Text

License Amendment Request (LAR) 12-01, Proposed Change to Establish the Reactor Protective System (RPS) Actuation Circuits Limiting Condition for Operation (LCO)
	ML12046A838
Person / Time
Site:	Fort Calhoun
Issue date:	02/10/2012
From:	Bannister D; Omaha Public Power District
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	LAR 12-01, LIC-12-0006
	Download: ML12046A838 (62)
	v • d • e

Omaha Publi Powe Obwdc 444 South 16th Street Mall Omaha, NE 68102-2247 LIC-1 2-0006 February 10, 2012 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Reference:

Docket No. 50-285

SUBJECT:

Fort Calhoun Station (FCS) License Amendment Request (LAR) 12-01, Proposed Change to Establish the Reactor Protective System (RPS)

Actuation Circuits Limiting Condition for Operation (LCO)

Pursuant to 10 CFR 50.90, the Omaha Public Power District (OPPD) hereby requests an amendment to the Renewed Facility Operating License No. DPR-40 for Fort Calhoun Station (FCS), Unit No. 1. The proposed amendment would establish the limiting condition for operation (LCO) requirements for the reactor protective system (RPS) actuation circuits in Technical Specification (TS) 2.15.

These TS revisions will result in the TS LCOs and surveillance requirements (SRs) being more aligned with NUREG-1432, Standard Technical Specifications, Combustion Engineering Plants, Revision 3, for RPS requirements. Specifically, this proposed change: renumbers LCO 2.15(1) through 2.15(4) to 2.15.1(1) through 2.15.1(4), renumbers LCO 2.15(5) to LCO 2.15.3 with an associated Table 2-6, and implements a new LCO 2.15.2 for the reactor protective system logic and trip initiation channels.

The Table of Contents is also revised to reflect the renumbering and addition of the LCO for the reactor protective system logic and trip initiation channels and the new Table 2-6.

Currently, the TS contain surveillance requirements in TS 3.1, Table 3-1, Item 12, for a quarterly functional test of the RPS logic units, and TS 3.1 Table 3-1, Item 10 for a prior to critical functional test of manual trips, but contains no LCO for the RPS logic units. These TS revisions will result in the TS LCOs and SRs for the RPS logic units and manual trips being similar to the Palisades plant which has a similar design for the reactor trip initiation channels.

Ac~ci Employment with Equal Opportunity

U. S. Nuclear Regulatory Commission LIC-1 2-0006 Page 2 The proposed TS changes conform to NRC regulation 10 CFR 50.36 for the contents of the Technical Specifications.

OPPD concludes that the proposed LAR presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment.

The enclosure contains OPPD's evaluation of the proposed changes, including the supporting technical evaluation, and the significant hazards consideration determination. Attachment 1 provides the existing TS and TS Bases pages marked-up to show the proposed changes to TS 2.15. Attachment 2 provides the associated retyped (clean) TS and TS Bases pages.

OPPD requests approval of the proposed amendment by March 1, 2013. Once approved, the amendment shall be implemented within 180 days.

There are no regulatory commitments associated with this proposed change.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State of Nebraska official.

If you should have any questions regarding this submittal or require additional information, please contact Mr. Bill R. Hansher at 402-533-6894.

I-declare under penalty of perjury that the foregoing is true and correct. Executed on February 10, 2012.

David i.Bannister Vice President and CNO DJB/BRH/brh

Enclosure:

OPPD's Evaluation of the Proposed Changes c: E. E. Collins, Jr., NRC Regional Administrator, Region IV L. E. Wilkins, NRC Project Manager J. C. Kirkland, NRC Senior Resident Inspector Director of Consumer Health Services, Department of Regulation and Licensure, Nebraska Health and Human Services, State of Nebraska

LIC-1 2-0006 Enclosure Page 1 OPPD's Evaluation of the Proposed Change(s)

Fort Calhoun Station (FCS) License Amendment Request (LAR) 12-01, Proposed Change to Establish the Reactor Protective System (RPS) Actuation Circuits Limiting Condition for Operation (LCO) 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

ATTACHMENTS:

1. Technical Specification and Information Only Bases Pages Markups

2. Retyped ("Clean") Technical Specifications and Information Only Bases Pages

LIC-12-0006 Enclosure Page 2 1.0

SUMMARY

DESCRIPTION This license amendment request (LAR) proposes a change to Renewed Facility Operating License No. DPR-40 for Fort Calhoun Station (FCS), Unit No. 1. The Omaha Public Power District (OPPD) proposes to revise the Technical Specification (TS) limiting condition for operation (LCO) 2.15, Instrumentation and Control Systems. Specifically, this proposed change: renumbers LCO 2.15(1) through 2.15(4) to 2.15.1(1) through 2.15.1(4), renumbers LCO 2.15(5) to LCO 2.15.3 with an associated Table 2-6, and implements a new LCO 2.15.2 for the reactor protective system logic and trip initiation channels.

The Table of Contents is also revised to reflect the renumbering and addition of LCO 2.15.2 for the reactor protective system logic and trip initiation channels and the relocation of TS 2.15(5) to a new LCO 2.15.3 and new Table 2-6. Currently, the TS contain surveillance requirements in TS 3.1, Table 3-1, Item 12, for a quarterly functional test of the RPS logic units, and.TS 3.1 Table 3-1, Item 10 for a prior to critical functional test of manual trips, but contains no LCO for the RPS logic units. These TS revisions will result in the TS LCOs and SRs for the RPS logic units and manual trips being similar to the Palisades plant which has a similar design for the reactor trip initiation channels.

2.0 DETAILED DESCRIPTION The proposed TS changes for LAR 12-01 are as follows:

" TS LCO 2.15, Instrumentation and Control Systems o Renumbered to LCO 2.15.1 for paragraphs (1) through (4) with appropriate renumbering of footnotes in Tables 2-2 through 2-5 that reference these paragraphs.

o Conversion of LCO 2.15(5) with its list of components into a new LCO 2.15.3 with the list of components being included into a new Table 2-6.

New TS LCO 2.15.2 for the reactor protective system logic and trip initiation channels.

Deletion of reactor protective system manual trip functional unit from TS Table 2-2 and inclusion into the new LCO 2.15.2.

These proposed changes to TS 2.15 will result in the FCS LCO 2.15 being more aligned with NUREG 1432, Standard Technical Specifications, Combustion Engineering Plants for the RPS.

LIC-12-0006 Enclosure Page 3

3.0 TECHNICAL EVALUATION

=

System Description===

The Reactor Protective System (RPS) is shown in Updated Safety Analysis Report (USAR)

Figures 7.2-1 and 7.2-2 (References 6.1 and 6.2). It consists of four channels of instrumentation. Each channel monitors 12 safety parameters, each parameter input is derived from an isolated instrument channel. Each parameter operates a two out of four coincidence logic matrix to maintain OR remove power from the Control Element Drive Mechanism (CEDM) clutches. Individual channel trips occur when the measurement reaches a preselected value. A typical measurement channel functional diagram is shown in USAR Figure 7.2-3. The channel trips are combined in six two-out-of-four matrices. Each individual measurement channel has inputs to three of the six logic matrices. The logic matrix trip relays are de-energized when two channels of the same measurement channel trip. Each two-out-of-four logic matrices provides trip signals to the interposing relays which in turn cause a direct trip of the contactors in the a-c supply to the CEDM clutch power supplies.

Any one of the six logic matrices will de-energize the 4 clutch power supplies. The logic matrices are arranged in a one,-out-of-six logic configuration. The clutch power supply DC outputs are ungrounded.

Reactor trip is accomplished by de-energizing the magnetic clutch holding coils and releasing the control element assemblies (CEAs) to drop into the core.

The Reactor Protective System was designed under the following bases to assure adequate protection for the reactor core (Reference 6.3):

a. Instrumentation conformed to the provisions of the proposed IEEE Standard for Nuclear Power Plant Protective Systems (IEEE 279, August 1968).

b. No single component failure can prevent safety action.

c. Four independent measurement channels, each complete with sensors, sensor power supply units, amplifiers, and bistable modules were provided for each parameter.

d. The channels are provided with a high degree of independence by separate connection of the sensors to the process systems. Separate raceways are used to segregate cable systems.

e. The four measurement channels provide trip signals to four independent trip paths.

f. A trip signal from any two trip units monitoring the same parameter causes a reactor trip.

g. When one of the four channels is taken out of service for maintenance, the protective system logic can be changed to a two-out-of-three coincidence for a reactor trip by bypassing the out-of-service channel.

h. The protective system AC power is supplied from four separate instrument buses.

i. Open circuiting, or loss of power supply for the channel logic, initiates an alarm and a channel trip.

j. All measurement channels and trip logic matrices assume the de-energized state to provide a tripping function.

LIC-12-0006 Enclosure Page 4

k. The Reactor Protective System can be tested with the reactor in operation or shutdown.

I. A manual trip, independent of the automatic trip system, is provided.

m. Trip signals are preceded by alarms to alert the operator of undesirable operating conditions in cases where the operator could avert a reactor trip by taking timely corrective actions.

n. The Reactor Protective System components are independent of the Power Range Monitor control channels.

Coincidence Logic Matrices The RPS has four separate channels having twelve trip units per channel. Each of the twelve trip units serves to monitor a different plant parameter. There are four trip units for each plant parameter monitored, one per channel.

There are six logic trip matrices (AB, BC, BD, AC, CD, and AD). For matrix AB, the normally open contact from the channel A trip unit 1 relay 1 (Al-1) is connected in parallel with the channel B trip unit 1 relay 1 (B1-1).

Each trip unit contains three sealed trip relays which have a single-pole, double-throw (SPDT) contact. The normally open contact from channel A trip unit 1 relay 1 (Al-1) is connected in parallel with the normally open contact from channel B trip unit 1 relay 1 (B1-1).

This is similarly done for the twelve contact combinations Al-1 through A12-1 and B1-1 through B12-1 and these twelve parallel combinations are connected in series to form a logic ladder. The trip unit relays, are energized in a reset condition, thus the normally open relay contacts are closed. The AB logic ladder serves to control four matrix relays which are energized in the reset condition. The three trip unit trip relays from the four channels are used to make six logic matrices in the same fashion as the AB matrix.

A normally open contact from each of the four matrix relays are connected in series with a normally open contact from the corresponding relays of the other five matrices to form four trip paths. A contact from one of the manual trip switches and the coil of an interposing relay are in series with the six matrix relay contacts for each of the four trip paths. Under normal operating conditions the four interposing relays are energized. A normally open contact from each of the interposing relays serves to control an M-contactor.

The four M-contactors are combined into pairs with the contacts of each pair connected in series. The series contacts of the two pairs serve to supply AC power to the CEDM clutch power supplies. The CEDM clutches are separated into two groups. The clutches in each group are supplied by parallel pairs of low voltage, d-c power supplies which are fed by an ungrounded supply via contacts from the two pairs of M-contactors. The parallel CEDM clutch power supplies assure that the inadvertent loss of one supply source will not de-energize the clutches.

A block diagram of the RPS is provided below:

LIC-1 2-0006 Enclosure Page 5 REACTOR PROTECTIVE BLOCK DIAGRAM

LIC-12-0006 Enclosure Page 6 Currently, the TS contain surveillance requirements in TS 3.1, Table 3-1, Item 12, for a quarterly functional test of the RPS logic units, and TS 3.1 Table 3-1, Item 10 for a prior to critical functional test of manual trips, but contains no LCO for the RPS logic units. Since there is no LCO, if an RPS logic unit or trip initiation channel becomes inoperable, TS 2.0.1 would apply. TS 2.0.1 specifies corrective measures to be employed for system conditions not covered by LCOs. The corrective measures include placing the unit in Hot Shutdown within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , in at least subcritical and less than 300 degrees F within an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and placing the unit in Cold Shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . In certain situations these actions are non-conservative compared to NUREG 1432, (Reference 6.4). As an example, NUREG 1432 requires the reactor trip breakers be opened immediately for certain conditions.

Applicability The proposed TS would revise the applicability for when the RPS logic and trip initiation channels are required to be operable to include whenever CEAs are capable of being withdrawn and the reactor coolant system is not at refueling boron concentration. The definition section of the TS includes definitions for both CEAs and non-trippable CEAs. As the RPS trip initiation channels do not trip non-trippable CEAs, the LCO only applies to CEAs that are defined as all full length shutdown and regulating control rods. By TS definition, Mode 4 is RCS less than 210 degrees F with a boron concentration of greater than or equal to shutdown boron concentration but less than refueling boron concentration, Mode 5is RCS less than 210 degrees F with a boron concentration of greater than or equal to refueling boron concentration.

RPS trip functions are not required while in modes of operation when the RCS boron concentration is at refueling boron concentration, or when no more than one trippable control rod is capable of being withdrawn, because the RPS function is already fulfilled. Refueling boron concentration provides sufficient negative reactivity to assure the reactor remains subcritical regardless of control rod position, and the safety analyses assume that the highest worth withdrawn full-length control rod will fail to insert on a trip. Therefore, under these conditions, the safety analyses assumptions will be met without the RPS trip function.

