41 3/4.7.13 SPENT FUEL POOL MINIMUM BORON CONCENTRATION ....................................... 3/4 7-42 3/4.7.14 CASK PIT POOL MINIMUM BORON CONCENTRATION ............................................. 3/4 7-43 3/4.7.15 CONTROL ROOM AIR-CONDITIONING SYSTEM (CRACS) ........................................ 3/4 7-44 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES OPERATING ... 3/4 8-1 SHUTDOWN ... 3/4 8-8 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. DISTRIBUTION - OPERATING .... 3/4 8-9 A.C. DISTRIBUTION - SHUTDOWN .... 3/4 8-10 D.C. DISTRIBUTION - OPERATING .... 3/4 8-11 D.C. DISTRIBUTION - SHUTDOWN .... 3/4 8-14 3/4.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES (DELETED) .... 3/4 8-15 February 27, 2002 SEQUOYAH - UNIT 1 IX Amendment No. 61, 227, 235, 247, 250, 265, 273 E2-6

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION (DELETED) 3/4 8-17 ISOLATION DEVICES (DELETED) ........................................ 3/4 8-20 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION ........................................ 3/4 9-1 3/4.9.2 INSTRUMENTATION ........................................ 3/4 9-2 3/4 9.3 DECAY TIME ........................................ 3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS ........................................ 3/4 9-4

/495 COMMUNICATIONS ( Deltd ...................................... ...... 3/ 9-5 3/.9.6 MANIPULATOR CRANE (Deleted) ........................................ 3/4 9-6 3/4.9.7 CRANE TRAVEL - SPENT FUEL PIT AREA (DELETED) .3/4 9-7 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION ALL WATER LEVELS .............. 3/4 9-8 LOW WATER LEVEL .............. 3/4 9-8a 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM .3/4 9-9 3/4.9.10 WATER LEVEL - REACTOR VESSEL .3/4 9-10 3/4.9.11 SPENT FUEL PIT WATER LEVEL .3/4 9-11 3/4.9.12 AUXILIARY BUILDING GAS TREATMENT SYSTEM .3/4 9-12 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN .34 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS .3/4 10-2 3/4.10.3 PHYSICS TESTS .3/4 10-3 3/4.10.4 REACTOR COOLANT LOOPS .34 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN (DELETED) .3/4 10-5 December 18, 2000 SEQUOYAH - UNIT 1 X Amendment No. 61, 204, 213,250, 264 E2-7

INDEX BASES SECTION PAGE 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE............................................................... B 3/4 4-4b (314.4.8 SPECIFIC ACTIVITY ....................................................................... B 314 4-5 314.4.9 RCS PRESSURE AND TEMPERATURE (PIT) LIMITS . ...........................................

B3/4 4-6 5314.4.10 STRUCTURAL INTEGRITY (DELETED) ....................................................... B 3/4 4-13 ...

9~~~~~~~~~E T 3/4.4.11 ........................................B 3/4 OLN YTMHA ET Dltd ..

ECO4-13 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS................................................... B 3/4 5-1 3/4.5.2 and 34.5.3 ECCS SUBSYSTEMS.1.................................................... B 3/4 5-1 3/4.5.4 BORON INJECTION SYSTEM (DELETED) ...... B 3/4 5-2 3/4.5.5 REFUELING WATER STORAGE TANK ...... B 3/4 5-3 3/4.5.6 SEAL INJECTION FLOW.................................................. B 3/4 5-4 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT................................................... B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS ...... B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES ........ B 3/4 6-3 3/4.6.4 COMBUSTIBLE GAS CONTROL................................................... B 3/4 64 3/4.6.5 ICE CONDENSER.................................................... B 3/4 64 3/4.6.6 VACUUM RELIEF LINES................................................... B 314 6-6 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE ..... B 3/4 7-1 13472SEMGNRATOR PBESSURE/TEMPERATURE LIMITATION (Deleted) ........... 7-3/4.7.3 COMPONENT COOLING WATER SYSTEM ..... B 3/4 7-3a November 9, 2004 SEQUOYAH - UNIT 1 XIII Amendment No. 157,197, 259,294, 297 E2-8

This Page affected by TS Change 00-06 INDEX BASES SECTION PAGE 3/4.7.4 ESSENTIAL RAW COOLING WATER SYSTEM ............................................. B 3/4 7-3a 3/4.7.5 ULTIMATE HEAT SINK (UHS) ............................................. B 3/4 7-4 3/4.7.6 FLOOD PROTECTION ............................................... B 3/47-4 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM ......................................... B 3/4 7-4 3/4.7.8 AUXILIARY BUILDING GAS TREATMENT SYSTEM ............................................. B 3/4 7-5 3/4.7.9 SNUBBERS 'DELETED) ...................................................................... B 3/4 7-5 3/4.7.11 FIRE SUPPRESSION SYSTEMS (DELETED) ............................................. B 3/4 7-7 3/4.7.12 FIRE BARRIER PENETRATIONS (DELETED) .. ............................................. B 3/4 7-8 3/4.7.13 SPENT FUEL POOL MINIMUM BORON CONCENTRATION ...................................... B 3/4 7-9 3/4.7.14 CASK PIT POOL MINIMUM BORON CONCENTRATION .......................................... B 314 7-13 3/4.7.15 CONTROL ROOM AIR-CONDITIONING SYSTEM (CRACS) ..................................... B 3/4 7-16 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 and 3/4.8.2 A.C. SOURCES AND ONSITE POWER DISTRIBUTION SYSTEMS ............. B 3/4 8-1 3/4.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES (DELETED) ............. B 3/4 8-2 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION .............................................. B 3/4 9-1 3/4.9.2 INSTRUMENTATION .... B 3/4 9-1 3/4.9.3 DECAYTIME . PNT B 3/4 9-1 3/4.9.5 COMMUNICATIONS (Deleted) ............................................................................ B / -

314.9.6 MANIPULATOR CRANE (Deleted) ......................................................................... l / -

3/4.9.7 CRANE TRAVEL - SPENT FUEL P EA (DELETED) ............................................B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION .................................... B 3/4 9-2 3/4.9.9 CONTAINMENT VENTILATION SYSTEM ............... ............................... B 3/4 9-3 February 27, 2002 SEQUOYAH - UNIT 1 XIV Amendment No. 157, 204, 227, 235, 265, 273 E2-9

REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 1 This specification has been deleted.

3.1.3.4 The Individual full longth (ehutdown and control) rod drop time from the fully withdrawn position#-

shall bo lo6 than or oqual to 2.7 sconds from beginning of docay of Mtationary grippor coil voltago to dachpot ontry with:

-T - e Ate r than or o q ula to 5411 0 °C, an d g

b. All reacrtor coolant pumnpe oprat;ng.

APPLICABILITY: MODES I and 2 ACT40N1

a. With tho drop time of any full length rod dotorminod to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding toRJ MODE A oF 2.

b. With the rod drop time6 wAthin limits but determined with 3 reactor coolant pumpe opeating, operation may procoed provided THERMAL POWER ic reetricted to Ie£E than or equal-to 71% of RATED THERMAL POWER SURVEILLANCE REQUIREMENTS 1.1.3.1 The rod drop time of full length rod6 shall be demonstrated through moacuroment prior to reactor critic~ality:

a. ierall rod6 folloYng each removal Of th reartor vecigo head,

b. For 6pocifically affected individual rods folloIng any maintenance on or modification to the rcontrnl rold dr.ive qstem wvhic^h couldl affect the drop time of thece rpecific rode, and

r. At lWact orne pre 1Q mnRthe.

\ 1tFully withdrawn ehal! be the condition where hutdown and control bankE are at a pocition within the itntoR'al Ž2:22 and

• 231 etepe WithdraWn, inclucI6Ve.

May 08, 1990 SEQUOYAH - UNIT 1 3/4 1-19 . Amendment No. 108, 138 E2-10

INSTRUMENTATION MOVABLE INCORE DETECTORS LIMITING CONDITION FOR OPERATION I This specification has been deleted.

3.3.3.2 Tho movablo incoro detection system shall be OPERABLE with:

a. At loast 7594 of tho dotector thimblos,

b. A minimum of 2 dotoctor thimbles por coro quadrant, and

c. Sufficiont movablo detoctors, drive, and readout oquipmont to map thoso thimblos.

APPLICABILITY: Whon tho movablo incoro dotection systom is used for:

a. Rocalibration of tho excoro neutron flux dotoction system,

b. Monitoring the QUADRANT POWER TILT RATIO, or

c. Moasuroment of FHand RFQ ACTION: -

/ Withtho movable incoro detection cystom inoporablo, do not use tho system for the above applicable monitoring or calibration functions. Tho provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.2 The movable incoro detection system shall bo demonstrated OPERABLE by normaliZing oach dotoctor output when required for:

a. Rocalibration of the oxcoro neutron flux dotoction cystom, or

b. Monitoring the QUADRANT POWER TILT RATIO, or

c. of n 7Z 'oasuoent (Z)A/

April 11, 2005 SEQUOYAH - UNIT 1 3/4 3-43 Amendment No. 19, 301 E2-11

MENTATION METEOROLOGICAL INSTRUMENTATION LIMITING CONDITION FOR OPERATION I This specification has been deleted.

