Text

Quality Assurance (QA) Program Changes, Technical Specification Basis Changes, and Updated Safety Analysis Report (USAR) Revision
	ML12226A171
Person / Time
Site:	Fort Calhoun
Issue date:	06/18/2012
From:	Herman J; Omaha Public Power District
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	LIC-12-0076
	Download: ML12226A171 (49)
	v • d • e

444 South 16th Street Mall Omaha, NE 68102-2247 June 18, 2012 LlC-12-0076 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

References:

1.

Docket No. 50-285

2.

Letter from OPPD (J. A. Reinhart) to NRC (Document Control Desk), 10 CFR 50.59 Report, Quality Assurance (QA) Program Changes, Technical Specification Basis Changes, and Updated Safety Analysis Report (USAR) Revision for Fort Calhoun Station (FCS), Unit No.1, dated June 18,2010 (LlC-10-0040)

SUBJECT:

10 CFR 50.59 Report, Quality Assurance (QA) Program Changes, Technical Specification Basis Changes, and Updated Safety Analysis Report (USAR) Revision for Fort Calhoun Station (FCS), Unit No.1 In accordance with 10 CFR 50.59(d)(2), the Omaha Public Power District (OPPD) submits Attachment 1 as the report of changes, tests, and experiments periormed pursuant to 10 CFR 50.59 for FCS. is provided to describe Quality Assurance (QA) Program (USAR Appendix A) changes, as required by 10 CFR 50.54(a)(4)(i). also contains a description of revised regulatory commitments that require Commission notification in accordance with NEI 99-04, "Guidelines for Managing NRC Cornmitment Changes," and modifications to the USAR made in accordance with NEI 98-03, "Guidelines for Updating Final Safety Analysis Reports."

In accordance with FCS Technical Specification 5.20.d, Attachment 3 provides a brief summary of the Technical Specification Basis Changes (TSBCs) made since the previous submittal (Reference 2) and Attachment 4 includes a copy of the revised pages.

The USAR is reissued in electronic format. Pursuant to 10 CFR 50.71 (e) and 10 CFR 50.4(b)(6), enclosed is one (1) original CD-ROM of the FCS USAR, which incorporates changes to the USAR made since the previous submittal (Reference 2) and includes changes made under the provisions of 10 CFR 50.59 but not previously submitted to the Commission. Attachment 5 contains a list of the files on the CD-ROM.

Employment with Equal Opportunity

U. S. Nuclear Regulatory Commission LlC-12-0076 Page 2 The Senior Resident Inspector is provided with an updated copy of the USAR by the FCS distribution process.

This information covers the period of June 19, 2010 through June 18, 2012.

In accordance with 10 CFR 54.37{b), a review of structures, systems, and components (SSCs) was performed.

No new SSCs subject to an aging management review or evaluation of time-limited aging analyses in accordance with 10 CFR 54.21 were identified.

No commitments to the NRC are made in this letter.

I declare under penalty of perjury that the foregoing is true and correct. (Executed on June 18, 2012).

Sincerely, B. Herman ivision Manager-Nuclear Engineering Attachments:

1.

Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59

2.

Quality Assurance Program Changes and Regulatory Commitments Revised in Accordance with NEI 99-04

3.

Summary of Technical Specification Basis Changes (TSBC)

4.

TSBC Pages

5.

List of Files on CD-ROM

Enclosure:

CD-ROM (1) of USAR Sections and Figures JBH/MLE/rnle c:

E. E. Collins, Jr., NRC Regional Administrator, Region IV L. E. Wilkins, NRC Project Manager J. C. Kirkland, NRC Senior Resident Inspector (w/o Enclosure)

LlC-12-0076 Page 1 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59

LlC-12-0076 Page 2 Abbreviations and Acronyms:

I I

I AFW - Auxiliary Feedwater NEI-Nuclear Energy Institute I

ANSI-American National Standards Institute NRC - Nuclear Regulatory Commission AOP - Abnormal Operating Procedure OPPD - Omaha Public Power District AR - Action Request POlL - Power Dependent Insertion Limit BAST - Boric Acid Storage Tank PRC - Plant Review Committee CD-ROM - Compact Disk Read-Only Memory QA Quality Assurance CEA - Control Element Assemb!y QATR - QuaJityAssurance Topical Rep_ort CEAPIS - CEA Position Indication System QR Qualified Reviewer CFR - Code of Federal Regulations RCA - Root Cause Analysis COLR - Core Operating Limits Report RCS - Reactor Coolant System i