Specification The proposed TS would apply to the six channels of RPS logic matrices and four channels of RPS trip initiation channels which do not currently have a separate LCO, and the two channels of manual trip initiation which are currently contained in TS 2.15, Table 2-2, Item 1.

Required Actions With one RPS Logic Matrix channel inoperable, it is proposed to restore the inoperable channel to operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The completion time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months provides the operator time to take appropriate actions and still ensures that any risk involved in operating with a failed channel is acceptable.

LIC-1 2-0006 Enclosure Page 7 With one RPS Manual Trip channel inoperable, the required action is to restore the inoperable channel to operable status prior to entering Mode 2 from Mode 3. This action is consistent with the current TS requirement contained in TS Table 2-2, Item 1. There are two installed Manual Trip channels. TS 2.15, Table 2-2, Item 1 requires that there be at least one operable. No safety analyses assume operation of the Manual Trip. Because of this, the required action is to restore the inoperable channel to operable status prior to entering Mode 2 from Mode 3 during the next plant startup.

With one RPS Trip Initiation logic channel inoperable, it is proposed to de-energize the affected clutch power supply within one hour. This removes the need for the affected channel by performing its associated safety function. With the clutch power supplies associated with one initiation logic channel de-energized, the remaining clutch power supplies prevent control rod clutches from de-energizing. The remaining clutch power supplies are in a one-out-of-two logic with respect to the remaining initiation logic channels in the clutch power supply path. This meets redundancy requirements, but testing on the operable channels cannot be performed without causing a reactor trip.

With two inoperable RPS Trip Initiation logic channels affecting the same trip leg, it is proposed to de-energize the affected clutch power supplies immediately. With both channels inoperable, the RPS function is lost if the affected clutch power supplies are not de-energized. Therefore, immediate action is required to de-energize the affected clutch power supplies. The immediate completion time is appropriate since there could be a loss of safety function if the associated clutch power supplies are not de-energized.

With the required actions not met, or with two RPS Manual Trip channels inoperable or with two or more RPS logic matrices inoperable, or with two or more RPS Trip Initiation logic channels inoperable that do not affect the same trip leg, it is proposed that within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; the station is placed in Hot Shutdown, it is verified that no more than one CEA can be withdrawn, or the RCS is verified to be at refueling boron concentration. In this condition the reactor must be placed in a Mode in which the LCO does not apply. The completion time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to be in Mode 3 is reasonable, based on operating experience, to reach the required Mode from full power conditions in an orderly manner and without challenging plant systems.

The requirements of TS 2.15(5) are being incorporated into a new LCO 2.15.3 with the list of components being included into a new Table 2-6. No changes are proposed for the requirements other than formatting to be more consistent with the remainder of TS 2.15 by listing the components in a Table.

These TS revisions will result in the TS LCOs and SRs for the RPS logic units and manual trips being similar to the Palisades plant which has a similar design for the reactor trip initiation channels.

LIC-12-0006 Enclosure Page 8 10 CFR Part 50.36 Criteria:

10 CFR 50.36(c)(2)(ii) states that "A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

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

(B) Criterion 2 - A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.,

(C) Criterion 3 - A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(D) Criterion 4 - A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety."

NRC-approved NUREG 1432, Standard Technical Specifications, Combustion Engineering Plants, Revision 3 (Reference 6.4), identifies an improved TS that was developed based on the screening criteria in the Commission's Final Policy Statement (Reference 6.5) and subsequently codified in 10 CFR 50.36. The RPS actuation circuits satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii) (described above) and their operability will be required by the proposed changes to FCS TS 2.15.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.1.1 Regulations Code of Federal Regulations Part 50:

10 CFR 50.36, Technical Specifications: 10 CFR 50.36(c)(2) states, "When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met." The revised actions will establish limiting conditions for operation for the RPS logic matrices and RPS trip initiation channels which do not currently have a separate LCO and therefore the proposed change will meet the requirements of this regulation.

LIC-1 2-0006 Enclosure Page 9 10 CFR 50.36(c)(3) criteria states that "surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met."

The revised actions will maintain the surveillance requirements on the RPS logic units and manual trip functions and therefore the proposed change will continue to meet the requirements of this regulation.

Fort Calhoun Station (FCS), Unit No. 1 was licensed for construction prior to May 21, 1971, and is committed to the draft General Design Criteria (GDC) published for comment in the Federal Register on July 11, 1967 (32 FR 10213) in lieu of 10 CFR 50, Appendix A. Appendix G of the FCS Updated Safety Analysis Report (USAR) shows that draft GDC 12, 14, 19, and 25 are most applicable to the proposed amendment.

CRITERION 12 - INSTRUMENTATION AND CONTROL SYSTEMS Instrumentation and controls shall be provided as required to monitor and maintain variables within prescribedoperatingranges.

This criterion is met. Instrumentation is provided for continuous measurement of all significant process variables. Controls are provided for the purpose of maintaining these variables within the limits prescribed for safe operation. The instrumentation conforms to applicable Institute of Electrical and Electronics Engineers (IEEE)standards.

The principalprocess variablesmonitored include neutron level (reactorpower);

reactor coolant temperature, flow, and pressure; pressurizer liquid level; and steam generatorlevel. In addition, instrumentation is provided for continuous automatic monitoring of radiation level. The instrumentation and control systems are described in detail in USAR Section 7.

The proposed license amendment request provides for addition of the RPS logic operability LCO. No physical changes are being made to the plant. This criterion will continue to be met.

CRITERION 14- CORE PROTECTION SYSTEMS Core protection systems, together with associated equipment, shall be designed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits.

This criterion is met. The reactoris protected by the reactorprotection system from reaching a condition at which fuel damage might occur. The protection

LIC-12-0006 Enclosure Page 10 system is designed to monitor the reactoroperatingconditions and initiate a fast shutdown if any of measured variables exceed the operatinglimits.

The signals which will provide automatic reactortrip are identified in Table 7.2-1 of the Fort Calhoun Station, Unit No. 1 USAR. The parametersand conditions which will initiate a trip are the following:

a) High Neutron Level (reactorpower) (A T power is a backup) b) High Startup Rate (low power level only) c) High PressurizerPressure d) Thermal Margin/Low Pressure e) Loss of Load f) Low Steam GeneratorPressure g) Low Reactor Coolant Flow h) Low Steam GeneratorLiquid Level i) ContainmentBuilding High Pressure j) Steam GeneratorDifferential Pressure The proposed license amendment request provides for addition of the RPS logic operability LCO. Applicability of the LCO will ensure that the RPS is capable of being tripped whenever a CEA is capable of being withdrawn and the RCS is not at refueling boron concentration. No physical changes are being made to the plant. This criterion will continue to be met.

CRITERION 19- PROTECTION SYSTEMS RELIABILITY Protectionsystems shall be designed for high functional reliabilityand in-service testability commensurate with the safety functions to be performed.

This criterionis met. Design of protection systems includes specification of high quality components, ample design capacity, component redundancy, and in-service testability. The following principaldesign criteria have been applied:

a) No single component failure shall prevent the protection systems from fulfilling their protective function when action is required.

b) No single component failure shall initiate unnecessaryprotection system action provided implementation does not conflict with the criterion above.

Testing facilities are built into the protection systems to provide for:

a) Preoperationaltesting to give assurancethat the protection systems can fulfill theirrequired functions.

b) In-service checking of protective channels from the process sensorto the channel trip unit (bistable).

LIC-1 2-0006 Enclosure Page 11 c) In-service testing of the channel trip units (bistables) and associated coincidence logic and the outputs of that logic through to the final actuator.

The proposed license amendment request provides for addition of the RPS logic operability LCO. Surveillance requirements to perform quarterly functional tests of the RPS logic units remain unchanged. No physical changes are being made to the plant. This criterion will continue to be met.

CRITERION 25 - DEMONSTRATION OF FUNCTION OF FUNCTIONAL OPERABILITY OF PROTECTION SYSTEM Means shall be included for testing protection systems while the reactor is in operation to demonstrate that no failure or loss of redundancyhas occurred.

This criterion is met. Protection systems, from the sensors up to the final protection element, will be capable of being checked during reactor operation, as follows:

a) Measurement channels used in protection systems will be checked by observing outputs of similar channels and cross checking with related measurements which are presented on indicators and recorders on the control board.

b) Trip units and logic will be tested by inserting a signal into the measurement channel ahead of the readout and, upon application of a trip level input, observing that a signal is passed through the trip units and the logic to the logic output relays.

c) The logic output relays will be tested individually for initiation of trip action.

The proposed license amendment request provides for addition of the RPS logic operability LCO. No physical changes are being made to the plant.

Currently, the TS contain surveillance requirements in TS 3.1, Table 3-1, Item 12, for a quarterly functional test of the RPS logic units, and TS 3.1 Table 3-1, Item 10 for a prior to critical functional test of manual trips. No changes to these surveillance requirements are being made. This criterion will continue to be met.

LIC-1 2-0006 Enclosure Page 12 4.1.2 Design Basis The design basis requirement for the RPS was discussed in the Technical Evaluation Section 3.0 as it relates to the RPS logic and trip functions and their associated USAR sections.

4.1.3 Approved Methodologies There are no new specific approved methodologies associated with this proposed TS change.

4.1.4 Analysis No new analyses were needed in support of this proposed LAR.

4.2 Precedent As noted in Section 3.0 above, precedent for adding the TS LCOs and SRs for RPS logic and trip initiation circuits is similar to NUREG 1432 for the reactor trip breakers. The design of the FCS RPS is similar to that of the Palisades plant in that there are M-contactors instead of reactor trip breakers. This submittal is consistent with these documents; however, no plant-specific precedence is cited in this LAR.

4.3 Significant Hazards Consideration The proposed change would modify Technical Specification (TS) 2.15 to include provisions for the reactor protective system logic and trip initiation circuits.

These proposed changes are aligned with NUREG 1432, Standard Technical Specifications, Combustion Engineering Plants, Revision 3. In addition, administrative changes are being made to TS 2.15(5), in that the TS is being reformatted to list components in a table consistent with the format of the remainder of TS 2.15.

The Omaha Public Power District (OPPD) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No.

The reactor protective system logic and trip initiation channels meets Criterion 3 of 10 CFR 50.36 for inclusion into Technical Specification (TS) as a component that is part of the primary success path and which functions or actuates to

LIC-12-0006 Enclosure Page 13 mitigate a design basis accident or transient. The TSs currently do not have limiting conditions for operation (LCO) specific for this circuitry, but does contain surveillance requirements. The addition of LCOs provides additional restrictions on the operation of the plant and provides required actions and time limits if these components are incapable of performing their function. As such, the proposed change does not increase the probability of an accident. The proposed changes do not alter the physical design of the RPS, or any other plant structure, system or component (SSC) at Fort Calhoun Station (FCS).

The proposed changes conform to the Nuclear Regulatory Commission's (NRC's) regulatory guidance regarding the content of plant TS as identified in 10 CFR 50.36 and NRC publication NUREG 1432.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed TS changes do not alter the physical design, safety limits, or safety analysis assumptions associated with the operation of the plant. Hence, the proposed changes do not introduce any new accident initiators, nor do they reduce or adversely affect the capabilities of any plant structure or system in the performance of their safety function.

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 amendment involve a significant reduction in a margin of safety?

Response: No.

The TS operability requirements for the RPS logic and trip initiation channels ensure there is adequate components operable to assure safe reactor operation and are necessary to ensure safety systems accomplish their safety function for design basis accident events. The proposed TS would revise the applicability for when the RPS logic and trip initiation channels are required to be operable to include whenever control element assemblies (CEAs) are capable of being withdrawn and the reactor coolant system (RCS) is not at refueling boron concentration.' When the RCS boron concentration is at refueling boron concentration, or when no more than one trippable control rod is capable of being withdrawn, the RPS function is already fulfilled. These proposed TS changes for the RPS are aligned with the applicability and operability requirements provided in NUREG 1432.

LIC-1 2-0006 Enclosure Page 14 Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, OPPD concludes that the proposed amendment presents 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.