3.3.3.4 The meteorological monitoring inetrumFntation- channol shown in Table 3.3 8 chall be GRERABLE.-

APPLICABILITY:- At all timos.

&ACTION:-

a. With one or moro requirod meteoroloilcal monitoring channlt inoporablo for moroe than 7 days, preparo and cubmit a Special Report to the CommiGciFn pursuant to Specification 6.9.2 Aithin the next 10 day utlining t cahe the-malfunction and the planc-for reStORng the channel(c) to OPERABLE statuc.

lb. The proviFionc of Specification 3.0.3 are not applicable.l SURVEILLANCE REQUIREMENTS 41.3.3.1 Each of the above meteorological monitoFing iRtrumentation channelsc hall be demonetrated OPERABLE by the peoformance of the CHANNEL CHECK and CHANNEL CALIBRATION- operatienr at April 11, 2005 SEQUOYAH - UNIT 1 3/4 3-47 Amendment No. 301 E2-12

,J ~METEOROLOGICAL MONITORING INSTRUMENTATION >

I This table has been deleted.

INSTRUMENT LOCAT4ON MINIMUM-OPE=RABLEP

1. WIND SPEED

. Chane;4l I Nominai EIovR780 MSL 4 b-Chane- 2-. Nominal Elov. 807 MSL 4

2. WIND DIRECTION

a. Ghaie41, Nominal Elov.780 MSL 4 b Ghanne2-2 7 Nominal Elov. 780MStL 4M 3.AIRT TMPERA.TURE DEL-TA T

/a. Channol 2, Nominal Elov. 780 to 1017 MSL 1 SEQUOYAH - UNIT 1 3/4 3-48 E2-13

TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

. This table has been deleted.

INSTRUMENT CHANNEL CHECK CHANNEL CALIBRATION

1. WIND SPEED

a. Nominal EIov.780 MSL SA

b. Nominal Elov. 897 MSL E SA

2. WIND DIRECTION

3. Nominal Eov. 780 MSL D SA

b. Nominal Elov. 897 MSL SA

3. AIR TEMPERATURE DELTA T

a. Nominal Elov. 780 897 MSL D SA

b. Nominal Elov. 780 1017 MSL D SA SEQUOYAH - UNIT 1 3/4 3-49 E2-14

REACTOR COOLANT SYSTEM 3/4.4.7 CHEMISTRY LIMITING CONDITION FOR OPERATION l This specification has been deleted.

3.4.7 The Reactor Coolant Syetom chGmictry shall be maintainod within tho limits specified in Tablo 3.1 APPLICIABILITY: At all timo./

AG~IOI MODES 1, 2, 3 and 1

a. With any ono or more chomistry parameter in excess of itUSteady State Limit but within its TrAnsient Limit, restore the parameter to within its Steady State Limit within 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> or be in at least HOT STAWNDY ~Wthin tho noxt 6 hoUr and in C OLD SH UT DOW N with4n the

,following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With anR nr more Ghoe;rtr' paramter in GXcoss of its Tranient Limit, be Rn at least HOT STANDBY;w;ithn 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> a;nd in COLID SHUTDOWQAN' within the folloying 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

At all othor times6 Wffith the concnRtFation Of oither- chlorideorAfuoido in the Roactor Coolant Systemn ce orf its Steady State Limit for morA than 2-1 hur r in c of Tr_

Tranient Limit, reduco the pressurizor pressure to los6 than or equal to 500 psig, if applicablo, and porform an enginein evaluation to determinR the effects of the out of limit conditonn on the structural integrity of the Reactor Coolant Syst9 m; determine~ that the Reactor Coolant System; remain! cetaR o ontinued oporationrior to incoasing the MrCeAurizer pressure abovo 500 psig or prior to proeeGdingtoMDE1 SURPVEILLANCE REQUIREMENTS 4.1.7 The Roactor Coolant System chomietry shall bo determined to be within the limits by analySie off those parameters at the frequencies specified In Tble 1.1 3.

SEQUOYAH - UNIT 1 3/4 4-16 September 17, 1980 E2-15

< ~TABLE 3.4-2 j ~REACTOR COOLANT SYSTEM CHEMISTRY LIMITS This table has been deleted.

T DPARAMETER STEADY STATE LIMIT TRANSIENT LIMIT DISSOLVED OXYGEN* --0.40 ppm 51.00 PPm CHLORIDE 0.15 ppm 1 L1.50 ppm FLUORIDE -0.5 ppf 1.50 P

• Lmit Rot appliTable ith Th 4-17 Septmbe 17 SEQUOYAH - UNIT 1 314 4-17 September 17, 1980 E2-16

~~ TABLE 4 .4-3 J ~REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS TThis table has been deleted.

PARAMETER MINIMUM ANALYSIS FREQUENCIES l lDISSOLVED OXYGEN* At leat once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CHLORIDE At loat onco per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

/IFLUORIDE At loast once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 0

Noat rogiterod with T IecGc than or equal to 2590 F SEQUOYAH - UNIT 1 3/4 4-18 E2-17

R R COOLANT S 3/4.4.11 REACTOR COOLANT SYSTEM HEAD VENTS LIMITING CONDITION FOR OPERATION I This specificationhas been deleted.

3.4.11 At loact ono Roactor Coolant Systom Hoad Vent (RCSHV) path shall bo OPERABLE.'

APPI ICABI lY: MODAESD 1, 2, anrd

)AGTIQN4

/AWith no RCSHV path OPER-BLE , ractoro at loast ono path to OPERABLE status within 30 dayr or bo in HOT STANDBY within 6 hourc and HOT SHUTDOWN within tho following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLA.NCE REQUIREMENTS 4.1.11 Each RCSHV path chall bo domonstratod OPERABLE at leact once por 18 monthc by:

a. Vorifying that tho upstroam manual icolation valvoe aro lockod in tho opon poSition,

b. Oporating oach romotoly controllod valvo through at loact ono cyclo from tho control room,

c. Verifying flow through oach RCSHV path.

\

• noporablo pathe muMt bo maintainod dloted with power removod from tho valvo actuatort. If any RCSHV path is declared inoporable whilo in an applicable MODE, power shall bo romovod from the valve actuatorRwithin ono hour.

April 11, 2005 SEQUOYAH - UNIT 1 3/4 4-28 Amendment Nos. 116, 123, 133, 213, 301 E2-18

PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSUREiTEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION

)1.This specification has been deleted.

3.7.2 Tho tomporatures of both tho primary and Socondary coolantS in tho 6toam gonoratorc shall boe \

groater than 70°F whon tho proccuro of oithor coolant in tho ctoam genorator ij groator than 200 pcig.

APPLICABILIT: At all timoe.

\ACTION:

With tho roguiromontS of tho abovo epocification not Ratisfiod:

a. Roduco tho etoam gonorator proesuro of tho applicablo 6ido to leEE than or oqual to 200 pEig within 30 minutoe, and

b. Porform an onginooring ovaluation to dotormino tho offoct of tho ovorproF6urization on tho etructural intogrity of tho stoam gonorator. Dotormino that the Stoam genorator romain9 accoptablo for continued oporation prior to incroasing its tamporaturos abovo 2000F.

SURVEILLANCE REQUIREMENTS 4.7.2 Tho proccur in oach sido of tho 6toam gonorator shall bo dotorminod to bo leOE than 200 pEig at loaet onco per hour whon the tomporaturo of eithor the pri mary or cocondary coolant ic laes than 70 0F.

March 25,1982 SEQUOYAH - UNIT 1 3/4 7-11 Amendment No. 12 E2-19

SY 3/4.7.10 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION 1 This specification has been deleted.

3.7.10 Ech rRalald rnorc contininr rFadioac09ti material either tn ~exess of 100 micrcriesr of hoeta-and/or gamma omitting material or 5 microcurioS of alpha omitting matorial rhall bo froo of greater than or equal to 0.005 micro curios of romovable contamination.

APPLICABILIT: At all times.

ACTION:

a. With araled4 sonurrA having removable contamination in ex-ess of the above limitr,/

imRmediately Wthdraw the rcAlrd urco fromn ure atnd:

rne

1. Either deconntaninate a-nd repair the eoaled sou rieo, Or

/2 Dispore of the soaled in acord-ance iource wth Rommiscwi;n Regulations.

b. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.10.1 Test Roquiremonts Each coaled ceurce shall be tested for leakage and/or contamination by:

a.The licenceR, or b.Othor porsons pecifically authorized by the Commission or an Agreement State.

The test method shall have a detection sensitivity of at least 0.005 microcuries per test sample.

1.7.10.2 Tact Each category of soaled ourFos (excluding rta-tup rurces

-roancis o and fisionn detectors preViously subjected to -oe flux) shall be _est__ed a-_tthe frequen der-_ibed WblOW

a. Sourcesr, in usea - At leart onc pe oths for all cronaledourceG coRtainingradioactive

1. With a half life greater than ;30 days (excluding Hydrogen 3), anFd

2. In any frnm other than gas.

l~~~~~~pi T XT WI 11,2005rT April 11, 2005 SEQUOYAH - UNIT 1 3/4 7-29 Amendment No. 12, 301 E2-20

AT SYSTEMS I This page Intentionallyleft blank l SURVEILLANCE REQUIREMENTS (Continued)

I~~b h SCtord siource not inir u tErch rcoaled corco ar.nd fiscion detector rhal1 bov rtese rorl ton

+

use or transfer to another tioneR uRn lers ttetedr 4.A.thin the proviousrix months Soalod sources and fission detetors transferred v..'ithout a Gortificate indicating the -cat test date shall be totted prorF to bein placed int use.