CR - Condition Report RFO - Refueling Outage CRS - Control Room Supervisor RTD - Resistance Temperature Detector CS - Containment Spray RW - Raw Water CW - Circulating Water SAO - Safety Analysis for Operability DCS - Distributed Control System SARC - Safety Audit and Review Committee DG - Diesel Generator SOC - Shutdown Cooling EC - Engineering Change SER Safety Evaluation Report ERFCS - Emergency Response Facility Computer System SFP - SRent Fuel Pool FCS - Fort Calhoun Station, Unit No. 1 SG - Steam Generator FCSG - Fort Calhoun Station Guideline SM - Shift Manager FSAR - Final Safety Analysis Report SO - Standing Order HEPA - High Efficiency Particulate Air SR - Surveillance Requirement I&C -Instrumentation & Control SSC - Structures, Systems and Components INPO -Institute of Nuclear Power Operation ST - Surveillance Test LCO - Limiting Conditions for Operation TM - Temporary Modification LOCA - Loss of Coolant Accident TS Technical Specification LPSI-Low Pressure Safety Injection TSBC - Technical Specification Basis Change LTOP - Low Temperature Overpressure Protection UFSAR - Updated Final Safety Analysis Report MCC - Motor Control Center USAR - Updated Safety Analysis Report MFW - Main Feedwater VCT - Volume Control Tank msl - Mean Sea Level WO Work Order

LlC-12-0076 Page 3 2010 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PRC. As a result, the language may be in future tense.

EC 35741 Rev. 1 Turbine Controls System Replacement (10 CFR 50.59 Evaluation Rev. 0)

Traveling Screen Replacement (10 CFR 50.59 Evaluation Rev. 0) proposed activity moves the basic turbine control and turbine trip, steam dump and bypass, reactor regulating systems functionality from an analog to a digital platform. while reducing point vulnerabilities and improving system dependability. As a result, there is no increase the frequency of occurrence of an accident previously evaluated in the USAR or in the ikelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in USAR.

Because the proposed activity involves changes to plant control systems, it has the potential to affect events classified under the categories "normal operation" or "incidents of moderate

. However. these do not have radiological consequences. Therefore, the proposed activity does not increase the consequences of an accident previously evaluated in the USAR and does not increase the consequences of a malfunction of an SSC important to safety previously evaluated in the USAR.

proposed activity does not introduce new failure modes. Therefore, the proposed activity does not create a possibility for an accident of a different type than any previously evaluated in USAR or a possibility for a malfunction of an SSC important to safety with a different result any previously evaluated in the USAR.

Since the proposed activity maintains the existing functionality of the turbine control and turbine steam dump and bypass, and reactor regulating systems, control bands and protection system setpoints are not changed. Therefore. the proposed activity does not result in a deSign basis limit for a fission product barrier as described In the USAR being exceeded or altered.

Finally. the proposed activity does not change any method of evaluation described in the USAR used in establishing the design bases or the safety analyses.

ThArAfnrA the nlAmAntAn without NRC <:Innrn\\l<:>1 this evaluation are:

1) This modification will transfer screen control functions from the screen wash panel AI-120 to local Geiger control panels mounted on each individual screen as well as a Distributed Control System (DCS) cabinet AI 339. Similar to the existing panel AI-120 screen controls, the Geiger local control Danel interface uses all analoa hand switches. indications. and meters. The Alli..::tinn

L1C-12-0076 Page 4 2010 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PRC. As a result, the language may be in future tense.

EC 50056 Temporary Modification Spliced Cable EA 124 Installed by EC 49036 50.5~:;~~u~tion pneumatic bubbler intake water level supervisory system will be replaced with digital radar level transmitters which will provide level input to the DCS cabinet. The DCS cabinet will consist of digital components which will process the radar level inputs, via redundant processors, and provide control inputs to the traveling screens local control panels during the automatic mode of traveling screen operation. The change from an analog to a digital control and indication system may be considered as adverse.