4.4 Conclusions 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.

5.0 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 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

6.1 USAR, Figure 7.2-1, Reactor Protective System 6.2 USAR, Figure 7.2-2, Reactor Protective System Functional Diagram 6.3 USAR Section 7.2, Instrumentation and Control, Reactor Protective Systems, Revision 14, dated August 19, 2010 6.4 NUREG 1432, Revision 3, Standard Technical Specifications, Combustion Engineering Plants, dated June 2004 6.5 NRC "Final Policy Statement on Technical Specifications Improvement for Nuclear Power Reactors" (58 FR 39132, dated July 22, 1993)

LIC-12-0006 Enclosure, Attachment 1 Page 1 Technical Specifications and Information Only Bases Pages Markups

TECHNICAL SPECIFICATION TABLE OF CONTENTS (Continued) 2.13 Limiting Safety System Settings, Reactor Protective System 2.14 Engineered Safety Features System Initiation Instrumentation Settings 2.15..1 Instrumentation and Control Systems 21,5.2 L Reactor Protective System (RPS)'Log -candTrip Initiation 2.15.3 Alternate Shutdown and Auxilary Feedwater Panel 2.16 River Level 2.17 Miscellaneous Radioactive Material Sources 2.18 DELETED 2.19 DELETED 2.20 Steam Generator Coolant Radioactivity 2.21 Post-Accident Monitoring Instrumentation 2.22 DELETED 2.23 Steam Generator (SG) Tube Integrity 3.0 SURVEILLANCE REQUIREMENTS 3.1 Instrumentation and Control 3.2 Equipment and Sampling Tests 3.3 Reactor Coolant System and Other Components Subject to ASME XI Boiler and Pressure Vessel Code Inspection and Testing Surveillance 3.4 DELETED 3.5 Containment Test 3.6 Safety Injection and Containment Cooling Systems Tests 3.7 Emergency Power System Periodic Tests 3.8 Main Steam Isolation Valves 3.9 Auxiliary Feedwater System 3.10 Reactor Core Parameters 3.11 DELETED 3.12 Radioactive Waste Disposal System 3.13 Radioactive Material Sources Surveillance 3.14 DELETED 3.15 DELETED 3.16 Residual Heat Removal System Integrity Testing 3.17 Steam Generator (SG) Tube Integrity 4.0 DESIGN FEATURES 4.1 Site 4.2 Reactor Core 4.3 Fuel Storage TOC - Page 2 Amendment No. 11,27,32,38,43,46,54,,69,84 aA 6, 93,97,104,122,436,152,L 160,17-6,183, 2114,230, 236, 216,248, 2-52

TECHNICAL SPECIFICATION TECHNICAL SPECIFICATIONS - TABLES TABLE OF CONTENTS TABLE SECTION 2-1 ESFS Initiation Instrum entation Setting Lim its ......................................................................... Section 2.14 2-2 Instrument Operating Requirements for RPS ................. Section 2.155i 2-3 Instrument Operating Requirements for Engineered Safety Features ................................. Section 2.151 2-4 Instrument Operating Conditions for Isolation Functions ...................................................... Section 2.1555 2-5 Instrumentation Operating Requirements for Other Safety Feature Functions .................... Section 2.15'F 26 Alternate Shutdown and Auxiliary FeedwaterPanelFunctions, .................... . Section 2.15.3 2-9 RC S Pressure Isolation Valves .................................................................................................. Section 2.1 2-10 Post-Accident Monitoring, Instrumentation Operating Limits ................................................... Section 2.21 2-11 RPS Limiting Safety System Settings ................................... Section 2.13 3-1 Minimum Frequencies for Checks, Calibrations, and Testing of RPS.: ..................................... Section 3.1 3-2 Minimum Frequencies for Checks, Calibrations and Testing of ............................................... Section 3.1 Engineered Safety Features, Instrumentation and Controls 3-3 Minimum Frequencies for Checks, Calibrations, and Testing ................................................... Section 3.1 of Miscellaneous Instrumentation and Controls 3-3a Minimum Frequency for Checks, Calibrations and Functional .................................................. Section 3.1 Testing of Alternate Shutdown Panels (AI-185 and AI-212) and Emergency Auxiliary Feedwater Panel (Al-1 79) Instrumentation and Control Circuits 3-4 Minim um Frequencies for Sam pling Tests ................................................................................. Section 3.2 3-5 Minim um Frequencies for Equipm ent Tests .............................................................................. Section 3.2 3-6 Reactor Coolant Pum p Surveillance .......................................................................................... Section 3.3 5.2-1 Minim um Shift C rew Com position ............................................................................................. Section 5.0 TOC - Page 4 Amendment No. 116,136,160, 246,248, 2,

TECHNICAL SPECIFICATION TECHNICAL SPECIFICATIONS - TABLES TABLE OF CONTENTS (ALPHABETICAL ORDER)

TABLE DESCRIPTION SECTION 2-1 ESFS Initiation Instrumentation Setting Limits ......................................................................... Section 2.14 2-4 Instrument Operating Conditions for Isolation Functions ...................................................... Section 2.155i 2-2 Instrument Operating Requirements for RPS ..................................................................... Section 2.15,ij 2-3 Instrument Operating Requirements for Engineered Safety Features ................................. Section 2.15.JI 2-5 Instrumentation Operating Requirements for Other Safety ................................................. Section 2.152 Features Functions 3-3a Minimum Frequency for Checks, Calibrations and Functional .................................................. Section 3.1 Testing of Alternate Shutdown Panels (Al-185 and AI-212) and Emergency Auxiliary Feedwater Panel (Al-1 79) Instrumentation and Control Circuits 3-2 Minimum Frequencies for Checks, Calibrations and Testing of ............................................... Section 3.1 Engineered Safety Features, Instrumentation and Controls 3-3 Minimum Frequencies for Checks, Calibrations, and Testing .................................................. Section 3.1 of Miscellaneous Instrumentation and Controls 3-1 Minimum Frequencies for Checks, Calibrations ....................................................................... Section 3.1 and Testing of RPS 3-5 Minimum Frequencies for Equipment Tests .............................................................................. Section 3.2 3-4 Minimum Frequencies for Sampling Tests ................................................................................. Section 3.2 5.2-1 Minimum Shift Crew Composition .............................................................................................. Section 5.0 2-10 Post-Accident Monitoring Instrumentation Operating Limits .................................................... Section 2.21 2-9 RCS Pressure Isolation Valves .................................................................................................. Section 2.1 3-6 Reactor Coolant Pump Surveillance .......................................................................................... Section 3.3 2-11 RPS Limiting Safety System Settings ...................................................................................... Section 2.13 TOC - Page 5 Amendment No. 116,125,1,2,160, 2-62

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.151,11 Instrumentation and Control Systems Applicability Applies to plant instrumentation systems.

Obiective To delineate the conditions of the plant instrumentation and control systems necessary to assure reactor safety.

Specifications The operability, permissible bypass, and Test Maintenance and Inoperable bypass specifications of the plant instrument and control systems shall be in accordance with Tables 2-2 through 2-5.

(1) In the event the number of channels of a particular system in service falls one below the total number of installed channels, the inoperable channel shall be placed in either the bypassed or tripped condition within one hour if the channel is equipped with a bypass switch, and eight hours if jumpers or blocks must be installed in the control circuitry. The inoperable channel may be bypassed for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from time of discovering loss of operability; however, if the inoperability is determined to be the result of malfunctioning RTDs or nuclear detectors supplying signals to the high power level, thermal margin/low pressurizer pressure, and axial power distribution channels, these channels may be bypassed for up to 7 days from time of discovering loss of operability. If the inoperable channel is not restored to OPERABLE status after the allowable time for bypass, it shall be placed in the tripped position or, in the case of malfunctioning RTDs or linear power nuclear detectors, the reactor shall be placed in hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If active maintenance and/or surveillance testing is being performed to return a channel to active service or to establish operability, the channel may be bypassed during the period of active maintenance and/or surveillance testing. This specification applies to the high rate trip-wide range log channel when the plant is at or above 10,4% power and is operating below 15% of rated power.

(2) In the event the number of channels of a particular system in service falls to the limits given in the column entitled "Minimum Operable Channels," one of the inoperable channels must be placed in the tripped position or low level actuation permissive position for the auxiliary feedwater system within one hour, if the channel is equipped with a bypass switch, and within eight hours if jumpers or blocks are required; however, if minimum operable channel conditions for SIRW tank low signal are reached, both inoperable channels must be placed in the bypassed condition within eight hours from time of discovery of loss of operability.

If at least one inoperable channel has not been restored to OPERABLE status after 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from time of discovering loss of operability, the reactor shall be placed in a hot shutdown condition within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; however, operation can continue without containment ventilation isolation signals available if the containment ventilation isolation valves are closed.

2.15 - Page 1 Amendment No. 8,20,54,65,88,109,494,209, 249T 2-63

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15'ATl Instrumentation and Control Systems (Continued)

If after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time of initiating a hot shutdown procedure at least one inoperable engineered safety features or isolation functions channel has not been restored to OPERABLE status, the reactor shall be placed in a cold shutdown condition within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This specification applied to the high rate trip-wide range log channel when the plant is at or above 10- % power and is operating below 15% of rated power.

(3) In the event the number of channels on a particular engineered safety features (ESF) or isolation logic subsystem in service falls below the limits given in the columns entitled "Minimum Operable Channels" or "Minimum Degree of Redundancy," except as conditioned by the column entitled "Permissible Bypass Conditions," sufficient channels shall be restored to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months so as to meet the minimum limits or the reactor shall be placed in a hot shutdown condition within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; however, operation can continue without containment ventilation isolation signals available if the ventilation isolation valves are closed. If after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time of initiating a hot shutdown procedure sufficient channels have not been restored to OPERABLE status, the reactor shall be placed in a cold shutdown condition within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(4) In the event the number of channels of those particular systems in service not described in (3) above falls below the limits given in the columns entitled "Minimum Operable Channels" or "Minimum Degree of Redundancy," except as conditioned by the column entitled "Permissible Bypass Conditions," the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If minimum conditions for engineered safety features or isolation functions are not met within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time of discovering loss of operability, the reactor shall be placed in a cold shutdown condition within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If the number of OPERABLE' high rate trip-wide range log channels falls below that given in the column entitled "Minimum Operable Channels" in Table 2-2 and the reactor is at or above 104% power and at or below 15% of rated power, reactor critical operation shall be discontinued and the plant placed in an operational mode allowing repair of the inoperable channels before startup or reactor critical operation may proceed.

If during power operation, the rod block function of the secondary CEA position indication system and rod block circuit are inoperable for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or the plant computer PDIL alarm, CEA group deviation alarm and the CEA sequencing function are inoperable for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the CEAs shall be withdrawn and maintained at fully withdrawn and the control rod drive system mode switch shall be maintained in the off position except when manual motion of CEA Group 4 is required to control axial power distribution.

(5) I* tho .....ont that the numbor of .p.rablo channo.. Of tho licted Altc.Rato Shutdown.

Pane eir the Auxiliary Fooidat op. P if.ano, intunatn or control circuits falls bolow the roguirod numbor of channoelc, oithor rostoro the roquirod numbor of cuh-annols to OPERABLE status within seven (7) days, or be in hot shutdown (Mode 3) within the next twolvo hours. This Spocificfation is,applicablo in Modos 1 and -2.

2.15- Page 2 Amendment No. 9, 20,54,65,99,125,157,494, 209, 2 TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control Systems (Continued)

,g-$:5.23Reactor Protective System (RPS)_Logic ald Tri-r nitiation FUnction/InstrUmont Rcguircd Numbor Or Conrol1 Parameqter Location of Canl

1. Reacti'~ity Control

a. Source Range Po)Wor Al-21:21

b. Reactor Wide Range 'AlI ý2121 Logarithmic Power

2. Reacto-r Coo-plant Systemn Pressure Centrel

a. Pressurizer Wide Ran*g .A1-179 1 Pressure (0-2500 psia)

3. De*ay Heat Rlemoval via Steam Ge*neators

a. ReactoFr Colant Hot Log A.-185 1 (Note- 1 T-empwatu-e

b. Reac*t*or C lant Cld Leg A- 185 1 (Note" TeRmperatwe G. Steam Generator + ro,.ureAl.179 1 per Stea Fn

eFate

d. Steam Generator Narrow Al-179 1 per Stea Fn Range Level i i Steam Generator Wide Al- 71 1 per Steam Range Level t~1( H I( Ji:II( Jr

4. Rleactor Coolant System

a. Press'urizer Level Al 185 1

b. Volume CGon*to* l Tank Level A-l185 1

c. Chargling PUMPG CH-1 B -and Al-185 1 its ass9ociatled controls6

d. Charging Isolation Valve Al-1851

5. Transfer Functions

a. All T-ransfer SWitchos,'Lockoult Al 1 84-591 Rays

b. All Transf*r SwitchesL*okIo*ut A 179 1 Relays Note 1: One reactor coolant hot leg temnperature indlication and one re-acto coolant cold leg temperature indication channel mus~t both be operable 91; the same steam generator(G . R Ao

,p-plica6biitv o heoeranoal;stausof R PS Logic and-TripIniiatonchannl in- MOD1E-S_ 1 an Nppes

)22; and,!

TECHNICAL SPECIFICATIONS When reactor coolant temperature (Tcod) is greater than 21 0°F or MODE 4 with more than one CEA rod capable of being withdrawn and RCS boron concentration less than REFUELING BORON CONCENTRATION.

Obiective To delineate the conditions of the plant instrumentation and control systems necessary to assure reactor safety.

Specification Six channels of RPS Logic matrices, four channels of RPS Trip Initiation Logic and two channels of RPS Manual Trip shall be OPERABLE.

Reguired Actions (1) With one RPS Logic Matrix channel inoperable, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

(2) With one RPS Trip Initiation Logic channel inoperable, de-energize the affected clutch power supply within one hour.

(3) With one RPS Manual Trip channel inoperable, restore the inoperable channel to OPERABLE status prior to entering MODE 2 from MODE 3.