G. Startup sources; and fission detectors Eac=Rh roaled ctartup source and fission d-etector chat! be toted -ARin 31 days prior to beiRn subjected to coe fluX or inrtalled in theoGre anRd followng repair or maintenance t the co urce 4.7.0.3 REport A reprt shall be prepared and sub-mitted to the CommrRission on anann -Iual bas~i if oarled soure- -r fission detector leakage tests reveal the prenceoef greater than or equal to 0-0015n microcurie6 of removable contamination.

March 25, 1982 SEQUOYAH - UNIT 1 3/4 7-30 Amendment No. 12 E2-21

RFEING OPERATIONS=

X3/4.9.5 COMMUNICATIONS g LIMITING CONDITION FOR OPERATION l This specification has been deleted.

3.9.5 Diroct communicationc shall bo maintainod botweon tho control room and poreonnol at tho rofueoing6tation.

APPLICABILITY: During CORE ALTERATIONS.

AGT4IN:

Whon diroct communicationr bethWoon tho control room and porGonnolat tho rofuoling ctation cannot bo maintained, SuSpond all CORE ALTERAJTIONS. Tho provicion6 of Spocification 3.0.3 are not applicablo.

SURVEILLA.NCE REQUIREMENTS A

4.9.5 frr9t cormnrnunt-W~iocRS betWeRn tho control room aRd poonen!l at tho r.fulling 6tation chall bo deomontratod- ;Athin ono heur prior to tho 6tart of and at least oncR8 pr 12 houlr during CORE

_LTrarERArr

.TIN  ;

SEQUOYAH - UNIT 1 3/4 9-5 E2-22

UELING OPERATIONS F 3/4.9.6 MANIPULATOR CRANE J LIMITING CONDITION FOR OPERATION\

1 This specificationhas been deleted.

3.9.6 Tho manipulator crane ;nd auxiliary hoist shall bo used for movement of drive rode or fuel asvomblies and shall bo OPERABLE with:

a- The manipulator crano uod for movement of fuel assemblios having:

4- A minimum capacity of 2750 pounds, snd 2-. An otorloRd cuit ff limit lees than or equal to 2700 poIndS.

b, The auxiliary honit usod for latching and unlatching driVe odac having:

47 A. minimu-m apaci; Of 610 pounds, and 2-. A load indicstor which shall be us od to provont lifting loads in OXCG£E of 600 pounds.

A.PPI-CG.lA.  : During moVement of drivo rods oFfuol arsemblies Withinthe roactor persuro Vessel.

With the roquiromonte for cranoe and/or hoist OPERABILITY not satisfiod, suspend U£0 of any inoporable mn-nipulatocrn an oauxiliary hoist from operatonRS iRnolVing the momene t Of driVe rFodS and fuol arssmblior withir .hn the rctr pnrosUlr vessel. Tho provision of Specification 3.0.3 aeRnot applicablo.

SURVEILLANCE REQUIREMENTS 1.9.6.1 Each manipulator crsno u-ed for movement of fuol sesomblios yithin the reactor prossuroe vessel ghall bo demonstrated OPERABLE within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of such operations by-porforming a load teet of at loast 2750 pounds; and domonstrating an automatic load cut off Vhon the crnolaid- o-xcood6 2700 pounds.

1.9..2 ac h auxiliary hoist and Rasoiated load indcra;tor us-ed for moemernt of drive rods;wthiRn the era~rtOF nreS6UF9aASSArrhtill hhedt-rgn.

- me n 1FAR tr= pthin r hgUF pnr0 p +ih 1 rt;;rt nf rUhih peotienRG by rr40nio aia- te-r1 of at leatt 61en poulRl&-

.ot March 25, 1982 SEQUOYAH - UNIT 1 3/4 9-6 Amendment No. 12 E2-23

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY .............................................. 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - Ta,0 > 200OF .................................. 3/4 1-1 Shutdown Margin - Tayg

• 200OF .................................. 3/4 1-3 Moderator Temperature Coefficient .................................. 3/4 1-4 Minimum Temperature For Criticality .................................. 3/4 1-6 3/4.1.2 BORATION SYSTEMS Flow Path - Shutdown (Deleted) .................................. 3/4 1-7 Flow Paths - Operating (Deleted) .................................. 3/4 1-7 Charging Pump - Shutdown (Deleted) .................................. 3/4 1-7 Charging Pumps - Operating (Deleted) .................................. 3/4 1-7 Borated Water Source - Shutdown (Deleted) .......... ........................ 3/4 1-7 Borated Water Sources - Operating (Deleted) ........... ....................... 3/4 1-7 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height .................................. 3/4 1-14 Position Indication Systems-Operating .................................. 3/4 1-17 Position Indication System-Shutdown (Deleted) .................................. 3/4 1-18 Rod Drop Time (Deleted) ......................... 3/4 1-19 Shutdown Rod Insertion Limit .......................... 3/4 1-20 Control Rod Insertion Limits .......................... 3/4 1-21 December 18, 2000 SEQUOYAH - UNIT 2 IV Amendment No. 255 E2-24

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 314.2.1 AXIAL FLUX DIFFERENCE (AFD) ............................................................. 3/4 2-1 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR-FQ(Z) ............................................................. 3/4 2-4 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR ............................................................ 3/4 2-8 3/4.2.4 QUADRANT POWER TILT RATIO ............................................................. 3/4 2-10 3/4 2.5 DNB PARAMETERS ............................................................. 3/4 2-13 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION ............................................................. 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION ........ 3/4 3-14 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation ............................................................. 3/4 3-40 ovare Incore Detect ors (Deleted) ............................................................. 3/4 3-44 ASeismic Instrumentation (Deleted) .............................. 3/4 3-45 eteo Instrumentation (Deleted).............................. 3/4 3-48 Remote Shutdown Instrumentation .................................................. 3/4 3-51 Chlorine Detection Systems (Deleted) ................................................. 3/4 3-55 Accident Monitoring Instrumentation .................................................. 3/4 3-56 Fire Detection Instrumentation (Deleted) ............................................... 3/4 3-59 Deleted .............................. 3/4 3-68 Explosive Gas Monitoring Instrumentation .............................. 3/4 3-69 September 7, 1999 SEQUOYAH - UNIT 2 V Amendment No. 54, 130, 134, 218, 236 E2-25

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION ........................................................ 3/4 4-1 Hot Standby ........................................................ 3/4 4-2 Hot Shutdown ........................................................ 3/4 4-3 Cold Shutdown ........................................................ 3/4 4-5 3/4.4.2 SAFETY VALVES -SHUTDOWN (Deleted) ........................................................ 3/4 4-6 3/4.4.3 SAFETY AND RELIEF VALVES - OPERATING Safety Valves Operating ........................................................ 3/4 4-7 Relief Valves Operating ........................................................ 3/4 4-8 3/4.4.4 PRESSURIZER ........................................................ 314 4-9 3/4.4.5 STEAM GENERATORS ........................................................ 3/44-10 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Instrumentation ........................................................ 3/4 4-17 Operational Leakage ........................................................ 3/4 4-18 Reactor Coolant System Pressure Isolation Valve Leakage ................................... 3/4 4-19 3 HEMISTRY (Deleted) ........................................................... 34-1...

/ C .7 Ax3/4.4.8 SPECIFIC ACTIVITY .................................................................. 3/4 4-24 l3/4.4.9 RCS PRESSURE AND TEMPERATURE (P/T) LIMITS Reactor Coolant System ............................ 3/4 4-28 Pressurizer (Deleted) ............................ 3/4 4-31 3/4.4.10 DELETED ...................................................... 3/4 4-32 3/4.4.11 REACTOR COOLANT SYSTEM HEAD VENTS (Deleted) ........................................... 3/4 4-33

.4.12 LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEM ................. 3/4 4-34 March 9, 2005 SEQUOYAH - UNIT 2 VI Amendment No. 106, 120, 138, 147, 198, 250, 284, 288 E2-26

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.6.3 CONTAINMENT ISOLATION VALVES ................................ 3/4 6-17 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Monitors (Deleted) .. 3/4 6-24 Electric Hydrogen Recombiners - W (Deleted) . ... 3/4 6-25 Hydrogen Mitigation System .. 3/4 6-26 3/4.6.5 ICE CONDENSER Ice Bed .... 3/4 6-27 Ice Bed Temperature Monitoring System (Deleted) .... 3/4 6-29 Ice Condenser Doors .... 3/4 6-30 Inlet Door Position Monitoring System (Deleted) .... 3/4 6-32 Divider Barrier Personnel Access Doors And Equipment Hatches .... 3/4 6-33 Containment Air Return Fans .... 3/4 6-34 Floor Drains .... 3/4 6-35 Refueling Canal Drains .... 3/4 6-36 Divider Barrier Seal .. . . 3/4 6-37 3/4.6.6 VACUUM RELIEF VALVES .... 3/4 6-39 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves .. 3/4 7-1 Auxiliary Feedwater System .... 3/4 7-5 Condensate Storage Tank .... 3/4 7-7 Activity .. 3/4 7-8 Main Steam Line Isolation Valves .... 3/4 7-10 ManFed sgion, Reg ulating, And-yass Valves. ............ 3/4 7-1lOa (3472 SEMGNRATOR PRESSURE/TEMPERATURE LIMITATION (Deleted) 3/4. -11....