The adverse effects have been considered and the following conclusions have been made as a result: 1) The shift in control and indication from an analog system to a digital system will improve accuracy, reduce maintenance (pneumatic air gaskets failure and tube clogging is no longer an issue), allow for greater flexibility and quality of indications and alarms to the operator (individual cell levels and hardware alarms will be available to the operator). The use of software adds a new type of malfunction to the system but the utilization of fully redundant processors (with redundant power supplies) prevents the introduction of any single point failure as well as any result that differs from those described in the USAR.

In the event the redundant DCS is completely disabled, the Geiger screens are still fully operational locally at the Geiger control panel.

The traveling screens control system is non-safety related.

None of the modified equipment is credited with accident prevention or mitigation or in mitigating the radiological consequences of a malfunction.

No new types of accident are created, and fission product barriers are not affected. No methods of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses are affected by this modification.

It is concluded that a License Amendment is not required prior to implementation of this modification.

to AuthorizelFeeder Cable EA124, 480 volt power to motor control center MCC-3A1 had a degraded cable which was section and required replacement.

The degraded EA 124 cable section was removed and replaced with new cable of the same size and electrical characteristics. The replacement cable section and the remaining existing cable were spliced together in the cable tray. The cable terminations to MCC-3A 1 were the same. This TM provides authorization for this spliced cable which is scheduled to be replaced by the end of the 2011 refueling outage. This 50.59 evaluation is being completed as a result of FCS taking a more conservative approach with respect to the 50.59 Applicability Screening performed as part of EC 49036 which installed this splice. Thus, installing this cable splice adversely affects the EA124 cable feeding the MCC components due to otential that likelihood of occurrence of a safety comDonent malfunction is more than minimall

LlC-12-0076 PageS 2010 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PRC. As a result, the language may be in future tense.

The cable splice installed under EC 49036 for the 480 volt power feeder cable to MCC-3A 1 was considered to be in compliance with the FCS USAR and plant design basis. The original cable and the cable splice are qualified to IEEE 383-1974 for Flame and LOCA performance. A failure of the cable splice would be no different than a failure of the original cable. A failure of either the original cable or the cable splice would result in the same loss of MCC*3A 1 and its connected loads. The loss of a MCC is bounded by the USAR and plant design basis. Therefore, the splice of cable EA124 is considered to be within the FCS design basis and USAR. However, this cable is now considered to be operable but degraded and the cable will be replaced during the 2011 refuelina outaae.

LlC-12-0076 Page 6 2011 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PRC. As a result, the language may be in future tense.

Temporary Modification to Drive T1231ReCOnfigUre T-123 loop by bypassing the loop RTD and the transmitter (TT-123) and using an From Brrr-122CA output signal from loop 2B cold leg temperature channel B (BfTT-122CA) to drive the reconfigured loop. This reconfiguration is necessary to compensate for failed RTD, T-123. This will require (10 CFR 50.59 Evaluation Rev. 0) installation (at AI-215) of a scaler to convert the 50-700QF range to 0-600°F range currently used the existing T-123 loop, a voltage to current converter to convert the voltage signal to 10-50ma signal, installing an instrument cable between the voltage to current converter and TT-123 Itransmitter, and finally connecting TT-123 output cable EB3392 to the converter cable. At AI-215, connect the output of BfTT-122CA to the scaler and voltage to current converter.

reconfigured loop is not an accident initiator and will not result in increase in the frequency occurrence of an accident previously evaluated in the UFSAR. The activity does not introduce possibility for a malfunction of an SCC with a different result because the activity does introduce a new failure mode. The reconfigured loop provides a signal to the Low Temperatu Overpressure Protection (LTOP) system. The LTOP is not a chapter 14 accident mitigating system, however, it provides peak pressure protection for transients that may occur while Reactor Coolant System (RCS) temperature is at -the LTOP enable temperature of 350°F, (indicated T-Cold). During review of calculation FC06863, Rev. 2 ilLTOP Setpoint Instrument Loop Uncertainty and L TOP Trip Curve Development," it was identified that the uncertainties introduced by the Temporary Modification will affect the LTOP setpoints Slightly in the non-conservativ",1 direction. The increase in uncertainty exceeds the analytical limit at some points of the variab LTOP protection curve.