(4) With two RPS Trip Initiation Logic channels affecting the same trip leg inoperable, de-energize the affected clutch power supplies immediately.

(5) With the required actions of (1), (2), or (4) not met, or with two RPS Manual Trip channels inoperable, or with two or more RPS Logic Matrices inoperable, or with two or more RPS Trip Initiation Logic channels inoperable for reasons other than (4):

a. be in HOT SHUTDOWN and verify no more than one CEA is capable of being withdrawn within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , OR

b. verify reactor coolant boron concentration is at REFUELING BORON CONCENTRATION within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

2.15 - Page 3 Amendment No.-2-0g, 249

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control Systems _(Continued)_

2 A3 Alternate Shutdown and Auxiliary Feedwater Panel Function/Instrument Required Number or Control Parameter Locra tion ofAChannel

6. Auxiliary Foodwator Conrols6

a. Steam Gonreator RC 2A and Al 1:79 1 2B Auxiliary Feedwater Irsolation Inboard and Outboard RI4l I I RA Valves Control

b. Stea~m-D-rivon Pump FW-10 Al-1:79- 1 Recirculation Valve Control G. Stoam Drivon Pump FW-10 Al-1791 S8toa1 IsolationR Valvo Control

d. Steam from Steamn Generator Al-179 1 RC 2A and RC 2B-- to M~AW10 Stoam Isolation Valve Control During plant operation, the comFplete insRGUmon8tation systems will normally boen'p~io This specificatfion outlines limniting codtosfor operation n9eesary to preserie the systemn when one Or moere of the channels are out of seryice. Reactor safety is provided by RPS, which automatically initiates appropriate_ a tin-t prevent exceeding established limi.5tsI. Saf ety is no9t coprmie, however, by contine d operation with Goei instrumentationR channels. out of srcsiepovinswere made forF this in the plant deeign The RPS and moest engineered safety feature c~hannels _are supplied with sufficin redundancy Ito,q provide *÷t ,*,' __I,--the *tio s capability f.,lh,,,*

, -,o for channel

,, test at power, eXcept for back(up channels suc-h as6 de~rived ciruit.i the ESP logic system.

W~hen onme of the four channels is taken out of sor~ico forF maintenance, RIPS logic can be changed to a two out Of three coincidence for a reactor trip by bypassing the removed ch-annel. If the bypass5 is6 not effec-ted, the out ofsrcGhannel (Power Removed) assumes a tripped condi~tion (xexcRept high rate o~f change of power, high power level and high pressrie pelso,~ which resul--ts, in a onRe out of-three1 chann~el logic. If in the-2-out-of- I logic system of the RPS one channel is bypassed and a second channel manually plac~ed in a tripped conRdition, the resulting lo~gic is 1 oeut-of -2. At rated power, the miium OPRERABLE high powchannell is 3 in ordeer te provide adequate power tilt doect8Gt*. Ifonly 2 channels arc OPERABLE, the reactorF power level isreduced to 70%

rated power which pro~tects the reactorF fromn possibly oxceeding design peaking factorsE due to undetected flux tilts and from exceeding dropped CEA peaking fatctrs.

Applies to the operational status of Alternate Shutdown and Auxiliary Feedwater Panel

!F~ctinsin MODES 1and 2.i d~lelineate h coniton offthe plant instrumentation' and cent rol systems necessaryto,

TECHNICAL SPECIFICATIONS S&6fi q. AiFTh u r r Po~iL~aa~ier~n Tble2-@9shall b~e OPERABLEJ

~j7With-the nnuIMbe r Af 5P RA,4kTLE~ccha nnýs oF6hr confiro.1Cl.uit s r~~~nth re"ired nmer ef chanmels retý,eýtheJr~equw~e dQnierf q L v~h ywen ()Ly§;_,hest.'PR Vý -iih Wh iFtfD 2.15 - Page 4 Amendment No. 88,125,152,173,191, 208, 2-49

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control Systems Basis During plant operation, the complete instrumentation systems will normally be in service.

This specification outlines limiting conditions for operation necessary to preserve the effectiveness of the reactor protective system (RPS) and engineered safety features (ESF) system when one or more of the channels are out of service. Reactor safety is provided by RPS, which automatically initiates appropriate action to prevent exceeding established limits. Safety is not compromised, however, by continued operation with certain instrumentation channels out of service since provisions were made for this in the plant design.

The RPS and most engineered safety feature channels are supplied with sufficient redundancy to provide the capability for channel test at power, except for backup channels such as derived circuits in the ESF logic system.

When one of the four channels is taken out of service for maintenance, RPS logic can be changed to a two-out-of-three coincidence for a reactor trip by bypassing the removed channel. Ifthe bypass is not effected, the out-of-service channel (Power Removed) assumes a tripped condition (except high rate-of-change of power, high power level and high pressurizer pressure),(') which results in a one-out-of-three channel logic. If in the 2-out-of-4 logic system of the RPS one channel is bypassed and a second channel manually placed in a tripped condition, the resulting logic is 1-out-of-2. At rated power, the minimum OPERABLE high-power level channel is 3 in order to provide adequate power tilt detection. Ifonly 2 channels are OPERABLE, the reactor power level is reduced to 70%

rated power which protects the reactor from possibly exceeding design peaking factors due to undetected flux tilts and from exceeding dropped CEA peaking factors.

An RPS Logic matrix channel consists of two matrix power supplies, four matrix relays and their associated contacts as well as all interconnecting wiring. An RPS Trip Initiation Logic channel consists of an M contactor and associated contacts, an interposing relay and all interconnecting wiring. Two RPS Trip Initiation Logic channels associated with the same pair of CEDM clutch power supplies are considered to affect the same trip leg.

Integrated into the trip initiation logic are two RPS Manual Trip channels. Manual Trip #1 operates by directly de-energizing all four M contactors in response to the operation of a manual pushbutton. Manual Trip #2 operates by de-energizing an undervoltage relay which results in the opening of two circuit breakers, CB-AB and CB-CD, which supply power to the CEDM clutch power supplies. Manual Trip channel #1 consists of manual trip pushbutton #1 and interconnecting wiring. Manual Trip channel #2 consists of manual trip pushbutton #2, circuit breakers CB-AB and CB-CD, and associated interconnecting wiring.

With one manual reactor trip channel inoperable, operation may continue until the reactor is shut down for other reasons. No safety analyses assume operation of the Manual trip.

Because of this, the Required Action is to restore the inoperable channel to OPERABLE status prior to entering MODE 2 from MODE 3 during the next plant startup.

2.15 - Page 45 Amendment No. 98,125,152,173,191, 208, 249 TSBC

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control Systems (Continued)

Basis (Continued)

The ESF logic system is a Class 1 protection system designed to satisfy the criteria of IEEE 279, August 1968. Two functionally redundant ESF logic subsystems "A" and "B" are provided to ensure high reliability and effective in-service testing. These logic subsystems are designed for individual reliability and maximum attainable mutual independence both physically and electrically. Either logic subsystem acting alone can automatically actuate engineered safety features and essential supporting systems.

All Engineered Safety Features are initiated by 2-out-of-4 logic matrices except containment high radiation which operates on a 1-out-of-2 basis. The number of installed channels for Containment Radiation High Signal (CRHS) is two. CRHS isolates the containment pressure relief, air sample and purge system valves.

Entry into Technical Specification 2.15DI(3) is made when conditions have caused one logic subsystem ("A" or "B") to become inoperable but the redundant logic subsystem remains operable. The loss of a prime initiation relay (which renders all 4 channels of a logic subsystem inoperable) is the condition most likely to cause entry into Technical Specification 2.15j1(3). In this situation, the remaining ESF logic subsystem still has the capability to automatically actuate engineered safety features equipment and essential supporting systems. The 48-hour completion time is commensurate with the importance of avoiding the vulnerability of a single failure in the remaining ESF logic subsystem.

Technical Specification 2.15,1i(3) will not be used upon loss of the common channels that affect both "A" and "B" subsystems prime initiators operability unless the permissible bypass condition is met. Upon exiting TS 2.155.(3) following the restoration of a prime initiation relay to OPERABLE status, if any channel(s) remain inoperable, the appropriate Limiting Conditions for Operation (LCO) (TS 2.15j,(1) orT.J(2)is applicable with the length of inoperability measured from time of discovery of: 1) prime initiation relay inoperable, or 2) channel inoperability, whichever is longer.

The ESF system provides a 2-out-of-4 logic on the signals used to actuate the equipment connected to each of the two emergency diesel generator units.

The rod block system automatically inhibits all CEA motion in the event a LCO on CEA insertion, CEA deviation, CEA overlap or CEA sequencing is approached. The installation of the rod block system ensures that no single failure in the control element drive control system (other than a dropped CEA) can cause the CEAs to move such that the CEA insertion, deviation, sequencing or overlap limits are exceeded. Accordingly, with the rod block system installed, only the dropped CEA event is considered an Anticipated Operational Occurrence (AOO) and factored into the derivation of the Limiting Safety System Settings (LSSS) and LCO. With the rod block function out-of-service, several additional CEA deviation events must be considered as AOOs. Analysis of these incidents indicates that the single CEA withdrawal incident is the most limiting of these events. An analysis of the at-power single CEA withdrawal incident was performed for Fort Calhoun for various initial Group 4 insertions, and it has been concluded that the LCO and LSSS are.

valid for a Group 4 insertion of less than or equal to 15%.

2.15 - Page 5, Amendment No. 21-94,20, 249 TSBC 06-001-0

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control Systems (Continued)

Basis (Continued)

Operability of the primary CEA position indication system (CEAPIS) channel and the secondary CEAPIS channel is required to determine CEA positions and thereby ensure compliance with the CEA alignment and insertion limits of TS 2.10.2. The primary CEAPIS channel utilizes the output of a synchro transmitter geared to the clutch output shaft. CEA position is displayed visually at the main control panel.

The secondary CEAPIS channel utilizes the output of a voltage divider network controlled by a series of reed switches. The reed switches are actuated by a permanent magnet attached to the rack assembly. Position information is supplied to the distributed control system (DCS) flat-panel touch monitors for simultaneous viewing of all CEA group positions.

Limit switches on the regulating CEAs and reed switches on the shutdown CEAs provide an additional means of determining CEA position when the CEAs are fully inserted or fully withdrawn. When the CEAs are fully inserted or fully withdrawn, this indication (displayed on the DCS) can be used in lieu of secondary CEAPIS data to meet the shiftly CHANNEL CHECK of primary CEAPIS. However, as limit switch indication is not fully independent of secondary CEAPIS, primary CEAPIS must be used to verify secondary CEAPIS data.

In MODES 1 and 2, CEA position indication is required to allow verification that the CEAs are positioned and aligned as assumed in the safety analysis. If one CEA position indication channel is inoperable for one or more CEAs, TS 3.1, Table 3-3, Item 4 (CEA position verification) must be performed within 15 minutes following any CEA motion in that group to ensure that the CEAs are positioned as required.

The operability of the Alternate Shutdown Panel (Al-1 85), including Wide Range Logarithmic Power and Source Range Monitors on AI-212, and Emergency Auxiliary Feedwater Panel (AI-179) instrument and control circuits ensures that sufficient capability is available to permit entry into and maintenance of the Hot Shutdown Mode from locations outside of the Control Room. This capability is required in the event that Control Room habitability is lost due to fire in the cable spreading room or Control Room.

Variances which may exist at startup between the more accurate AT-Power and Nuclear Instrumentation Power (NI-Power) are not significant for enabling of the trip functions. By 15% of rated power as measured by the uncalibrated NI Power, the Axial Power Distribution (APD) and Loss of Load (LOL) trip functions are enabled while the High Rate of Change of Power trip is bypassed.

The APD trip function acts to limit the axial power shape to the range assumed in the setpoint analysis. Significant margins to local power density limits exist at 15% power, as well as power levels up to at least 30% (where NI calibration occurs).

2.15 - Page 67Z Amendment No. 208, 249 TSBC-04-001 -0 TSBC-08-003-0 TSBC-08-008-0 TSBC-1 1-006-0

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control Systems (Continued)

Basis (Continued)

The LOL trip function acts as an anticipatory trip for the high pressurizer pressure and high power trips in order to limit the severity of a LOL transient. This trip is not credited in the USAR Chapter 14 Safety Analyses and any variance between AT-Power and NI-Power has no effect on the safety analysis.

The High Rate of Change of Power trip acts to limit power excursions from low power levels and bypassing of this trip at a high power level is conservative. This trip is not credited in the USAR Chapter 14 Safety Analyses for Mode 1 operation. Any variance between AT-Power and NI-Power has no effect on the safety analysis.

Steam generator blowdown isolation ensures that the auxiliary feedwater system performs its design function of maintaining adequate steam generator (SG) water level for decay heat removal once the auxiliary feedwater actuation signal (AFAS) is actuated. The steam generator blowdown isolation function consists of two trains (logic subsystems). Each train closes one SG blowdown isolation valve to each SG. Each SG has redundant (Train A and Train B) blowdown isolation valves. Four clutch power relays initiate closure of the SG blowdown isolation valves with each clutch power relay closing one valve when the reactor trips. Failure of one clutch power relay to initiate SG blowdown isolation or failure of one train will not prevent single valve isolation of SG blowdown flow.