3/4.7.3 COMPONENT COOLING WATER SYSTEM .3/4 7-12 September 20, 2004 SEQUOYAH - UNIT 2 Vill Amendment No. 188, 203, 222, 268, 286 E2-27

This Page affected by TS Change 00-06 l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 314.7.4 ESSENTIAL RAW COOLING WATER SYSTEM ............................................. 3/4 7-13 3/4.7.5 ULTIMATE HEAT SINK ............................................. 3/4 7-14 3/4.7.6 FLOOD PROTECTION PLAN (DELETED) ............. ................................ 3/4 7-15 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM ........................................... 3/4 7-17 3/4.7.8 AUXILIARY BUILDING GAS TREATMENT SYSTEM ............................................. 3/4 7-19 3/4.7.9 SNUBBERS (DELETED) ............................................. 3/4 7-21 3471SELDSUCE CONTAMINATION (Deleted ..................................

314.7.11 FIRE SUPPRESSION SYSTEMS (DELETED) ....................................... 314 7-43 314.7.12 FIRE BARRIER PENETRATIONS (DELETED) .............................................. 314 7-52 3/4.7.13 SPENT FUEL POOL MINIMUM BORON CONCENTRATION ....................................... 3/4 7-53 3/4.7.14 CASK PIT POOL MINIMUM BORON CONCENTRATION ............................................. 3/4 7-54 3/4.7.15 CONTROL ROOM AIR-CONDITIONING SYSTEM (CRACS) ............................................. 3/4 7-55 314.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES Operating ............................................. 314 8-1 Shutdown ............................................. 3/4 8-9 314.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. Distribution - Operating ............................................. 3/4 8-10 A.C. Distribution - Shutdown .............................................. 3/4 8-11 D.C. Distribution - Operating ............................................. 3/4 8-12 D.C. Distribution - Shutdown ............................................. 314 8-15 February 27, 2002 SEQUOYAH - UNIT 2 IX Amendment No. 218, 225, 238, 256, 262 E2-28

I INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 314.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment Penetration Conductor Overcurrent Protective Devices (DELETED) ................................................ 3/4 8-16 Motor Operated Valves Thermal Overload Protection (DELETED) ............ ........... 314 8-18 Isolation Devices (DELETED) ................................................ 3/4 8-21 3/4.9 REFUELING OPERATIONS 314.9.1 BORON CONCENTRATION ............................................................. 314 9-1 314.9.2 INSTRUMENTATION ............................................................. 314 9-3 3/4.9.3 DECAY TIME ............................................................. 3/4 9-4 314.9.4 CONTAINMENT BUILDING PENETRATIONS.3/ ....................................... . 314 9-5

,314.9.5 CMMUNICATIONS (Deleted) ............................................ 314 9-6 .

3/.9.6 MANIPULATOR CRANE (Deleted) ........................................... 3149-7 .. ,

314.9.7 CRANE TRAVEL - SPENT FUEL PIT AREA (DELETED). 3/4 9-8 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION All Water Levels. 3/4 9-9 Low Water Level ........................................................... 3/4 9-10 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM ..................................................... 3/4 9-11 314.9.10 WATER LEVEL - REACTOR VESSEL ......................... .................................. 314 9-12 3/4.9.11 WATER LEVEL-SPENT FUEL PIT ........................................................... 3/4 9-13 3/4.9.12 AUXILIARY BUILDING GAS TREATMENT SYSTEM ..................................................... 3/4 9-14 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN ........................................................... 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS .......................... 314 10-2 3/4.10.3 PHYSICS TESTS ........................................................... 3/4 10-3 February 22, 2000 SEQUOYAH - UNIT 2 X Amendment No. 53,194, 203, 241 E2-29

INDEX BASES I I SECTION PAGE 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE .. .................................................... B 314 4-4 3/ 4.4.8 SPECIFIC ACTIVITY ............................................................................ B 3 / 4-5 314.4.9 RCS PRESSURE AND TEMPERATURE (P/T) LIMITS ................................................... B 3/4 4-7 3/4.4.10 STRUCTURAL INTEGRITY (DELETED) ................................................... B 3/4 4-14 3/.45.11 RENCYO COOLANT SYSTEM HEAD VENTS (Deleted) ........................................ B 3S4 4-14 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS ..................................... B 3/4 5-1 314.5.2 and 3/4.5.3 ECCS SUBSYSTEMS ..................................... B 3/4 5-1 3/4.5.4 BORON INJECTION SYSTEM ..................................... B 3/4 5-2 3/4.5.5 REFUELING WATER STORAGE TANK ..................................... B 3/4 5-2 3/4.5.6 SEAL INJECTION FLOW ....................................... B 3/4 5-4 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT ....................................... B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS ..................................... B 314 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES ..................................... B 3/4 6-3 3/4.6.4 COMBUSTIBLE GAS CONTROL ..................................... B 3/4 6-4 3/4.6.5 ICE CONDENSER ..................................... B 3/4 6-4 3/4.6.6 VACUUM RELIEF VALVES ..................................... B 3/4 6-6 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE ................................................................... B 3/4 7-1

{ 3/4.72 STEAMGENERAOR PRESURE/TEPERATURE LIMITATION B 3/4 7-3A (Deleted) . .......................................

3/4.7.3 COMPONENT COOLING WATER SYSTEM ............................................................... B 3/4 7-3a September 15, 2004 SEQUOYAH - UNIT 2 XIII Amendment No. 147, 188, 250, 284 E2-30

I This Page affected by TS Change 00-06 l INDEX BASES SECTION PAGE 3/4.7.4 ESSENTIAL RAW COOLING WATER SYSTEM .............................................. B 3/4 7-3a 3/4.7.5 ULTIMATE HEAT SINK .............................................. B 3/4 7-4 3/4.7.6 FLOOD PROTECTION .............................................. B 3/4 7-4 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM .......................................... B 3/4 7-4 3/4.7.8 AUXILIARY BUILDING GAS TREATMENT SYSTEM .............................................. B 3/4 7-5 3/4.7.9 SNUBBERS..................................B 314 7-5 3/4.7.11 FIRE SUPPRESSION SYSTEMS (DELETED)........................B 3/4 7-7 3/4.7.12 FIRE BARRIER PENETRATIONS (DELETED) ............................... B 3/4 7-8 3/4.7.13 SPENT FUEL POOL MINIMUM BORON CONCENTRATION ............... B 3/4 7-9 3/4.7.14 CASK PIT POOL MINIMUM BORON CONCENTRATION ...... B 3/4 7-13 314.7.15 CONTROL ROOM AIR-CONDITIONING SYSTEM (CRACS).............................................. B 3/4 7-16 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 and 3/4.8.2 A.C. SOURCES AND ONSITE POWER DISTRIBUTION SYSTEMS ..... B 3/4 8-1 3/4.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES (DELETED) ....... B 3/4 8-2 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION .............................................. B 3/4 9-1 3/4.9.2 INSTRUMENTATION ................................................ B 3/4 9-1 3/4.9.3 DECAY TIME................................................................................................................. B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS .............................................. B 3/4 9-1 3/4.9.5 COMMUNICATIONS (Deleted) ................................................ B 3/4 9-2 3/4..6MA NIULATR CANE ................................................. B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL PIT AREA (DELETED) .............................................. B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION ..................................... B 3/4 9-2 3/4.9.9 CONTAINMENT VENTILATION SYSTEM .............................................. B 3/4 9-3 February 27, 2002 SEQUOYAH - UNIT 2 XIV Amendment No. 194, 218, 225, 256, 262 E2-31

/ RECTIITY CONTROL SYSTEMS _

ROD DROP TIME LIMITING CONDITION FOR OPERATION 1 This specification has been deleted.

3.1.3.4 Tho individual full longth (Shutdown and control) rod drop timo from tho fully withdrav.n poition-6hall bo 10££ than or oGqual to 2.7 soconds from boginning of docay of stationary grippor coil voltago to darshpot entr,' with:

a. Tvgrojator than or erual to 541 10 F, and

b. All roactor coolant pumps oporating.

APPLICABILITY: Modoc 1 and 2.

a. With tho drop timo of any full longth rod dotorminod to oxcood tho abovo limit, rostoro tho rod drop timo to within tho abovo limit prior to procooding to MODE I or 2.

b. With tho rod drop timoc within limitc but dotorminod with 3 roactor coolant pump6 oporating, oporation may procood providod THERMAL POWER is roctrictod to IO£E than or equal to 71% of RA.TED THERMAL POWER.