While the temporary modification was successful in restoring loop functionality and it was evaluated to be acceptable for the purpose of the XC 105 calorimetric calculation, the operability the L TOP circuit has been affected. During power operation, the L TOP system is disabled and only enabled during plant cool-down and heat-up. Therefore, there is no plant operability at power. However, since the L TOP is enabled and required to be operable during down, the L TOP circuitry driven by the T-123 was declared inoperable. Reference CR 2011-0560.

Temporary Modification to Drive T123!Reconfigure T-123 loop by bypassing the loop RTD and the transmitter (TT-123) and using From Bm-122CA signal from loop 2B cold leg temperature channel B (BfTT-122CA) to drive the reconfigured This reconfiguration is necessary to compensate for failed RTD, T-123. This will EC 49330 (10 CFR 50.59 Evaluation Rev. 1) in"'~allation (at AI-215) of a scaler to convert the 50-700QF range to 0-600°F range currently the existing T-123 loop, a voltage to current converter to convert the voltage signal to 10-50m

~~.~n:~~ an instrument cable between the voltaae to current converter and TT-1

LlC-12-0076 Page 7 2011 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PRC. As a result, the language may be in future tense.

. 50.59 EV"~.9~~~Sl.Jmmary transmitter, and finally connecting TT-123 output cable EB3392 to the converter cable.

connect the output of BITT-122CA to the scaler and voltage to current converter.

The reconfigured loop is not an accident initiator and will not result in increase in the frequency of occurrence of an accident previously evaluated in the UFSAR. The activity does not introduce the possibility for a malfunction of an SCC [SSC] with a different result because the activity does not introduce a new failure mode. The reconfigured loop provides a signal to the Low Temperature Overpressure Protection (L TOP) system. The L TOP is not a chapter 14 accident mitigating system, however, it provides peak pressure protection for transients that may occur while the Reactor Coolant System (RCS) temperature is at the LTOP enable temperature of 350°F, (indicated T-Cold). During review of calculation FC06863, Rev. 2 "LTOP Setpoint Instrument Loop Uncertainty and L TOP Trip Curve Development," it was identified that the uncertainties introduced by the Temporary Modification will shift the L TOP analysis generated setpoint limits slightly closer to the plant L TOP setpoints. This slight shift was evaluated to be within the assumed analytical limits for the L TOP protection curve and still protected by the plant L TOP setpoints.

The temporary modification was successful in restoring loop functionality and operability, and the functionality and operability of both, the XC105 calorimetric calculation and the LTOP circuitry are not impacted based on the loop calibration results when the temporary modification was installed in July 2010 and again on March 8, 2011.

The Control Rod Power Dependent Insertion Limit (POlL) is currently calculated for alarming EC 50590 I XC411 Calculation Modification purposes in both the Distributed Control System (DCS) and the Emergency Response Facility Computer System (ERFCS). In the DCS, it is also used to generate a Rod Block if the POlL is violated.

Reactor power is a required input into these POlL calculations.

Reactor power is calculated identically in both the DCS and ERFCS using the Delta-T method. This reactor power is available as pOint X_411 in the DCS and point XC411 in the ERFCS. This evaluation will only refer to XC411, although the changes affect both points.

The XC411 calculation contains a constant that is used to add a conservative margin to the result of the Delta-T calculation. This EC will revise that constant (K406) from +4% to +1% to reduce the amount of conservatism in the calculation. This EC will also implement a filter on the output of XC411 to reduce the severe fluctuations caused by fluctuating temperature inputs into the calculation. This filter will be a first order lag filter with a 30-second time constant. This will result

'---____-'--______________--'",in XC411 reachina 63% of anv reactor Dower steD chanae within 30 seconds.

LlC-12-0076 Page 8 2011 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PRC. As a result, the language may be in future tense.

EC 32387-Turbine Controls Replacement EC will also remove the requirement that XC411 be greater than the secondary calorimetric Innw~r value XC105 in procedure RE-ST-C EA-OOO1, "Power Dependent Insertion Limits, Deviation Sequence Monitoring Test". The previous requirement will be replaced with a tolerance

/+3% that will be used to verify XC411 is calculating properly.

In addition, since the XC411 calculation is currently duplicated in the ERFCS and the DCS, 11 calculation will be removed from the ERFCS and the DCS will provide X_ 411 to th ERFCS for use as XC411.