References (1) USAR, Section 7.2.7.1 2.15- Page 641 Amendment No. 208, 249 TSBC-08-008-0 TSBC-09-009-0

TECHNICAL SPECIFICATIONS TABLE 2-2 Instrument Operating Requirements for Reactor Protective System Test, Maintenance Minimum Minimum Permissible and Functional Operable Degree of Bypass Inoperable No. Unit Channels Redundancy Condition Bypass Li_ Not Used 2 High Power Level 1(c) Thermal Power (e) 2 (b)(c)

Input Bypassed below 10-4% of Rated Power(a)(d) 3 Thermal Margin/Low 1 Below 104% of (e) 2 (b)

Pressurizer Pressure Rated Power(a)(d) 4 High Pressurizer 1 None (e) 2 (b)

Pressure 5 Low R.C. Flow 1 Below 104% of (e) 2 (b)

Rated Power (a)(d) 6 Low Steam Generator 2/Steam 1/Steam None (e)

Water Level Gen(b) Gen 7 Low Steam Generator 2/Steam 1/Steam Below 600 (e)

Pressure Gen(b) Gen psia(a)(d) 8 Containment High 2 (b) 1 During Leak Test (e)

Pressure 9 Axial Power 1(C) Below 15% of (e)

Distribution Rated Power(g) 10 High Rate Trip-wide 2 (b) Below 104% and (e)

Range Log Channels above 15% of Rated Power(a)(g) 11 Loss of Load 2 (b) Below 15% of (e)

Rated Power(g) 12 Steam Generator 2 (b) None (e)

Differential Pressure

a. Bypass automatically removed.

b. Specification 2.15 r"(2) is applicable.

2.15 - Page 79 Amendment No. 60,77,88,153,194, 24-

TECHNICAL SPECIFICATIONS TABLE 2-2 (Continued)

c. If two channels are inoperable, load shall be reduced to 70% or less of rated power.

d. For low power physics testing this trip may be bypassed up to 101% of rated power.

e. Specification 2.15ýi(1) is applicable.

f. Deleted.

g. For each channel, the same bistable automatically activates the Loss of Load and Axial Power Distribution (APD) trips and automatically bypasses the high rate trip at 15% of rated power. Only the APD trip is a Limiting Safety System Setting. Therefore, the bistable is set to actuate within the APD tolerance band.

2.15 - Page BE10 Amendment No. 60,77,88,194, 249

TECHNICAL SPECIFICATIONS TABLE 2-3 Instrument Oneratina Reauirements for Enaineered Safety Features Test, Minimum Minimum Permissible Maintenance Functional Operable Degree of Bypass and Inoperable No. Unit Channels Redundancy Condition Bypass 1 Safety Injection A Manual 1 None None N/A B High Containment Pressure Logic Subsystem A 2 (a)(d)(I) 1 During Leak (f)

Logic Subsystem B 1 Test 2 (a)(d)O)

C Pressurizer Low/Low Pressure Logic Subsystem A 2 (a)(d)(1) 1 Reactor Coolant (f)

Logic Subsystem B 2(a)(d)(1) 1 Pressure Less Than 1700 psia(b) 2 Containment Spray A Manual(m) 1 None None N/A B High Containment Pressure Logic Subsystem A 1 During Leak 2 (a)(c)(d)(I) 1 (f)

Logic Subsystem B 2 (a)(c)(d)(I) Test C Pressurizer Low/Low Pressure Logic Subsystem A 2 (a)(c)(d)(I) 1 Reactor Coolant (f)

Logic Subsystem B 2 (a)(c)(d)(J) 1 Pressure Less Than 1700 psia(b)

D Steam Generator Low/Low Pressure Logic Subsystem A 2/Steam 1/Steam Steam Generator (f)

Gen (a)(c)(d)(1)

Gen Pressure Less Than Logic Subsystem B 2/Steam 1/Steam 600 psia(n)

Gen(a)(c)(d)(I) Gen 3 Recirculation A Manual 1 None None N/A B SIRW Tank Low Level Logic Subsystem A 2 (a)(k)(I) 1 None (J)

Logic Subsystem B (a)(k)(I) 1 2

4 Emerqency Off-Site Power Trip A Manual 1 (e) None None N/A B Emergency Bus Low Voltage (Each Bus)

-Loss of Voltage 2 (d) 1 Reactor Coolant (f) 2 (a)(d) 1 Temperature Less

-Degraded Voltage Than 300' F 2.15- Page g*l Amendment No. 41,65,88,173,184,19,249, 266

TECHNICAL SPECIFICATIONS TABLE 2-3 (Continued)

Test, Minimum Minimum Permissible Maintenance Functional Operable Degree of Bypass and Inoperable No. Unit Channels Redundancy Condition Bypass 5 Auxiliary Feedwater A Manual 1 None None N/A B Auto. Initiation Logic Subsystem A Operating Modes Logic Subsystem B 3, 4, and 5

-Steam Generator 1 (h)

Low Level

-Steam Generator 3 (a)(g)(I) 1 (i)

Low Pressure

-Steam Generator 3 (a)(g)(I) 1 (i)

Differential Pressure a Circuits on ESF Logic Subsystems A and B each have 4 channels.

b Auto removal of bypass above 1700 psia.

c Coincident containment high pressure, pressurizer low/low pressure, and steam generator low pressure signals are required for initiation of containment spray.

d If minimum OPERABLE channel conditions are reached, one inoperable channel must be placed in the tripped condition or low level actuation position for auxiliary feedwater system within eight hours from the time of discovery of loss of operability. Specification 2.151 (2) is applicable.

e Control switch on incoming breaker.

f If one channel becomes inoperable, that channel must be placed in the tripped or bypassed condition within eight hours from time of discovery of loss of operability. Specification 2.15*j(1) is applicable.

g Three channels required because bypass or failure results in auxiliary feedwater actuation block in the affected channel.

h Specification 2.15,1(1) is applicable.

2.15 - Page410191 Amendment No. 65,98,194,249, 25

TECHNICAL SPECIFICATIONS TABLE 2-3 (Continued)

If the channel becomes inoperable, that channel must be placed in the bypassed condition within eight hours from time of discovery of loss of operability. If the channel is not returned to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from time of discovery of loss of operability, one of the eight channels may continue to be placed in the bypassed condition provided the Plant Review Committee has reviewed and documented the judgment concerning prolonged operation in bypass of the inoperable channel. The channel shall be returned to OPERABLE status no later than during the next cold shutdown. If one of the four channels on one steam generator is in prolonged bypass and a channel on the other steam generator becomes inoperable, the second inoperable channel must be placed in bypass within eight hours from time of discovery of loss of operability. If one of the inoperable channels is not returned to OPERABLE status within seven days from the time of discovery of the second loss of operability, the unit must be placed in hot shutdown within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

If one channel becomes inoperable, that channel must be placed in the bypassed condition within eight hours from time of discovery of loss of operability. If the channel is not returned to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from time of discovery of loss of operability, one of the eight channels may continue to be placed in the bypassed condition provided the Plant Review Committee has reviewed and documented the judgment concerning prolonged operation in bypass of the inoperable channel. The channel shall be returned to OPERABLE status no later than during the next cold shutdown. If a channel is in prolonged bypass and a channel on the opposite train becomes inoperable, the second inoperable channel must be placed in bypass within eight hours from time of discovery of loss of operability. If one of the inoperable channels is not returned to OPERABLE status within seven days from the time of discovery of the second loss of operability, the unit must be placed in hot shutdown within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

k Specification 2.15ý:1](2) is applicable.

Specification 2.15'1,(3) is applicable. If ESF Logic Subsystems A and B are inoperable, enter Specification 2.0.1.

m Steam Generator Low Pressure permissive is required for actuation.

n Auto removal of bypass prior to exceeding 600 psia.

2.15 - Page 1-o Amendment No. 88,173,494,,249, 25

TECHNICAL SPECIFICATIONS TABLE 2-4 Instrument Operating Conditions for Isolation Functions Test, Minimum Minimum Permissible Maintenance Functional Operable Degree of Bypass and Inoperable No. Unit Channels Redundancy Condition Bypass 1 Containment Isolation A Manual 1 None None N/A B Containment High Pressure Logic Subsystem A 2 (a)(e)(g) 1 During Leak (f)

Logic Subsystem B 2(a)(e)(g) 1 Test C Pressurizer Low/Low Pressure Logic Subsystem A 2 (a)(e)(g) 1 Reactor Coolant (f)

Logic Subsystem B 2(a)(e)(g) 1 Pressure Less Than 1700 psia(b) 2 Steam Generator Isolation A Manual 1 None None N/A B Steam Generator Isolation 1 None None N/A (i) Steam Generator

,Low Pressure Logic Subsystem A 2/Steam 1/Steam Steam Generator (f)

Gen2(a)(m)() Gen Pressure Less Than 600 psia(c)

Logic Subsystem B 2/Steam l/Steam Gen (a)(e)(g) Gen (ii) Containment High Pressure Logic Subsystem A 2(a)(e)(g) 1 During Leak (f)

Logic Subsystem B 2(a)(e)(g) 1 Test 3 - Ventilation Isolation A Manual 1 None None N/A B Containment High Radiation Logic Subsystem A 1 (d)(g) None If Containment (f)

Logic Subsystem B l(d)(g) None Relief and Purge Valves are Closed 4 Steam Generator Blowdown Isolation A Manual 1 (h) None Operating Modes N/A 3,4, & 5 B Reactor Trip 2 (h)(i) None Operating Modes (j)

Trains A and B 3, 4, & 5 OR ifat least one valve for each steam generator is closed 2.15 - Page 2-4 Amendment No. 88,93,108,152,153,173,184,194,249, 26&, 2-43

TECHNICAL SPECIFICATIONS TABLE 2-4 (Continued) a Circuits on ESF Logic Subsystems A and B each have 4 channels.

b Auto removal of bypass prior to exceeding 1700 psia.

c Auto removal of bypass prior to exceeding 600 psia.

d A and B trains are both actuated by either the Containment or Auxiliary Building Exhaust Stack initiating channels. The number of installed channels for Containment Radiation High Signal is two for purposes of Specification 2.15ýjl(1).

e If minimum operable channel conditions are reached, one inoperable channel must be placed in the tripped condition within eight hours from the time of discovery of loss of operability. Specification 2.15'1 (2) is applicable.

f If one channel becomes inoperable, that channel must be placed in the tripped or bypassed condition within eight hours from the time of discovery of loss of operability. Specification 2.1521(1) is applicable.

g Specification 2.151-j(3) is applicable. If ESF Logic Subsystems A and B are inoperable, enter Specification 2.0.1.

h. "Minimum Operable Channels" for steam generator blowdown isolation refers to the minimum number of trains (logic subsystems) which are required to be operable to provide manual or automatic SG blowdown isolation.

If both trains become inoperable, power operation may continue provided at least one SG blowdown isolation valve for each steam generator is closed OR be in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 3 in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Specifications 2.153.(1), (2), (3) and (4) are not applicable; TS LCO 2.0.1 is not applicable.

j. If one train becomes inoperable, that train may be placed in the bypassed condition. If the train is not returned to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time of discovery of loss of operability, operation may continue as long as one SG blowdown isolation valve to each steam generator is closed. If the train is not returned to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time of discovery, with blowdown not isolated to both SGs, be in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 3 in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Specifications 2.15iFj(1), (2), (3) and (4) are not applicable; TS LCO 2.0.1 is not applicable.

2.15 - Page Amendment No. 88,108,152,173,194,249, 263

TECHNICAL SPECIFICATIONS TABLE 2-5 Instrumentation Operatinq Requirements for Other Safety Feature Functions Minimum Minimum Permissible Functional Operable Degree of Bypass No. Unit Channels Redundancy Condition 1 CEA Position Indication 1 (a) None None Systems 2 Pressurizer Level 1 None Not Applicable NOTES:

(a) If one channel of CEA position indication is inoperable for one or more CEAs, requirements of specification 2.15C1 are modified for item 1 to "Perform TS 3.1, Table 3-3. Item 4 within 15 minutes following any CEA motion in that group." Specifications 2.15D1(1), (2), and (3) are not applicable.