SURVEILLNCE REQUIREMENTS 4.1.3.4 The rod drop timo of full longth rods shall bo domonstratod through meaeuromont prior to roactor

a. For all rods following oach romoval of tho reactor voccol head,

b. ForF 6pocifically affoc-tod- individui-al rondM follown n mintonan-.coR on o-r moedification to tho; contrmfol roddrivoytom whih ould -ffoct tho drop timor of thoGo SpoeifiG rodr,, aRd

c. At loact onco por 18 month .6 It Fully withdra"'n shall ho the condition whoro chutdoen and control bankc aro at a position within tho intorval of 22^2 and -231 6top6 withdrawn, inclu6ivo.

October 4, 1995 SEQUOYAH - UNIT 2 3/4 1-19 Amendment Nos. 20, 98,130, 203 E2-32

INSTRUMENTATION MOVABLE INCORE DETECTORS LIMITING CONDITION FOR OPERATION I This specificationhas been deleted.

3.3.3.2 Tho movablo incoro dotoction 6ytom shall bo OPER'ABLE with:

3. lnat 7.5 __of the dtector thimbles,

b. A minimum of 2 dotrctor thimblo F GororeequadraRnt, ard r-c. Sffirantmovl-h detoctort, dr;iV, -nd roeadeut oeuipmont to rap thore thimblocra.

.A.PPLIBLI-TY:-I-- - thm-n ovabo-vn-cr-edetecio cyrctem ic- urd for:

RaRocairioo-;n of thto oxrpr noutron flux dotoction 6rctwfr,/

b.MonitoIrin tho QUADRA.NT POWER TIL-T RA.TIO, Or

c. Moacuromont of r \

ACTION:-

With tho movablo incoro dotoction cyctom inoporablo, do not uEO tho ysterm for tho abovo applicablo monitoring or calibration functionR. Tho provicionc of Spocification 3.0.3 aro not applicablo.

SURVEILLANCE REQUIREMENTS 1.3.3.2 Tho movablo incoro dotoction rystom shall bo dormonstratod OPERRABLE by normalizing oach dotoctor output Whon roquirod for:

a. Rocalibration of tho oxcoro noutron flux dotoction Systom, or

b. Monitoring tho QUADRANT POWER TILT RATIO, or

c. Moa4uromoe9t o1-fz Q/ /-_a-F April 11, 2005 SEQUOYAH - UNIT 2 3/4 344 Amendment No. 21, 290 E2-33

INSTRUMENTATION METEOROLOGICAL INSTRUMENTATION LIMITING CONDITION FOR OPERATION 4 This specificationhas been deleted.

3.3.3.4 The metoroleGcl!al moni;oritng nctrumentatioRn Ghannolc rchown in Tablo 3.3 8 shall-beI APPLICABILITY: At all timos.

ACTION-

a. With ono or morM requirod meteorological monitoring channolR inoperablo for moro than 7 dayc, prepare and submit a Special Report to the Cor mision pursuant to Specification 6.9.2 the nxt 10 days outliningtof 1vthint th malfunction and tho planS for rortoring the channol(s) to OPERABLE status.

b. The provirionr of Specification 3.0.3 are not applicable.l I)SURVEILLA.NCE REQUIREMENTS l.3.3. Each of the above meteorological monitoring instmUmontation channelc shall be demonetratod

\OPERABLE by the porformanco of the CHANNEL CHECK and CHANNEL CALIBRATION eporatipnFat theroquenciec 6h1n0 in Table 1.3 5.

April 11, 2005 SEQUOYAH - UNIT 2 3/4 3-48 Amendment No. 290 E2-34

SEQUOYAH - UNIT 2 3/4 3-49 E2-35

X ~TABLE 4.3-5\

_Jz METEOROLOGICAL MONITORING INSTRUMENTATION\

SURVEILLANCE REQUIREMENTS

. This table has been deleted.

INSTRUMENT GHANNEIL- GCANNEL G/HECKK CALIBRATION

\

1. WIND SPEED

a. Nominal Elov. 780 MSL SA

b. Nominal Elov. 897 MSL Q SA

2. WIND DIRECTION\

a. Nominal Elov. 780 MSL 4 SA

-b. Nominal Eeov. 897 MSL 4 SA

23. AIR TEMPERATURE DELTA T

a. Nominal Row. 780 807 MSL 4 SA hb. Nominal Elov. 780 1017 MSL 4 SA SEQUOYAH - UNIT 2 3/4 3-50 E2-36

3447CHEMISTRY LIMITING CONDITION FOR OPERATION I This specification has been deleted.

3.4.7 The Reactor Coolant SyStem chomistr; shall bo maintained vithin the limits specified in Table 3.1 APPLICABILITY: At all times.

MODES 1, 2, 3 and 1:

a. With any ono or moro chemistry paramotor in oxcosn of it, Steady Stato Limit but h thin its TranRient LImit, rostore the parameter to vthin its Steady State Limit vithin 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> or be in at loaot HOT STANDBRY ydthin tha neGxt 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in CL S TDOWN Wthin the-folloyAng 30 hoUrM

. With any oRno Or morO rhemiRtn, parametterin oxrcrr of itr Tranrient Limit, ba ir at loart HOT STANDBY WRthin r hours andIn COLD HUtTDOWNh'tin . .. tho ARn 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

At All Other Time:w WIth tho concor ntr-tion of ritho>r chhloneri or florideo in tho Reactor Coolant SysteRm iRnexGrces of its Stead Stte LImit for more than 24 hnour orCnio Xcsr Of 1t6 Transiont I mit, rdurcR thR presu6riZer pF66Freure to lorr thn;n or equal to 500 psig, if applicable, and perform an ongInoorinr evalwann to determine the effects of theRout of limit nondition on tho strwutral i nterfity of tho I Roactor Coolant SySemFR determin that the Reactor Coolant System remnaiRn aceptable for I continu6ed operaton prior t increasing the prresurizer pressurE aboVe 500 psig Or prior to pIroceeding t MODE 4.

SSURVEILLA.NCE REQUIRIEMENTS I .A.77The ReaorCoolant Syrtem Ghemistr.'shall be determined to be ithin the limits by analysisr of those parameters at the frequencies specified in Table 1.1 3.

SEQUOYAH - UNIT 2 3/4 4-21 E2-37

SEQUOYAH - UNIT 2 3/4 4-22 E2-38

( ~TABLE 4.4-3X t ~REACTOR COOLANT SYSTEM\

/ CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTSl 1 This table has been deleted.

SAMPLE AND PtARAMETrER ANALYSIS FREQUENCY DISSOLVED OXYGEN* At loart onco per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> CHLORIDE At least onco per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> R

FLUORIDE At lea<t once por 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

Ntruird thT.. lsoc than or equal to 2500 F SEQUOYAH - UNIT 2 3/4 4-23 E2-39

RECOCOOLANT

/~~ SYSTEM 33/.4.11 REACTOR COOLANT SYSTE HADVNS_ ,

LIMITING CONDITION FOR OPERATION I This specificationhas been deleted.

3.4.11 At least one Roactor Coolant SyStem Hoad Vent (RCSHV) path shall be OPERABLE.*a APPLICABILITY: MODES 1, 2, and 3.

ACTIGON:

With no RCSH'.V path OPERABLE*, restore at least one path to OPERAEBLE statuswiithin 30 days or be in HOT STANDBY Ylithin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and HOT SHUTDOWN Within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.11 Each RCSHV path shall be domonstrated OPERA.BLE at loast onco por 18 months by:

a. Vorifying that tho upstream manual icolation valvec aro locked in the opon position,

b. Oporating each remotely controlled valve through at loast ono cycle from tho control room, and G. erifyiRg flow through each RCSHkt path.

• !noporable paths must bo maintainod closed with poWer removed from the valvo actuators. If any RCSHV path ij declarod inoperablo while in an applicable MODE, power shall bo removed from tho valvoA actuatoem wthin oRA hour. J April 11, 2005 SEQUOYAH - UNIT 2 3/4 4-33 Amendment No. 106, 112,120, 138, 203, 290 E2-40

PLANT SYSTEMS

)314.7.2 STEAM GENERATOR PRESSURE/EMPERTRELMTIO LIMITING CONDITION FOR OPERATION I This specificationhas been deleted.

3.7.2 Tho tomporaturoe of both tho primary and socondary coolantc in tho stoam gonorators shall bo groator than 70 0F Wvhon tho procSuro of oithor coolant in tho stoam gonorator iEgroator than 200 pcig.

APPLICABILITY: M all timeos.

AGN49 With tho roquiromonte of tho abovo Spocification not catirfiod:

aS Roduco tho 6toam gonorator procsuro of tho applicablo cido to lIeoc than or oqual to 200 peig WIthin 30 minutoc, and bv Porform an onginooring ovaluation to dotormino tho offoct of tho ovorprescurization on tho structural integrity of tho 6team gonorator. Determine that the stoam generator remainc acceptable for continued oporation prior to incroaGing ite temperatures abovo 00°F SURVEILANCE REQUIREMENTS 4-7.2 The proecuro in oach cido of the stoam gonerator chall bo dotorminod to bo loss than 200 p6i9 at loaet Roncpor hour when the teompratur of either the pnrimina or socongldary nolant ic loAr than 700 Fz.