The changes made to the XC411 calculation, used in the ERFCS and the DCS to calculate th POlL values that are used for alarming and Rod Block, do not have an adverse impact on systems, accident analysis or methodologies. The changes reduce, but do not eliminate th conservatism contained in the calculation. This change results in a more accurate calculation the reactor power value used in the ERFCS and DCS POlL calculations. The actual POlL values specified in the Core Operating Limits Report (COLR), and the methods used to determine th values, are not affected by this change. Therefore, a License Amendment is not required irnnlcrn",nt these....h",nnc."

SystemlThe proposed activity consists of (1) an upgrade of the controls for the main turbine and (2) changes to the steam dump and bypass and reactor regulating systems. It is part of an overall (10 CFR 50.59 Evaluation Rev. 1) phased project to upgrade select plant instrumentation and control (I&C) systems and integrate into a single distributed control system (DCS). Control system functions will be performed using digital instead of analog devices. The turbine emergency trip system will use electrical trains, and the mechanical trips will be eliminated. The operator interface currentlyl performed using analog meters, indicating lights, push buttons, and rotary knobs on the main control board will be performed using touchscreen monitors and associated workstations.

The proposed activity is being undertaken for two reasons: (1) to improve plant control system reliability by replacing components that are or will soon become obsolete and by eliminating single point vulnerabilities and (2) to automate certain tasks currently performed by the operators. The proposed activity moves the basic turbine control and turbine trip, steam dump and bypass, and reactor regulating systems functionality from an analog to a digital platform, while reducing single point vulnerabilities and improving system dependability. As a result, there is no increase in the frequency of occurrence of an accident previously evaluated in USAR or in the likelihood of occurrence of a malfunction of an SSC imoortant to EC 32387

LlC-12-0076 Page 9 2011 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PRC. As a result, the language may be in future tense.

.Ef* Qhal1~.*.*.*.

~p.~~Evaluation

'c. Number previously evaluated in the USAR.

Because the proposed activity involves changes to plant control systems, it has the potential to affect events classified under the categories "normal operation" or "incidents of moderate frequency." However, these do not have radiological consequences. Therefore, the proposed activity does not increase the consequences of an accident previously evaluated in the USAR and does not increase the consequences of a malfunction of an SSC important to safety previously evaluated in the USAR.

The proposed activity does not introduce new failure modes. Therefore, the proposed activity does not create a possibility for an accident of a different type than any previously evaluated in the USAR or a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the USAR.

Since the proposed activity maintains the existing functionality of the turbine control and turbine trip, steam dump and bypass, and reactor regulating systems, control bands and protection system setpoints are not changed. Therefore, the proposed activity does not result in a design basis limit for a fission product barrier as described in the USAR being exceeded or altered.

Finally, the proposed activity does not change any method of evaluation described in the USAR used in establishing the design bases or in the safety analyses.

Therefore, the proposed activitv can be imolemented without prior NRC aoproval.

FC08034 Calculation FC08034:

Diesel FUel\\A calculation is being prepared to determine the adequacy of onsite diesel fuel inventory for Usage During a Severe Flooding Event meeting the 7 day diesel generator (DG) operating time described in Technical Specification 2.7 Basis section as applied to a site-flooding scenario. This calculation is being prepared to address the concern raised in condition report (CR) 2011-3842. For the flooding scenario, in contrast to the description in TS 2.7 Basis and USAR 8.4, the calculation makes an assumption that fuel storage tank FO-10 is unavailable. It was determined that onsite diesel fuel inventory is not adequate to meet the 7 day DG operating requirement.

Therefore, the calculation proposes procedure changes to AOP-01 (Acts of Nature) that can be made to ensure that the 7 day requirement for onsite diesel fuel inventory can be met without relying on FO-10. The proposed procedure change will require operating a single DG to maintain safe shutdown loads as opposed to oDeratinq both DGs as currentlv prescribed in AOP-01.

In the 10 CFR 50.59 screenin

LlC-12-0076 Page 10 2011 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PRC. As a result, the language may be in future tense.

process, it was determined that an evaluation is necessary because the calculation is not crediting FO-10 whereas USAR section 8.4.1 states that FO-10 is available for use. Further, a single DG will be procedurally directed for use with the other DG removed from service. These changes are concluded to have a potentially adverse effect on USAR described design functions.