2.15 - Page 1-4LT-6 Amendment No. -54,65,110,249, 265, 267

TECHNICAL SPECIFICATIONS TABLE 2-6 Alternate Shutdown and Auxiliary Feedwater Panel Functions Function/Instrument Required Number or Control Parameter Location of Channels

1. Reactivity Control

a. Source Range Power AI-212 1

b. Reactor Wide Range AI-212 1 Logarithmic Power

2. Reactor Coolant System Pressure Control

a. Pressurizer Wide Range Al-179 Pressure (0-2500 psia)

3. Decay Heat Removal via Steam Generators

a. Reactor Coolant Hot Leg Al-185 1 (Note 1)

Temperature

b. Reactor Coolant Cold Leg A-i 85 1 (Note 1)

Temperature

c. Steam Generator Pressure Al-1 79 1 per Steam Generator

d. Steam Generator Narrow A1-179 1 per Steam Range Level Generator

e. Steam Generator Wide Al-179 1 per Steam Range Level Generator

4. Reactor Coolant System Inventory Controls

a. Pressurizer Level Al-185 1

b. Volume Control Tank Level A1-185 1

c. Charging Pump CH-1 B and Al-1 85 1 its associated controls

d. Charging Isolation Valve Al-185 Control 1

5. Transfer Functions

a. All Transfer Switches/Lockout Al-1 85 Relays

b. All Transfer Switches/Lockout A1-179 Relays Note 1: One reactor coolant hot leg temperature indication and one reactor coolant cold leg temperature indication channel must both be operable on the same steam generator (i.e.,

RC-2A or RC-2B).

2.15 - Page 17 Amendment No.

TECHNICAL SPECIFICATIONS TABLE 2-6 (Continued)

Alternate Shutdown and Auxiliary Feedwater Panel Functions Function/Instrument Required Number or Control Parameter Location of Channels

6. Auxiliary Feedwater Controls

a. Steam Generator RC-2A and Al-1 79 1 2B Auxiliary Feedwater Isolation Inboard and Outboard Valves Control

b. Steam-Driven Pump FW-10 AI-179 1 Recirculation Valve Control

c. Steam-Driven Pump FW-10 Al-1 79 1 Steam Isolation Valve Control

d. Steam from Steam Generator Al-1 79 1 RC-2A and RC-2B to FW-1 0 Steam Isolation Valve Control 2.15 - Page 18 Amendment No.

LIC-12-0006 Enclosure, Attachment 2 Page 1 Retyped ("Clean")

Technical Specifications and Information Only Bases Pages

TECHNICAL SPECIFICATION TABLE OF CONTENTS (Continued) 2.13 Limiting Safety System Settings, Reactor Protective System 2.14 Engineered Safety Features System Initiation Instrumentation Settings 2.15.1 Instrumentation and Control Systems 2.15.2 Reactor Protective System (RPS) Logic and Trip Initiation 2.15.3 Alternate Shutdown and Auxiliary Feedwater Panel 2.16 River Level 2.17 Miscellaneous Radioactive Material Sources 2.18 DELETED 2.19 DELETED 2.20 Steam Generator Coolant Radioactivity 2.21 Post-Accident Monitoring Instrumentation 2.22 DELETED 2.23 Steam Generator (SG) Tube Integrity 3.0 SURVEILLANCE REQUIREMENTS 3.1 Instrumentation and Control 3.2 Equipment and Sampling Tests 3.3 Reactor Coolant System and Other Components Subject to ASME XI Boiler and Pressure Vessel Code Inspection and Testing Surveillance 3.4 DELETED 3.5 Containment Test 3.6 Safety Injection and Containment Cooling Systems Tests 3.7 Emergency Power System Periodic Tests 3.8 Main Steam Isolation Valves 3.9 Auxiliary Feedwater System 3.10 Reactor Core Parameters 3.11 DELETED 3.12 Radioactive Waste Disposal System 3.13 Radioactive Material Sources Surveillance 3.14 DELETED 3.15 DELETED 3.16 Residual Heat Removal System Integrity Testing 3.17 Steam Generator (SG) Tube Integrity 4.0 DESIGN FEATURES 4.1 Site 4.2 Reactor Core 4.3 Fuel Storage TOC - Page 2 Amendment No. 11,27,32,38,13,16,51,60,8,*86, 93,97,104,4122,4136,152, 160,17-6,183, 211,230, 236, 246,218, 2.52

TECHNICAL SPECIFICATION TECHNICAL SPECIFICATIONS - TABLES TABLE OF CONTENTS TABLE SECTION 2-1 ESFS Initiation Instrumentation Setting Limits ......................................................................... Section 2.14 2-2 Instrument Operating Requirements for RPS ....................................................................... Section 2.15.1 2-3 Instrument Operating Requirements for Engineered Safety Features ................................. Section 2.15.1 2-4 Instrument Operating Conditions for Isolation Functions ...................................................... Section 2.15.1 2-5 Instrumentation Operating Requirements for Other Safety Feature Functions .................... Section 2.15.1 2-6 Alternate Shutdown and Auxiliary Feedwater Panel Functions ........................................... Section 2.15.3 2-9 RCS Pressure Isolation Valves .................................................................................................. Section 2.1 2-10 Post-Accident Monitoring, Instrumentation Operating Limits ................................................... Section 2.21 2-11 RPS Limiting Safety System Settings ...................................................................................... Section 2.13 3-1 Minimum Frequencies for Checks, Calibrations, and Testing of RPS ....................................... Section 3.1 3-2 Minimum Frequencies for Checks, Calibrations and Testing of ............................................... Section 3.1 Engineered Safety Features, Instrumentation and Controls 3-3 Minimum Frequencies for Checks, Calibrations, and Testing ................................................... Section 3.1 of Miscellaneous Instrumentation and Controls 3-3a Minimum Frequency for Checks, Calibrations and Functional .................................................. Section 3.1 Testing of Alternate Shutdown Panels (AI-185 and AI-212) and Emergency Auxiliary Feedwater Panel (Al-1 79) Instrumentation and Control Circuits 3-4 Minimum Frequencies for Sampling Tests ................................................................................. Section 3.2 3-5 Minimum Frequencies for Equipment Tests .............................................................................. Section 3.2 3-6 Reactor Coolant Pump Surveillance .......................................................................................... Section 3.3 5.2-1 Minimum Shift Crew Composition .............................................................................................. SectiOn 5.0 TOC - Page 4 Amendment No. 116,136,6G0, 246,248, 25

TECHNICAL SPECIFICATION TECHNICAL SPECIFICATIONS - TABLES TABLE OF CONTENTS (ALPHABETICAL ORDER)

TABLE DESCRIPTION SECTION 2-6 Alternate Shutdown and Auxiliary Feedwater Panel Functions ........................................... Section 2.15.3 2-1 ESFS Initiation Instrumentation Setting Limits ......................................................................... Section 2.14 2-4 Instrument Operating Conditions for Isolation Functions ...................................................... Section 2.15.1 2-2 Instrument Operating Requirements for RPS ....................................................................... Section 2.15.1 2-3 Instrument Opdrating Requirements for Engineered Safety Features ................................. Section 2.15.1 2-5 Instrumentation Operating Requirements for Other Safety ................................................. Section 2.15.1 Features Functions 3-3a Minimum Frequency for Checks, Calibrations and Functional .................................................. Section 3.1 Testing of Alternate Shutdown Panels (AI-185 and AI-212) and Emergency Auxiliary Feedwater Panel (Al-1 79) Instrumentation and Control Circuits 3-2 Minimum Frequencies for Checks, Calibrations and Testing of ............................................... Section 3.1 Engineered Safety Features, Instrumentation and Controls 3-3 Minimum Frequencies for Checks, Calibrations, and Testing .................................................. Section 3.1 of Miscellaneous Instrumentation and Controls 3-1 Minimum Frequencies for Checks, Calibrations ....................................................................... Section 3.1 and Testing of RPS 3-5 Minimum Frequencies for Equipment Tests .............................................................................. Section 3.2 3-4 Minimum Frequencies for Sampling Tests ................................................................................. Section 3.2 5.2-1 Minimum Shift Crew Composition .............................................................................................. Section 5.0 2-10 Post-Accident Monitoring Instrumentation Operating Limits ................................................... Section 2.21 2-9 RCS Pressure Isolation Valves .................................................................................................. Section 2.1 3-6 Reactor Coolant Pump Surveillance .......................................................................................... Section 3.3 2-11 RPS Limiting Safety System Settings ...................................................................................... Section 2.13 TOC - Page 5 Amendment No. 116,125,* 42,41-6, 262

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15.1 Instrumentation and Control Systems Applicability Applies to plant instrumentation systems.

Obiective To delineate the conditions of the plant instrumentation and control systems necessary to assure reactor safety.

Specifications The operability, permissible bypass, and Test Maintenance and Inoperable bypass specifications of the plant instrument and control systems shall be in accordance with Tables 2-2 through 2-5.

(1) In the event the number of channels of a particular system in service falls one below the total number of installed channels, the inoperable channel shall be placed in either the bypassed or tripped condition within one hour if the channel is equipped with a bypass switch, and eight hours if jumpers or blocks must be installed in the control circuitry. The inoperable channel may be bypassed for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from time of discovering loss of operability; however, if the inoperability is determined to be the result of malfunctioning RTDs or nuclear detectors supplying signals to the high power level, thermal margin/low pressurizer pressure, and axial power distribution channels, these channels may be bypassed for up to 7 days from time of discovering loss of operability. If the inoperable channel is not restored to OPERABLE status after the allowable time for bypass, it shall be placed in the tripped position or, in the case of malfunctioning RTDs or linear power nuclear detectors, the reactor shall be placed in hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If active maintenance and/or surveillance testing is being performed to return a channel to active service or to establish operability, the channel may be bypassed during the period of active maintenance and/or surveillance testing. This specification applies to the high rate trip-wide range log channel when the plant is at or above 10-4% power and is operating below 15% of rated power.

(2) In the event the number of channels of a particular system in service falls to the limits given in the column entitled "Minimum Operable Channels," one of the inoperable channels must be placed in the tripped position or low level actuation permissive position for the auxiliary feedwater system within one hour, if the channel is equipped with a bypass switch, and within eight hours if jumpers or blocks are required; however, if minimum operable channel conditions for SIRW tank low signal are reached, both inoperable channels must be placed in the bypassed condition within eight hours from time of discovery of loss of operability.

If at least one inoperable channel has not been restored to OPERABLE status after 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from time of discovering loss of operability, the reactor shall be placed in a hot shutdown condition within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; however, operation can continue without containment ventilation isolation signals available if the containment ventilation isolation valves are closed.

2.15 - Page 1 Amendment No. 8,20,51,,85,,08,1"1,208, 2-4-, 2-63

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15.1 Instrumentation and Control Systems (Continued)

If after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time of initiating a hot shutdown procedure at least one inoperable engineered safety features or isolation functions channel has not been restored to OPERABLE status, the reactor shall be placed in a cold shutdown condition within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This specification applied to the high rate trip-wide range log channel when the plant is at or above 10-% power and is operating below 15% of rated power.

(3) In the event the number of channels on a particular engineered safety features (ESF) or isolation logic subsystem in service falls below the limits given in the columns entitled "Minimum Operable Channels" or "Minimum Degree of Redundancy," except as conditioned by the column entitled "Permissible Bypass Conditions," sufficient channels shall be restored to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months so as to meet the minimum limits or the reactor shall be placed in a hot shutdown condition within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; however, operation can continue without containment ventilation isolation signals available if the ventilation isolation valves are closed. If after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time of initiating a hot shutdown procedure sufficient channels have not been restored to OPERABLE status, the reactor shall be placed in a cold shutdown condition within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(4) In the event the number of channels of those particular systems in service not described in (3) above falls below the limits given in the columns entitled "Minimum Operable Channels" or "Minimum Degree of Redundancy," except as conditioned by the column entitled "Permissible Bypass Conditions," the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . If minimum conditions for engineered safety features or isolation functions are not met within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time of discovering loss of operability, the reactor shall be placed in a cold shutdown condition within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If the number of OPERABLE high rate trip-wide range log channels falls below that given in the column entitled "Minimum Operable Channels" in Table 2-2 and the reactor is at or above 10-4% power and at or below 15% of rated power, reactor critical operation shall be discontinued and the plant placed in an operational mode allowing repair of the inoperable channels before startup or reactor critical operation may proceed.

If during power operation, the rod block function of the secondary CEA position indication system and rod block circuit are inoperable for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or the plant computer PDIL alarm, CEA group deviation alarm and the CEA sequencing function are inoperable for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the CEAs shall be withdrawn and maintained at fully withdrawn and the control rod drive system mode switch shall be maintained in the off position except when manual motion of CEA Group 4 is required to control axial power distribution.

2.15 - Page 2 Amendment No. 9, 20,54,65,99,125,157,194, 208, 24-

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control Systems 2.15.2 Reactor Protective System (RPS) Logic and Trip Initiation Appl~icability Applies to the operational status of RPS Logic and Trip Initiation channels in MODES 1 and 2; and, When reactor coolant temperature (TcoId) is greater than 21 0°F or MODE 4 with more than one CEA capable of being withdrawn and RCS boron concentration less than REFUELING BORON CONCENTRATION.

Obmective To delineate the conditions of the plant instrumentation and control systems necessary to assure reactor safety.

Specification Six channels of RPS Logic matrices, four channels of RPS Trip Initiation Logic and two channels of RPS Manual Trip shall be OPERABLE.

Required Actions (1) With one RPS Logic Matrix channel inoperable, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

(2) With one RPS Trip Initiation Logic channel inoperable, de-energize the affected clutch power supply within one hour.