SEQUOYAH - UNIT 2 3/4 7-11 E2-41

/PANTSYTEMS 13/ .710 SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION

/ This specification has been deleted.

3.7.10 Each soalod courcA containing radioactive matorial either in eXCeO£ of 100 microcuriAc of beta and!or gamma omitting material or 5 microcuries of alpha omitting material, chall be free of greater than or equal to 0.005 microcuries of removablo contamination.

APPLICABILIY: At all timc.

ACTION:-

a1 With a soalod courco having r contamination in OXCeEc of the abevo Iimits,

-movablo immodiatoly wthdraw the coaled source from Uce and either:

4- Decontamine andr repair the sealedreurce, or 2_. Dicrpne of the rto1ad cornei n arorndanGe Y.ith Crmnmicion RogulatinR6.

b- The proviri9nR of Specifications 3.0.3 are not applicable.

SUR'.'EILLANCE REQUIREMENTS 4.7.10.1 Test Requiremente Each soaled gource chall be tected for leakage and!or contamination by:

aC The liceneRe, or b, Other persons Specifically authorized by the Commission or an Agreement State.

The test method shall have a detection sensitivity of at least 0.005 microcurieo per test rample.

1.7.10.2 Test Frequencies Each category of soaled 6eurc£e (excluding startup rourcoc and ficion /

detectors previou6ly Gubjected to cor flux) chall be tested at the frdquency described below.

\n Surncre in usen_ At learst nRepr eix months foralIl coaled surnescrontaininrg radioactive 4, With a half-life greater than 39 days (excluding HYdrogen 3), and 2-. In any form other thangas.

April 11, 2005 SEQUOYAH - UNIT 2 3/4 7-41 Amendment No. 290 E2-42

{ PANTSYSTEMS

) 1 This page Intentionally left blank SURVEILL\ANCE REQUIREMENTS (Continued) b- Storod courcoe not in ueo Each soaled sourca and fircion dotactor shall be tacted prior to UEO or tranefor to anothor Iiconaeo un!oss toteowithi Rthe prvewouc £ix monthe. Soaled sourcos and fir,1on dotectore tranctarrod without-a n-RORicato indicating tho last toet date shall be toetod prior to boingl placad into UEO

/ . Startup sourcoc and fi49mon detoctors -Each soalad ctartup eource and ficiGon dotoctor ehall bo toetod within 31 days prior to boing EUbjActad tO coro flux or inetallod in tho coro and following ropair or maintananco to tho sourco.

1.7.10.3 Roports A roport shall bo proparod and submitted to the Commicion on an annual basic if sealod sourca or fic6ion dotoctor leakago tsets roveal tho praeonca of groator than or aqual to 0.005 microcurioe of ramovabla contamination.

SEQUOYAH - UNIT 2 3/4 7-42 E2-43

/ 3/4.9.5 COMMUNICATIONS LIMITING CONDITION FOR OPERATION

) This specificationhas been deleted.

3.9.5 Diroct communications shall bo maintained betWeon tho control room and porSonnRl st tho eftation.

fuo1elin APPLICABILITY: During CORE ALTERATIONS.

ACTION.

When direct comrmunicationS betWeen the control room and prorsnnel at the roFue4linig rtation c.annot be m~ainta!nod, susrpend all CORE ALTER-4.TIO-NSR. The previision Of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.5 Diroct communications betweoon the control room and personnel at the refueling station shall bo demonstrated ono hour prior to the start of and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE Rvithin ALTERATIONS.

SEQUOYAH - UNIT 2 3/4 9-6 E2-44

LIMEITNG OPERATIONS

)314.9.6 MANIPULATOR CRANE l LIMITING CONDITION FOR OPERATION

/ This specification has been deleted.

3.9.6 Tho manipulator crano and auxiliary hoist shall bo usod for movomont of drivo rods or fuol assemblies and shall be OPERABLE with:

a-. Tho manipulator crano ueod for movomont of fuol assemblies having:

4? A minimum capacity of 2750 pounds, and 2- An overload cut off limit loAs than or equal to 2700 pounds.

b- The auxiliary hoist used for latching and unlatching drive rod having:\

4? A minimum capacity of 610 pounds, and 2, A load indicator which shall be used to prevent lifting loads in eXCes£ of 600 pounds.

APPLICABILITY: During movement of drive reds or fuel assemblies, within the reactor prs6sure vesel.

AGCTON: J Ath the requir emernts for cranae andl/er heirst OPERA.II nRot satisfied, uspend use of any inoperable maenipulaetor cranand/er auxiliary hWist frFom oprations involViRg the movement of drive rods and fu1el assemnbli;es Wthin the reactor peressure sel. The previsions Of SpeifiatioanR3.0.3 are ont applicable.

SURVEILLANCE REQUIREMENTS 4.9.6.1 Each manipulator crane used for movement of fuel assemblioe Within the reactor pressure vessel shall be demonstrated OPERABLE AWthin 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the rtant of such operations by performing a load test of at least 2750 pounds and demonstrating an automatic electrical load cut off when the crane lead exceeds 2700 pounds.

4.9.6.2 Each auxiliary hoist and associated load indicator ured for movement of drive rods AWthin the reactor pressure Vessel shall be demonstrated OPER ABLERWthin 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of such operations by porforming a load teot of at least 610 pounds.

December23, 1982 SEQUOYAH - UNIT 2 314 9-7 Amendment No. 9 E2-45

ENCLOSURE 3 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT (SQNj UNITS 1 AND 2 Changes to Technical Specifications Bases Pages I. AFFECTED PAGE LIST Unit 1 Unit 2 B 3/4 1-4a B 3/4 1-4a B 3/4 3-2a B 3/4 3-2a B 3/4 3-3 B 3/4 3-3 B 3/4 4-4r B 3/4 4-5 B 3/4 4-5 B 3/4 4-13 B 3/4 4-14 B 3/4 7-3 B 3/4 7-3 B 3/4 7-7 B 3/4 7-6a B 3/4 9-2 B 3/4 9-2 II. MARKED PAGES See attached.

E3-1

REACTIVITY CONTROL SYSTEMS BASES The maximum rod drop time restriction Is consistent with the assumed rod drop time used in the accident analyses. Measurement with Tag greater than or equal to 541 0 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 LC'O' are .Md 3/4.1.3.4 ROD DROP TIME (Deleted) 3/1.1 .3.1 ROD DROP TIME and 3/4.1.3.5 SHUTDOWN ROD INSERTION LIMIT Fully withdrawn for shutdown and control rod banks is defined as a condition where the rod banks are positioned in a range of 222 to 231 steps fully withdrawn. This range is defined to permit axial repositioning of rod banks to mitigate rod wear on internal guide surfaces.

November 21, 1995 SEQUOYAH - UNIT 1 B 3/4 1-4a Amendment No. 108, 215 E3-2

INSTRUMENTATION BASES 314.3.3 MONITORING INSTRUMENTATION 314.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

Relative to the control room instrumentation Isolation function, one set of process radiation monitors acts to automatically initiate control room isolation. The actuation instrumentation consists of redundant radiation monitors. A high radiation signal from the detector will initiate its associated train of the Control Room Emergency Ventilation System (CREVS). The CREVS is also automatically actuated by a safety injection (SI) signal from either unit. The SI function is discussed in LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS) Instrumentation." In addition, the control room operator can manually initiate CREVS.

3 MOVABLE INCORE DETECTORS 31.3.3.2 This specificationis deleted.

The OPERABILITY4 of the movable incore d-etA-cto-rrs Wi.th th9 specified FRinimum comRplement of-8Rrure6 that the mneasurtements, notained fromR UGG Of thi6 6y~t9FA accurately FGprosont the Gqipet spatial noutron flux distribution of the reactor corG.Tho OPER'\BILITY of this 6ystom is demORstFatod by irradiating each detector used and dotormining the acceptability of its voltage curve.

For the purpsos of measuring Fo(X ,,Z)eoF F,-(,Y) a full incore flux map is used. Quartor coro flux maps, as defined in WCAP 8618, Juno 1976, may be used in recalibration of the oxcoro neutron flux detection system, and full incoro flux maps or symmetric incoro thimbles may be used for monitoring the 4 AI hRG9P~FRneQape =-6 sieeqe May 31, 2000 SEQUOYAH - UNIT 1 B 3/4 3-2a Amendment Nos. 54, 190, 223, 238, 256 E3-3

INSTRUMENTATION BASES 3/4.3.3.3 SEISMIC INSTRUMENTATION This specification issdeleted..

g /43..4METEOROLOGICAL INSTRUMENTATION\

Tis specificationIs deleted.

The OPERABILITY of the mneteorological instrumentation ens-reg that suffciet me~t9orlgic9al -

data icavailable for eotimating potential radiation doser to the public as a result of routine or accidental fad4i.a3v.a atm.ph5.

E IThi6 Gapability Is Fequired to evaluate the need fIT liniiatig prtective mneasures to pro-tA-ct thA health and safety of the public and is Gonsi6tGnt With the) 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 HOT 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 This specification deleted.

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, Revision 2, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1980.