The proposed calculation which does not credit diesel fuel storage tank FO-10 for maintaining emergency power to safe shutdown equipment during a flooding event does not require prior approval because analysis has demonstrated that adequate fuel inventories exist without it for this scenario. The proposed activity to operate a Single DG for 7 days following the onset of a site flooding event with the plant in a cold shutdown condition does not require prior approval because

1) a single DG is capable of providing power to all equipment required to maintain safe shutdown,

2) the non-running DG is available immediately should a failure of the operating DG occur, and 3) with the Dlant in cold shutdown TS 2.7 Dermits disablina the auto-start feature of a DG.

EC 54058 I 01-SC-3 Procedure Change to Allow A procedure change to 01-SC-3, Alternate Shutdown Cooling Utilizing Containment Spray Pumps, Closing of HCV-335 While on Alternate will allow closing HCV-335, SHUTDOWN CLG HT EXCHS AC-4A&B INLET HEADER Shutdown Cooling ISOLATION VALVE, with alternate shutdown cooling in operation. The procedure change also pins HCV-341, SHUTDOWN CLG HT EXCHS AC-4A&B OUTLET TEMPERATURE CONTROL VALVE, in the open position, and closes FCV-326, SHUTDOWN CLG HT EXCHS AC-4A & 4B LPSI BYPASS FLOW CONTROL VALVE at the SM/CRS discretion. Shutdown cooling will be maintained manually at approximately 1500 gpm by throttling the LPSI Loop Injection Valves with HCV-341 pinned open/throttled. Indication of LPSI flow is available on ERF page 194 (SFLPSI),

which is the summation of flow transmitters, F328, F330, F332, and F334.

Current procedure guidance and UFSAR described flow paths use valves, HCV-335, FCV-326, and HCV-341, for normal Shutdown Cooling with LPSI pumps. Procedure guidance found in 01 SC-3, uses both HCV-335 and HCV-341 as part of the flow path from the Loop 2 Hot Leg to the suction of the pumps and back through the SOC Heat Exchangers (AC-4A and AC-4B).

Furthermore, cooling flow is maintained at approximately 1500 gpm by adjusting FCV-326. HCV 341 is used to adjust flow through the SOC Heat Exchangers to maintain the desired RCS temperature and/or cooldown rate.

This change is necessary to allow closing of HCV-335 while on alternate shutdown cooling.

Current procedural guidance has HCV-335 open while on alternate shutdown cooling.

LlC-12-0076 Page 11 2011 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PRC. As a result, the language may be in future tense.

The proposed change to close HCV-335 while on alternate Shutdown Cooling with Containment Spray Pumps does not affect the Shutdown Cooling Systems ability to maintain reactor coolant temperature. Furthermore, the proposed changes to close FCV-326 and to open/throttle HCV-341 do not affect the ability to maintain Shutdown cooling while on alternate Shutdown Cooling with Containment Spray Pumps. Containment spray pumps can be considered as available shutdown COOling pumps with RCS temperature < 120 degrees F and minimum vent of 47 square inches.

Current procedure and UFSAR described flow paths have HCV-335 open and FCV-326, HCV-341 throttled to maintain RCS temperature.

The closed position of HCV-335 during altemate shutdown cooling is not an accident initiator and will not result in an increase in the frequency of occurrence of an accident previously evaluated in the UFSAR. The activity does not introduce the possibility for a malfunction of an SSC with a different result because the activity does not introduce a new failure mode. HCV-335 in the closed position with AC-4A or AC-4B in operation along with HCV-341 pinned open, and the LPSI Injection valves throttled during alternate shutdown cooling will provide sufficient flow and flow indications for alternate SDC. This evaluation was initiated because the UFSAR only addresses one shutdown cooling flow path for normal operation and does not describe the altemate SDC flowoath.

LlC-12-0076 Page 12 2012 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PRC. As a result, the language may be in future tense, EC53392 OI-CW-1 Attachment 18 Implementation of a procedure change to OI-CW-1 will add a new attachment that allows the development of a raw water flow path from the discharge tunnel, backwards through the circulating water system to the intake cells (full description of this flow path is provided in attachment 1). This flow path would be implemented to maintain the intake cell level during flooding conditions if the river gates (CW-14A-F, traveling screen sluice gates) were to sand-in to the extent that flow into the intake cell was insufficient to provide for raw water pump flow. This alignment will also allow for an alternate flow path that is less vulnerable to sanding and more controllable from the intake structure. The flow control in the existing configuration is limited because of the vulnerability of the intake gate motor operators which become submerged at a river level of 1010 feet msl. After the alternate flow path is established, the flow rate to the intake cells can be controlled using valves that are accessible from the intake structure and capable of manual actuation for fine control.