(3) With one RPS Manual Trip channel inoperable, restore the inoperable channel to OPERABLE status prior to entering MODE 2 from MODE 3.

(4) With two RPS Trip Initiation Logic channels affecting the same trip leg inoperable, de-energize the affected clutch power supplies immediately.

(5) With the required actions of (1), (2), or (4) not met, or with two RPS Manual Trip channels inoperable, or with two or more RPS Logic Matrices inoperable, or with two or more RPS Trip Initiation Logic channels inoperable for reasons other than (4):

a. be in HOT SHUTDOWN and verify no more than one CEA is capable of being withdrawn within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , OR

b. verify reactor coolant boron concentration is at REFUELING BORON CONCENTRATION within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

2.15 - Page 3 Amendment No.-208, 249

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control Systems 2.15.3 Alternate Shutdown and Auxiliary Feedwater Panel Applicability Applies to the operational status of Alternate Shutdown and Auxiliary Feedwater Panel Functions in MODES 1 and 2.

Objective To delineate the conditions of the plant instrumentation and control systems necessary to assure reactor safety.

Specification The Alternate Shutdown and Auxiliary Feedwater Panel Functions/Instrumentation or Control Parameters in Table 2-6 shall be OPERABLE.

Required Actions (1) With the number of OPERABLE channels or control circuits less than the required number of channels, restore the required number of channels to OPERABLE within seven (7) days.

(2) With the required actions of (1) not met, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.15 - Page 4 Amendment No. 88,125,152,473,194, 208, 249

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control Systems (Continued)

Basis During plant operation, the complete instrumentation systems will normally be in service.

This specification outlines limiting conditions for operation necessary to preserve the effectiveness of the reactor protective system (RPS) and engineered safety features (ESF) system when one or more of the channels are out of service. Reactor safety is provided by RPS, which automatically initiates appropriate action to prevent exceeding established limits. Safety is not compromised, however, by continued operation with certain instrumentation channels out of service since provisions were made for this in the plant design.

The RPS and most engineered safety feature channels are supplied with sufficient redundancy to provide the capability for channel test at power, except for backup channels such as derived circuits in the ESF logic system.

When one of the four channels is taken out of service for maintenance, RPS logic can be changed to a two-out-of-three coincidence for a reactor trip by bypassing the removed channel. If the bypass is not effected, the out-of-service channel (Power Removed) assumes a tripped condition (except high rate-of-change of power, high power level and high pressurizer pressure),(') which results in a one-out-of-three channel logic. If in the 2-out-of-4 logic system of the RPS one channel is bypassed and a second channel manually placed in a tripped condition, the resulting logic is 1-out-of-2. At rated power, the minimum OPERABLE high-power level channel is 3 in order to provide adequate power tilt detection. If only 2 channels are OPERABLE, the reactor power level is reduced to 70%

rated power which protects the reactor from possibly exceeding design peaking factors due to undetected flux tilts and from exceeding dropped CEA peaking factors.

An RPS Logic matrix channel consists of two matrix power supplies, four matrix relays and their associated contacts as well as all interconnecting wiring. An RPS Trip Initiation Logic channel consists of an M contactor and associated contacts, an interposing relay and all interconnecting wiring. Two RPS Trip Initiation Logic channels associated with the same pair of CEDM clutch power supplies are considered to affect the same trip leg.

Integrated into the trip initiation logic are two RPS Manual Trip channels. Manual Trip #1 operates by directly de-energizing all four M contactors in response to the operation of a manual pushbutton. Manual Trip #2 operates by de-energizing an undervoltage relay which results in the opening of two circuit breakers, CB-AB and CB-CD, which supply power to the CEDM clutch power supplies. Manual Trip channel #1 consists of manual trip pushbutton #1 and interconnecting wiring. Manual Trip channel #2 consists of manual trip pushbutton #2, circuit breakers CB-AB and CB-CD, and associated interconnecting wiring.

With one manual reactor trip channel inoperable, operation may continue until the reactor is shut down for other reasons. No safety analyses assume operation of the Manual trip.

Because of this, the Required Action is to restore the inoperable channel to OPERABLE status prior to entering MODE 2 from MODE 3 during the next plant startup.

2.15 - Page 5 Amendment No. 88,125,152,173,194, 208, 244 TSBC

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control Systems (Continued)

Basis (Continued)

The ESF logic system is a Class 1 protection system designed to satisfy the criteria of IEEE 279, August 1968. Two functionally redundant ESF logic subsystems "A" and "B" are provided to ensure high reliability and effective in-service testing. These logic subsystems are designed for individual reliability and maximum attainable mutual independence both physically and electrically. Either logic subsystem acting alone can automatically actuate engineered safety features and essential supporting systems.

All Engineered Safety Features are initiated by 2-out-of-4 logic matrices except containment high radiation which operates on a 1-out-of-2 basis. The number of installed channels for Containment Radiation High Signal (CRHS) is two. CRHS isolates the containment pressure relief, air sample and purge system valves.

Entry into Technical Specification 2.15.1(3) is made when conditions have caused one logic subsystem ("A" or "B") to become inoperable but the redundant logic subsystem remains operable. The loss of a prime initiation relay (which renders all 4 channels of a logic subsystem inoperable) is the condition most likely to cause entry into Technical Specification 2.15.1(3). In this situation, the remaining ESF logic subsystem still has the capability to automatically actuate engineered safety features equipment and essential supporting systems. The 48-hour completion time is commensurate with the importance of avoiding the vulnerability of a single failure in the remaining ESF logic subsystem.

Technical Specification 2.15.1(3) will not be used upon loss of the common channels that affect both "A" and "B" subsystems prime initiators operability unless the permissible bypass condition is met. Upon exiting TS 2.15.1(3) following the restoration of a prime initiation relay to OPERABLE status, if any channel(s) remain inoperable, the appropriate Limiting Conditions for Operation (LCO) (TS 2.15.1 (1) orTS 2.15.1(2) is applicable with the length of inoperability measured from time of discovery of: 1) prime initiation relay inoperable, or 2) channel inoperability, whichever is longer.

The ESF system provides a 2-out-of-4 logic on the signals used to actuate the equipment connected to each of the two emergency diesel generator units.

The rod block system automatically inhibits all CEA motion in the event a LCO on CEA insertion, CEA deviation, CEA overlap or CEA sequencing is approached. The installation of the rod block system ensures that no single failure in the control element drive control system (other than a dropped CEA) can cause the CEAs to move such that the CEA insertion, deviationsequencing or overlap limits are exceeded. Accordingly, with the rod block system installed, only the dropped CEA event is considered an Anticipated Operational Occurrence (AOO) and factored into the derivation of the Limiting Safety System Settings (LSSS) and LCO. With the rod block function out-of-service, several additional CEA deviation events must be considered as AOOs. Analysis of these incidents indicates that the single CEA withdrawal incident is the most limiting of these events. An analysis of the at-power single CEA withdrawal incident was performed for Fort Calhoun for various initial Group 4 insertions, and it has been concluded that the LCO and LSSS are valid for a Group 4 insertion of less than or equal to 15%.

2.15 - Page 6 Amendment No. 1-25,194,209, 2-49 TSBC 06-001-0 TSBC

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control Systems (Continued)

Basis (Continued)

Operability of the primary CEA position indication system (CEAPIS) channel and the secondary CEAPIS channel is required to determine CEA positions and thereby ensure compliance with the CEA alignment and insertion limits of TS 2.10.2. The primary CEAPIS channel utilizes the output of a synchro transmitter geared to the clutch output shaft. CEA position is displayed visually at the main control panel.

The secondary CEAPIS channel utilizes the output of a voltage divider network controlled by a series of reed switches. The reed switches are actuated by a permanent magnet attached to the rack assembly. Position information is supplied to the distributed control system (DCS) flat-panel touch monitors for simultaneous viewing of all CEA group positions.

Limit switches o n the regulating CEAs and reed switches on the shutdown CEAs provide an additional means of determining CEA position when the CEAs are fully inserted or fully withdrawn. When the CEAs are fully inserted or fully withdrawn, this indication (displayed on the DCS) can be used in lieu of secondary CEAPIS data to meet the shiftly CHANNEL CHECK of primary CEAPIS. However, as limit switch indication is not fully independent of secondary CEAPIS, primary CEAPIS must be used to verify secondary CEAPIS data.

In MODES 1 and 2, CEA position indication is required to allow verification that the CEAs are positioned and aligned as assumed in the safety analysis. If one CEA position indication channel is inoperable for one or more CEAs, TS 3.1, Table 3-3, Item 4 (CEA position verification) must be performed within 15 minutes following any CEA motion in that group to ensure that the CEAs are positioned as required.

The operability of the Alternate Shutdown Panel (Al-1 85), including Wide Range Logarithmic Power and Source Range Monitors on AI-212, and Emergency Auxiliary Feedwater Panel (A1-179) instrument and control circuits ensures that sufficient capability is available to permit entry into and maintenance of the Hot Shutdown Mode from locations outside of the Control Room. This capability is required in the event that Control Room habitability is lost due to fire in the cable spreading room or Control Room.

Variances which may exist at startup between the more accurate AT-Power and Nuclear Instrumentation Power (NI-Power) are not significant for enabling of the trip functions. By 15% of rated power as measured by the uncalibrated NI Power, the Axial Power Distribution (APD) and Loss of Load (LOL) trip functions are enabled while the High Rate of Change of Power trip is bypassed.

The APD trip function acts to limit the axial power shape to the range assumed in the setpoint analysis. Significant margins to local power density limits exist at 15% power, as well as power levels up to at least 30% (where NI calibration occurs).

2.15 - Page 7 Amendment No. 2-08, 249 TSBC-04-001 -0 TSBC-08-003-0 TSBC 08-008-0 TSBC-1 1-006-0

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.15 Instrumentation and Control Systems (Continued)

Basis (Continued)

The LOL trip function acts as an anticipatory trip for the high pressurizer pressure and high power trips in order to limit the severity of a LOL transient. This trip is not credited in the USAR Chapter 14 Safety Analyses and any variance between AT-Power and NI-Power has no effect on the safety analysis.

The High Rate of Change of Power trip acts to limit power excursions from low power levels and bypassing of this trip at a high power level is conservative. This trip is not credited in the USAR Chapter 14 Safety Analyses for Mode 1 operation. Any variance between AT-Power and NI-Power has no effect on the safety analysis.

Steam generator blowdown isolation ensures that the auxiliary feedwater system performs its design function of maintaining adequate steam generator (SG) water level for decay heat removal once the auxiliary feedwater actuation signal (AFAS) is actuated. The steam generator blowdown isolation function consists of two trains (logic subsystems). Each train closes one SG blowdown isolation valve to each SG. Each SG has redundant (Train A and Train B) blowdown isolation valves. Four clutch power relays initiate closure of the SG blowdown isolation valves with each clutch power relay closing one valve when the reactor trips. Failure of one clutch power relay to initiate SG blowdown isolation or failure of one train will not prevent single valve isolation of SG blowdown flow.

References (1) USAR, Section 7.2.7.1 2.15 - Page 8 Amendment No. 208, 249 TSBC-08-008-0 TSBC-09-009-0

TECHNICAL SPECIFICATIONS TABLE 2-2 Instrument Ooeratina Reauirements for Reactor Protective System Test, Maintenance Minimum Minimum Permissible and Functional Operable Degree of Bypass Inoperable No. Unit Channels Redundancy Condition Bypass 1 Not Used 2 High Power Level 1 (c) Thermal Power (e)

Input Bypassed below 104% of Rated Power(a)(d) 3 Thermal Margin/Low 1 Below 104% of (e) 2 (b)

Pressurizer Pressure Rated Power(a)(d) 4 High Pressurizer 1 None (e)

Pressure 5 Low R.C. Flow 1 Below 104% of (e)

Rated Power (a)(d) 6 Low Steam Generator 2/Steam 1/Steam None (e)

Water Level Gen(b) Gen 7 Low Steam Generator 2/Steam 1/Steam Below 600 (e)

Pressure Gen(b) Gen psia(a)(d) 8 Containment High 1 During Leak Test (e)

Pressure 9 Axial Power 2 (b)(c) 1(C) Below 15% of (e)

Distribution Rated Power(g) 10 High Rate Trip-wide 2 (b) Below 104% and (e)

Range Log Channels above 15% of Rated Power(a)(g) 11 Loss of Load Below 15% of (e)

Rated Power(g) 12 Steam Generator None (e)

Differential Pressure

a. Bypass automatically removed.

b. Specification 2.15.1(2) is applicable.

2.15 - Page 9 Amendment No. 60,77,99,153,194, 2-49 I

TECHNICAL SPECIFICATIONS TABLE 2-2 (Continued)

c. If two channels are inoperable, load shall be reduced to 70% or less of rated power.

d. For low power physics testing this trip may be bypassed up to 10-1% of rated power.

e. Specification 2.15.1 (1) is applicable.

f. Deleted.

g. For each channel, the same bistable automatically activates the Loss of Load and Axial Power Distribution (APD) trips and automatically bypasses the high rate trip at 15% of rated power. Only the APD trip is a Limiting Safety System Setting. Therefore, the bistable is set to actuate within the APD tolerance band.