The postaccident monitoring instrumentation limiting condition for operation provides the requirement of Type A and Category 1 monitors that provide information required by the control room operators to:

• Permit the operator to take preplanned manual actions to accomplish safe plant shutdown.

• Determine whether systems important to safety are performing their intended functions.

September 7, 1999 SEQUOYAH - UNIT 1 B 314 3-3 Amendment No. 62, 81, 149, 159, 238, 245 E3-4

RATR COOLANT SYSTEM BASES 3/4.4.7 CHEMISTRY 1This specification Is deleted.

Tho limitations on Reactor Cooiant System chemistry ensuro that corrosion of the Roactor Coolant Systom is minimized and roducos tho potontial for Reactor Coolant Systom leakage or failuro duo to stross corrosion. Maintaining tho chemistry Within tho Stoady Stato Wmits providos adequate corrosion protoction to onsuro tho structural integrity of tho Roactor Coolant Systom ovor tho lifo of tho plant. Tho associatod offocts of oxcoeding tho oxygon, chlorido and fluorido limits aro time and tomporaturo dopondont. Corrosion rstudios show that oporation may bo continued with containmont concentration lovels in eXCoss of tho Steady Stato Limits, up to tho Transient Limits, for the spocifiod limited timo intorvals without having a significant effoct on tho stwctural integrity of the Reactor Coolant Systom.

Tho time intorval pormitting continued oporation within tho restrictions of tho Transiont Limits providos time for taaking correctivo actions to restore the containmont concontrations to within tho Stoady Stato tifmi4& 7 August 4, 2000 SEQUOYAH - UNIT 1 B 3/4 4-4r Amendment No. 36,189, 214 E3-5

REACTOR COOLANT SYSTEM BASES Th surveillan rG qirmet provide adequate assuranne that cncnentrations, in exCes of th 9 wll e dtocod n suffic-ient time to take Gorrective action limis J-3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM. The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Sequoyah Nuclear Plant site, such as site boundary location and meteorological conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.35 microcuries/gram DOSE EQUIVALENT 1-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 0.35 microcuries/gram DOSE EQUIVALENT 1-131 but within the limits shown on Figure 3.4-1 should be limited to no more than 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> per year since the activity levels allowed by Figure 3.4-1 increase the 2-hour thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture. A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S) while relying on the ACTIONS. This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to power operations.

Reducing T.,g to less than 500OF prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.

Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

April 11, 2005 SEQUOYAH - UNIT 1 B 3/4 4-5 Amendment No. 117, 237, 301 E3-6

REACTOR COOLANT SYSTEM BASES 3/4.4.10 DELETED 314.4.11 REACTOR COOLANT SYSTEM HEAD VENTS This specification is deleted The funetien ef the RGS headl Yent: is le Feme':e niwi endensabhez or steaff frezm the Feeetcfr v-ssel head, Thuis ystem i; de ig^ned to mitigate a possible condition of inadequate core cooling, inadoguate natural circulation, or inability to dapressurizo the RHR System initiated conditions reOulting from the accumulation of non-condensable gases in the Reactor Coolant System. The reactor vessel head ver.t is deaigned ,It redunmant safety grade vemt pa1. a.

November 9, 2004 SEQUOYAH - UNIT 1 B 314 4-13 Amendment No. 116, 133,157, 208, 294, 297 E3 -7

PLANT SYSTEMS BASES 3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 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 with the assumptions used in the accident analyses.

3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses.

3/4.7.1.6 MAIN FEEDWATER ISOLATION, REGULATING. AND BYPASS VALVES Isolation of the main feedwater (MFW) system is provided when required to mitigate the consequences of a steam line break, feedwater line break, excessive feedwater flow, and loss of normal feedwater (and station blackout) accident. Redundant isolation capability is provided on each feedwater line consisting of the feedwater isolation valve (MFIV) and the main feedwater regulating valve (MFRV) and its associated bypass valve. The safety function of these valves is fulfilled when closed or isolated by a closed manual isolation valve. Therefore, the feedwater isolation function may be considered OPERABLE if its respective valves are OPERABLE, if they are maintained in a closed and deactivated position, or if isolated by a closed manual valve. The 72-hour completion time to either restore, close, or isolate an inoperable valve takes into account the redundancy afforded by the remaining OPERABLE valves and the low probability of an event occurring that would require isolation of the MFW flow paths during this time period. The 8-hour completion time for two inoperable valves in one flow path takes into account the potential for no redundant system to perform the required safety function and a reasonable duration to close or isolate the flow path. Although the steam generator can be Isolated with the failure of two valves in parallel, the double failure could be an indication of a common mode failure and should be treated the same as the loss of the isolation function. The 7-day frequency to verify that an inoperable valve is closed or isolated is reasonable based on valve status indications available in the control room, and other administrative controls to ensure the valves are closed or isolated.

2 34..2STEAM GENERATOR PRESSURE/TEMPERATURE LIMIAONa

( 1 Thispecification is deleted.

t The limitation on Eteam generator pro~sweu and temper~ature enswur' that the proseuro induied 64stress6s 1n the Fsteam.; generator de ;not oxceed the mnaximnum allowvable fractur8 toug1hn9ss str96 11imits.

I The limitatiaeis of :70mano 2nA pOig arte bara heRo anseaF qt~oeReate: DT- _f 251F nad Rare svuffinaient to-PF8VRwrwt brittl8 frantUF&ta June 8, 1998 SEQUOYAH - UNIT 1 B 3/4 7-3 Amendment No. 232 E3-8

PLANT SYSTEMS BASES

/34.710 EALED SOURCE CONTAMINTO J Thisspecificationisdelefed.

/Tho HimitatiGR6 on remoGvablo contamination for sourorquin leak terting, including alpha- \

omittrs, iS bao -o10 CFR 709.39(c) limnitS for pIlutoium. Thi£ liiato ... on6ur that leakage from byproduct, cource, and rpocial nuclear material courcec Will not oxceed allowablo intako valuec. Soalod courcoF aro classified into throo groupc according to their Uce, With requiromente rurveillance Gommensurate vith the probability of damage to a cource in that group. Tho1e 6ourcec which aro frequontly handled are roquirod to be tected moro ofton than thoRo which aro not. Soalod cource vhich l aro continuou61y enclored within a 6hielded mechanism (i.e., cealed courc within radiation monitoring J or boron measuring devicGe) are conGidred tobo to be ored and need not bo tected unlocs they are removed from tho 1hiRlded m9chanism.

This Specification is deleted.

August 28, 1998 SEQUOYAH - UNIT I B 3/4 7-7 Amendment No. 39, 227, 235 E3-9

REFUELING OPERATIONS BASES

/MMUNICATIONS 4 This specification is deleted.

Tho roguiromont for communications capability on9uroe that refuoling station personnol can bo

promptly informed of 6ignificant changoc in tho facility status or coro reactivity conditions during CORE )

ALTER'ATIONS.

3/4.9.6 MANIPULATOR CRANE l This specificationis deleted.

The OPERRABILITY roeuiromonts for the manipulator cranoe eonuro that: 1) manipulator cranoS will be used for movEmont of drivo rode and fuel acsomblioe, 2) each crano has cufficient load capacity to lift a drivo rod or fuel assembly, and 3) tho coro intornal6 and proseuro VOeSO6 are protocted from C N TRAVELs 'Snt FUEL PT AREA gagd during lifting 0pARAGA CRAN SPNT TRVEL- UELPIT REA~~L This specification is deleted.

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that;

1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140OF as required during the REFUELING MODE, and 2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification. The minimum required flow rate of 2000 gpm ensures decay heat removal, minimizes the probability of losing an RHR pump by air-entrainment from pump vortexing, and minimizes the potential for valve damage due to cavitation or chatter. Losing an RHR pump is a particular concern during reduced RCS inventory operation. The 2000 gpm value is limited by the potential for cavitation in the control valve and chattering in the 10-inch check valve.

The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor pressure vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling.

Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

June 14, 1995 SEQUOYAH - UNIT 1 B 3/4 9-2 Amendment No. 134, 167, 204 E3-10

REACTIVITY CONTROL SYSTEMS BASES The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with Tag greater than or equal to 5410F 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.

//..3.4 ROD DROP TIME (Deleted) 2-314.4.2-4 ROD-DR-OP TIME and3/4.1.3.5 SHUTDOWN ROD INSERTION LIMIT v Fully withdrawn for shutdown and control rod banks is defined as a condition where the rod banks are positioned in a range of 222 to 231 steps fully withdrawn. This range is defined to permit axial repositioning of rod banks to mitigate rod wear on Internal guide surfaces.

November 21, 1995 SEQUOYAH - UNIT 2 B 3/4 1-4a Amendment No. 98, 205 E3-11

INSTRUMENTATION BASES 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

Relative to the control room instrumentation isolation function, one set of process radiation monitors acts to automatically initiate control room Isolation. The actuation instrumentation consists of redundant radiation monitors. A high radiation signal from the detector will initiate its associated train of the Control Room Emergency Ventilation System (CREVS). The CREVS is also automatically actuated by a safety injection (SI) signal from either unit. The Si function is discussed in LCO 3.3.2, 'Engineered Safety Feature Actuation System (ESFAS) Instrumentation.' In addition, the control room operator can manually initiate CREVS.