2.15 - Page 10 Amendment No. 60,77,88, 104, 249 I

TECHNICAL SPECIFICATIONS TABLE 2-3 Instrument Operating Requirements for Engineered Safety Features Test, Minimum Minimum Permissible Maintenance Functional Operable Degree of Bypass and Inoperable No. Unit Channels Redundancy Condition Bypass 1 Safety Iniection A Manual 1 None None N/A B High Containment Pressure Logic Subsystem A 2 (a) 1 During Leak (f)

Logic Subsystem B 2 (a)(d)(I) 1 Test C Pressurizer Low/Low Pressure Logic Subsystem A 2 (a)(d)(I) 1 Reactor Coolant (f)

Logic Subsystem B 2 (a)(d)(I) 1 Pressure Less Than 1700 psia(b) 2 Containment Spray A Manual(m) 1 None None N/A B High Containment Pressure Logic Subsystem A 2 (a)(c)(d)(I) 1 During Leak (f)

Logic Subsystem B 1 Test 2 (a)(c)(d)()

C Pressurizer Low/Low Pressure Logic Subsystem A 2 (a)(c)(d)(1) 1 2(a)(c)(d)(1 1 Reactor Coolant (f)

Logic Subsystem B Pressure Less Than 1700 psia(b)

D Steam Generator Low/Low Pressure Logic Subsystem A 2/Steam 1/Steam Steam Generator (f)

Gen (a)(c)(d)(1)

Gen Pressure Less Than Logic Subsystem B 2/Steam 1/Steam 600 psia(n)

Gen(a)(c)(d)(1) Gen 3 Recirculation A Manual 1 None None N/A B SIRW Tank Low Level Logic Subsystem A 2 (a)(k)(1) 1 None (j) 2 (a)(k)(1) 1 Logic Subsystem B 4 Emergency Off-Site Power Trip A Manual 1 (e) None None N/A B Emergency Bus Low Voltage (Each Bus)

-Loss of Voltage 2 (d) 1 Reactor Coolant 2 (a) (d) 1 (f)

-Degraded Voltage Temperature Less Than 300' F 2.15 - Page 11 Amendment No. 41,66,88,173,181,194,2Q0, I

TECHNICAL SPECIFICATIONS TABLE 2-3 (Continued)

Test, Minimum Minimum Permissible Maintenance Functional Operable Degree of Bypass and Inoperable No. Unit Channels Redundancy Condition Bypass 5 Auxiliary Feedwater A Manual 1 None None N/A B Auto. Initiation Logic Subsystem A Operating Modes Logic Subsystem B 3, 4, and 5

-Steam Generator 2 (a)(d)(1) 1 (h)

Low Level

-Steam Generator 3 (a)(g)(1) 1 (i)

Low Pressure

-Steam Generator 3 (a)(g)(1) 1 (i)

Differential Pressure a Circuits on ESF Logic Subsystems A and B each have 4 channels.

b Auto removal of bypass above 1700 psia.

c Coincident containment high pressure, pressurizer low/low pressure, and steam generator low pressure signals are required for initiation of containment spray.

d If minimum OPERABLE channel conditions are reached, one inoperable channel must be placed in the tripped condition or low level actuation position for auxiliary feedwater system within eight hours from the time of discovery of loss of operability. Specification 2.15.1(2) is applicable.

e Control switch on incoming breaker.

f If one channel becomes inoperable, that channel must be placed in the tripped or bypassed condition within eight hours from time of discovery of loss of operability. Specification 2.15.1 (1) is applicable.

g Three channels required because bypass or failure results in auxiliary feedwater actuation block in the affected channel.

h Specification 2.15.1 (1) is applicable.

2.15 - Page 12 Amendment No. 65,88,194,249, 2-651

TECHNICAL SPECIFICATIONS TABLE 2-3 (Continued)

If the channel becomes inoperable, that channel must be placed in the bypassed condition within eight hours from time of discovery of loss of operability. If the channel is not returned to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from time of discovery of loss of operability, one of the eight channels may continue to be placed in the bypassed condition provided the Plant Review Committee has reviewed and documented the judgment concerning prolonged operation in bypass of the inoperable channel. The channel shall be returned to OPERABLE status no later than during the next cold shutdown. If one of the four channels on one steam generator is in prolonged bypass and a channel on the other steam generator becomes inoperable, the second inoperable channel must be placed in bypass within eight hours from time of discovery of loss of operability. If one of the inoperable channels is not returned to OPERABLE status within seven days from the time of discovery of the second loss of operability, the unit must be placed in hot shutdown within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

If one channel becomes inoperable, that channel must be placed in the bypassed condition within eight hours from time of discovery of loss of operability. If the channel is not returned to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from time of discovery of loss of operability, one of the eight channels may continue to be placed in the bypassed condition provided the Plant Review Committee has reviewed and documented the judgment concerning prolonged operation in bypass of the inoperable channel. The channel shall be returned to OPERABLE status no later than during the next cold shutdown. If a channel is in prolonged bypass and a channel on the opposite train becomes inoperable, the second inoperable channel must be placed in bypass within eight hours from time of discovery of loss of operability. If one of the inoperable channels is not returned to OPERABLE status within seven days from the time of discovery of the second loss of operability, the unit must be placed in hot shutdown within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

k Specification 2.15.1(2) is applicable.

Specification 2.15.1(3) is applicable. If ESF Logic Subsystems A and B are inoperable, enter Specification 2.0.1.

m Steam Generator Low Pressure permissive is required for actuation.

n Auto removal of bypass prior to exceeding 600 psia.

2.15 - Page 13 Amendment No. 89,173,194,2- , 2558 I

TECHNICAL SPECIFICATIONS TABLE 2-4 Instrument Operating Conditions for Isolation Functions Test, Minimum Minimum Permissible Maintenance Functional Operable Degree of Bypass and Inoperable No. Unit Channels Redundancy Condition Bypass 1 Containment Isolation A Manual 1, None None N/A B Containment High Pressure Logic Subsystem A 2(a)(e)(g) 1 During Leak (f)

Logic Subsystem B 2(a)(e)(g) 1 Test C Pressurizer Low/Low Pressure Logic Subsystem A 2 (a)(e)(g) 1 Reactor Coolant (f)

Logic Subsystem B 2(a)(e)(g) 1 Pressure Less Than 1700 psia(b) 2 Steam Generator Isolation A Manual 1 None None N/A B Steam Generator Isolation 1 None None N/A (i) Steam Generator Low Pressure Logic Subsystem A 2/Steam 1/Steam Steam Generator (f)

Gen(a)(e)(9) Gen Pressure Less Than 600 psia(c)

Logic Subsystem B 2/Steam 1/Steam Gen(a)(e)(g) Gen (ii) Containment High Pressure Logic Subsystem A 2 (a)(e)(g) 1 During Leak (f)

Logic Subsystem B 2(a)(e)(g) 1 Test 3 Ventilation Isolation A Manual 1 None None N/A B Containment High Radiation Logic Subsystem A 1 (d)(g) None IfContainment (f)

Logic Subsystem B l(d)(g) None Relief and Purge Valves are Closed 4 Steam Generator Blowdown Isolation A Manual 1 (h) None Operating Modes N/A 3,4, & 5 B Reactor Trip 2 (h)(i) None Operating Modes (j)

Trains A and B 3, 4, & 5 OR ifat least one valve for each steam generator is closed 2.15 - Page 14 Amendment No. 88,93,108,152,153,173,184,190,249, I

TECHNICAL SPECIFICATIONS TABLE 2-4 (Continued) a Circuits on ESF Logic Subsystems A and B each have 4 channels.

b Auto removal of bypass prior to exceeding 1700 psia.

c Auto removal of bypass prior to exceeding 600 psia.

d Aand B trains are both actuated by either the Containment or Auxiliary Building Exhaust Stack initiating channels. The number of installed channels for Containment Radiation High Signal is two for purposes of Specification 2.15.1 (1).

e Ifminimum operable channel conditions are reached, one inoperable channel must be placed in the tripped condition within eight hours from the time of discovery of loss of operability. Specification 2.15.1(2) is applicable.

Ifone channel becomes inoperable, that channel must be placed in the tripped or bypassed condition within eight hours from the time of discovery of loss of operability. Specification 2.15.1 (1) is applicable.

g Specification 2.15.1(3) is applicable. IfESF Logic Subsystems A and B are inoperable, enter Specification 2.0.1.

h. "Minimum Operable Channels" for steam generator blowdown isolation refers to the minimum number of trains (logic subsystems) which are required to be operable to provide manual or automatic SG blowdown isolation.

Ifboth trains become inoperable, power operation may continue provided at least one SG blowdown isolation valve for each steam generator is closed OR be in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 3 in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Specifications 2.15.1(1), (2), (3) and (4) are not applicable; TS LCO 2.0.1 is not applicable.

j. Ifone train becomes inoperable, that train may be placed in the bypassed condition. Ifthe train is not returned to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time of discovery of loss of operability, operation may continue as long as one SG blowdown isolation valve to each steam generator is closed. Ifthe train is not returned to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time of discovery, with blowdown not isolated to both SGs, be in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 3 in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Specifications 2.15.1(1), (2), (3)and (4) are not applicable; TS LCO 2.0.1 is not applicable.

2.15 - Page 15 Amendment No. 88,108,152,1739,194,249, 263 1

TECHNICAL SPECIFICATIONS TABLE 2-5 Instrumentation Operating Requirements for Other Safety Feature Functions Minimum Minimum Permissible Functional Operable Degree of Bypass No. Unit Channels Redundancy Condition 1 CEA Position Indication 1 (a) None None Systems 2 Pressurizer Level 1 None Not Applicable NOTES:

(a) If one channel of CEA position indication is inoperable for one or more CEAs, requirements of specification 2.15.1 are modified for item 1 to "Perform TS 3.1, Table 3-3. Item 4 within 15 minutes following any CEA motion in that group." Specifications 2.15.1(1), (2), and (3) are not applicable.

2.15- Page 16 Amendment No.-4,65,1 10,2*9, 265, 267 I 2-"8

TECHNICAL SPECIFICATIONS TABLE 2-6 Alternate Shutdown and Auxiliary Feedwater Panel Functions Function/Instrument Required Number or Control Parameter Location of Channels

1. Reactivity Control

a. Source Range Power AI-212 1

b. Reactor Wide Range AI-212 1 Logarithmic Power

2. Reactor Coolant System Pressure Control

a. Pressurizer Wide Range Al-179 Pressure (0-2500 psia)

3. Decay Heat Removal via Steam Generators

a. Reactor Coolant Hot Leg Al-185 1 (Note 1)

Temperature

b. Reactor Coolant Cold Leg Al-1 85 1 (Note 1)

Temperature

c. Steam Generator Pressure AM- 79 Generator 1 per Steam

d. Steam Generator Narrow Al-1 79 Generator Range Level 1 per Steam

e. Steam Generator Wide Al-1 79 Generator Range Level

4. Reactor Coolant System Inventory Controls

a. Pressurizer Level Al-1 85 1

b. Volume Control Tank Level Al-1 85 1

c. Charging Pump CH-1 B and Al-1 85 1 its associated controls

d. Charging Isolation Valve Al-185 Control

5. Transfer Functions

a. All Transfer Switches/Lockout AM-85 1 Relays

b. All Transfer Switches/Lockout AM-79 Relays Note 1: One reactor coolant hot leg temperature indication and one reactor coolant cold leg temperature indication channel must both be operable on the same steam generator (i.e., RC-2A or RC-2B).

2.15 - Page 17 Amendment No.

TECHNICAL SPECIFICATIONS TABLE 2-6 (Continued)

Alternate Shutdown and Auxiliary Feedwater Panel Functions Function/Instrument Required Number or Control Parameter Location of Channels

6. Auxiliary Feedwater Controls

a. Steam Generator RC-2A and Al-179 1 2B Auxiliary Feedwater Isolation Inboard and Outboard Valves Control

b. Steam-Driven Pump FW-10 A1-179 1 Recirculation Valve Control

c. Steam-Driven Pump FW-10 Al-1 79 1 Steam Isolation Valve Control

d. Steam from Steam Generator Al-1 79 1 RC-2A and RC-2B to FW-10 Steam Isolation Valve Control 2.15 - Page 18 Amendment No.

v t e Letter Sequence Request
TAC:ME8038, (Approved, Closed)
Initiation Request, Request Acceptance... Administration Questions 13 July 2012 31 October 2012 31 August 2012 Answers 1 October 2012 Results Approval Other: ML120830328
MONTHYEAR2012-03-23 [Table View]

LIC-12-0006, License Amendment Request (LAR) 12-01, Proposed Change to Establish the Reactor Protective System (RPS) Actuation Circuits Limiting Condition for Operation (LCO)

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LIC-12-0006, License Amendment Request (LAR) 12-01, Proposed Change to Establish the Reactor Protective System (RPS) Actuation Circuits Limiting Condition for Operation (LCO)

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5.0 ENVIRONMENTAL CONSIDERATION

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