3(4.3.3.2 MOVABLE INCORE DETECTORS 1 This specification Is deleted.

Tho OPERA'BILITY of tho movablo incoro dotoctors with tho spocifiod minimum complomont of gunimont oneurou that tho moaurromonts obtainod from usCof this cstRom accurately rFpFroSnt tho spatial nrUtFron flux diRtribltion of tho reactor cor. Thr. OER'II PWstom oTV f thiFs is domontrGatod by irradiating each dotector ueod and dotormining the accoptability of its voltago curvo.

FIor tho purFpo of moawuiRn F,4X,,Z) o7r OF 4(X, a full inro- flux m-nap is useod. Qua-rtor cor flux maps, a define in WCAP 8618, J 1976, may be sord i;n recalibration of tho excoro noRuton flux I d otoction systm, anrd full inoroF flux mnapsor symmetrir incoroR thimblesr,may be ued for monRitoring* tho QUADRANT POWER TILT PATIO ;on oRno PReowor Range Channel is inrablo.l /

May 31, 2000 SEQUOYAH - UNIT 2 B 3/4 3-2a Amendment No. 46, 72,182, 214, 228, 247 E3-12

INSTRUMENTATION BASES 3/4.3.3.3 SEISMIC INSTRUMENTATION This specificatin poelten a 3R INSTRUMENTATION 1Tis specif cation is deleted.

The QPERABILIlY of the reteshutownal instrumentation ensuresthat suffiieont petabFolgit is d available fer estimating pmtential radiaceof dHses to the publiY tfcy s a oFe-t f tinthepotaenidental capblit fofradiasseqen coldst-own atmospheoe. This oapabilityf the cnied to evaluate the need fori requiredting theevent contsures to phbteat the health and safetyof the publit and Is Dnesistnrt with the o Thi specidati of Rdeglt ide 1.23, "ed." 91ral Programs," Februay 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 HOT 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 Cerion 19 of 1F0 CFR 50.

3/4.3.3.6 CHLORINE DETECTION SYSTEMS This specification deleted.

314.3.3.7 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, Revision 2, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1980.

The postaccident monitoring instrumentation limiting condition for operation provides the requirement of Type A and Category I monitors that provide information required by the control room operators to:

Permit the operator to take preplanned manual action to accomplish safe plant shutdown September 7, 1999 SEQUOYAH - UNIT 2 B 314 3-3 Amendment No. 35, 46, 54, 72,135, 149, 228, 236 E3-13

REACTOR COOLANT SYSTEM BASES 3 HEMISTRY 4 This specificationIs deleted.

The limitations on Roactor Coolant System chomiStry ensure that corrosion of the Roactor Coolant system is minimized and reduceS the potential for Reactor Coolant System leakage or failure \

duo to stress corocion. Maintaining the chemistry AWthin the Steady Stato Limits provides adequate corrosion protection to the structural integrity of the Reactor Coolant Systom over the life of the Fnsure plant. The asrociated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependont. Corrosion studies show that operation may be continued with contaminant concentration leveIl in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time interals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Stead" S-tat e;ILi'mit.r, 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2-hour doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Sequoyah site, such as site boundary location and meteorological conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.35 microcuries/gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes In THERMAL POWER. Operation with specific activity levels exceeding 0.35 microcuries/gram DOSE EQUIVALENT 1-131 but within the limits shown on Figure 3.4-1 should be limited to no more than 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> per year since the activity levels allowed by Figure 3.4-1 increase the 2-hour thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture. A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S) while relying on the ACTIONS. This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to power operations.

April 11, 2005 SEQUOYAH - UNIT 2 B 314 4-5 Amendment No. 107, 227, 290 E3-14

0 REACTOR COOLANT SYSTEM BASES 3/4.4.10 DELETED 3/4.4.11 REACTOR COOLANT SYSTEM HEAD VENTS I This specificationis deleted.

The function of the RCS head ve.ontc i£to remove non condencabloc or ntram from tho roactor Vo£EOI hoad. This syrtom I6 decignod to mitigate a poccible condition of inadequate coro cooling, inadequato natural circulation, or inability to deprFseurize tho RHR Systom initiated conditions rosulting from the accumulation of non condonRable ga6ec in the Reactor Coolant Syetom. Tho reactor vaesel head vont Ic decigned vAth redundant safety grade vent path6.

September 15, 2004 SEQUOYAH - UNIT 2 B 3/4 4-14 Amendment No. 106, 120, 147, 148,198, 277, 284 E3-15

PLANT SYSTEMS BASES 314.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 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 with the assumptions used in the accident analyses.

314.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses.

3/4.7.1.6 MAIN FEEDWATER ISOLATION. REGULATING. AND BYPASS VALVES Isolation of the.main feedwater (MFW) system Is provided when required to mitigate the consequences of a steam line break, feedwater line break, excessive feedwater flow, and loss of normal feedwater (and station blackout) accident. Redundant isolation capability is provided on each feedwater line consisting of the feedwater isolation valve (MFIV) and the main feedwater regulating valve (MFRV) and its associated bypass valve. The safety function of these valves is fulfilled when closed or isolated by a closed manual isolation valve. Therefore, the feedwater isolation function may be considered OPERABLE if its respective valves are OPERABLE, if they are maintained in a closed and deactivated position, or if isolated by a closed manual valve. The 72-hour completion time to either restore, close, or isolate an inoperable valve takes into account the redundancy afforded by the remaining OPERABLE valves and the low probability of an event occurring that would require isolation of the MFW flow paths during this time period. The 8-hour completion time for two inoperable valves in one flow path takes into account the potential for no redundant system to perform the required safety function and a reasonable duration to close or isolate the flow path. Although the steam generator can be isolated with the failure of two valves in parallel, the double failure could be an indication of a common mode failure and should be treated the same as the loss of the isolation function. The 7-day frequency to verify that an inoperable valve is closed or isolated is reasonable based on valve status indications available in the control room, and other administrative controls to ensure the valves are closed or isolated.

(34.72 TEAM GENERATOR PRESSUREITEMPERATURE LIMITAIO 1Tis specificationis deleted. A INhe litmaitatiof on steam pressure and temperatwro eRSUro that the primsuowfiindt t Fenator

<sresesin the steam generatorr, do not eXGGod th9 mRaximum allowable fracture toughness; 6tr#s imts.

prevent brittle fracture-.

June 8, 1998 SEQUOYAH - UNIT 2 B 3/4 7-3 Amendment No. 222 E3-16

BASES

/ 34.710SEALED SOURCE CONTAMINATIONN 1Tis specificationis deleted, The limitstione oR FBFROVable GGntaFAInatiGR for G9UFoe6 roqui4rin leak toeting, including vpha omittore , baod o 'IOn0CR 7n.39(c) limits for pluto-nium. Thi limitation wi- onWurlthat leakage frFom byproduct, sourco, and pepcial nuczloar mstorisl courcec will not excood aIlowablo intako values. Soalod eource soclaifiod are into three groupe according to their use, with curVoilIanco requirement6 CommenEUrate with tho probability of damage to a eource in that group. Those source which are frFquently handlod are roquirod to bo tested moro ofton than thoso which aro not. Soaled eource which are continuouely onclosed within a shielded mechanism (i.o., soalod eourcoe within radiation monitoring or boron moaguring devicoe) are coneidored to bo etored and need not bo toetod unlocs they are romovod from tho shiolded mechanism.

August 28, 1998 SEQUOYAH - UNIT 2 B 3/4 7-6a Amendment Nos. 31, 143, 225 E3-17

REFUELING OPERATIONS BASES l This specification Is deleted.

Tho OE Broquiromont Tf GOmFURIcatioRS or mapabnlity that Fefuelir:ig stationiputorel crano cror, plrlptly ibudformedovomot of rifantrhadges i the fauolity statum o2cah reantivity cnfitiontloas durinpaci lAL-TFRATIQN9S.

/ 3/4.9.6 MANIPULATOR CRANE This specif cation is deleted.

The OPERABILITY Fequirom9Rt6 for the m~anipulator Granes eRsur9 that: 1) mnanipulator Gcran!es xvill b m sd fo~r mov9Femetof drive ro!de and fuel a6sseblieG,^) eachrchint hasr~sufi9Rientadrcapacity to lift a drivo rod or fuAl aceombly, assembly, and 3) the core internals and pro6sure Voc6ol are protected from oxceccivo lifting force in the ovent they are inadvortently engaged during lifting operationes.-

3/4.9.7 CRANE TRAVEL - SPENT FUEL PIT AREA This specification is deleted.

314.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that;

1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140OF as required during the REFUELING MODE, and 2) sufficient coolant circulation is maintained thru the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification. The minimum required flow rate of 2000 gpm ensures decay heat removal, minimizes the probability of losing an RHR pump by air-entrainment from pump vortexing, and minimizes the potential for valve damage due to cavitation or chatter. Losing an RHR pump is a particular concern during reduced RCS inventory operation. The 2000 gpm value is limited by the potential for cavitation in the control valve and chattering in the 10-inch check valve.

The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor pressure vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling.

Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

June 14, 1995 SEQUOYAH - UNIT 2 B 3/4 9-2 Amendment No. 121, 157, 194 E3-18