The new attachment to OI-CW-1 allows repositioning some of the components in the CW system as to create a new RW flow path from the discharge tunnel into the condenser waterboxes and backwards through the condenser inlet valves to the intake tunnel and CW cells. This flow path would be implemented to maintain the intake cell level during flooding conditions if the river gates (CW-14A-F, traveling screen sluice gates) were to sand-in to the extent that flow into the intake cell was insufficient to provide for raw water pump flow. This alignment will also provide a better control of flow into the CW suction cells, as it does not rely on operation of the river sluice gates (CW-14A-F), which could be either submerged or inaccessible.

The changes included in this attachment consist of repositioning certain CW system components (motor operated valves, check valves and CW cell sluice valves) as to provide an additional flow path from the RW discharge into the intake cells. There are no changes to the RW system normal alignment that could affect its heat removal function. The evaluation concludes that the proposed change does not increase the frequency or consequences of a previously evaluated accident nor does it create a possibility of a different type of accident. The changes do not impact any fission product barrier and does not involve a change in methodology described in the UFSAR.

The evaluation also concludes that there are no new failure modes or mechanisms that could increase either the likelihood or the consequences of a malfunction of a SSC, nor can any failure of a component result in different consequences than previously evaluated in the UFSAR.

LlC-12-0076 Page 13 2012 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PRC. As a result, the language may be in future tense.

Therefore, the changes described in the new attachment can be implemented without requiring rior NRC aooroval.

Implementation!LiCense Renewal (LR) Commitment #31 (USAR 15.4) is to perform a one-time inspection of the Circulating Water (CW) discharge tunnel to verify aging of the CW tunnel as part of the structures EC 55478 Change USAR 15.4 Schedule monitoring program. This activity would change LR Commitment #31 implementation schedule (as listed in USAR 15.4) from "prior to the period of extended operation" (August 9,2013) to "RFO 27". The intent of LR #31 is to verify that aging effects are not occurring in the CW tunnel or that the aging effect is progressing at such a slow rate it will not impact the intended function during the PEO (USAR 15.2). A structural inspection has not been performed on the internal portion of the CW discharge tunnel but a structural inspection has been performed on the accessible portions of the Intake Structure building including the intake cells. The discharge tunnel is part of the Intake Structure and both are founded on piles to bedrock.

These aging effects could adversely affect the safety related function of the Intake Structure to support a safe shutdown of the plant (USAR 5.11).

As discussed in NEI 99-04, "Guideline for Managing NRC Commitment Changes," 10 CFR 50.59 should be applied to changes in commitments that are embodied in UFSAR descriptions of the facility or procedures (FCSG-23 4.2.1 N4a). The change under consideration is a delay to the implementation schedule in the USAR 15.4, License Renewal Commitment listing for commitment

31. Delaying LR commitment #31 implementation schedule from "prior to the period of extended operation" to "RFO 27" was evaluated for adverse effects on the ability of the Intake Structure to support safe shutdown of the plant (USAR 5.11).

Extending the inspection implementation schedule will not impact the intended function of the CW discharge tunnel based on the consensus that aging effects are not occurring or are progressing at a slow rate. In conclusion, this activity does not require a License Amendment.

This EC was subsequently canceled and the inspection of the CW discharge tunnel was erformed as reauired.

HCV-335, the shutdown cooling (SOC) heat exchangers AC-4A and AC-4B inlet header isolation HCV-335 Seat Leakage Test WO valve has shown evidence of leakage past its seat when closed. HCV-335's safety function in the 438017 closed position is to keep the low pressure safety injection (LPSI) and containment spray (CS) flow paths distinct during accident conditions so that each path retains adequate flow to perform its accident mitiaation function.

Therefore, a troubleshootina plan is required to auantifv the

L1C-12-0076 Page 14 2012 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PRC. As a result, the language may be in future tense.

01-AFW-4, Revision 43: Elimination Of FW-6 And FW-10 as Startup or Shutdown Pumps During Normal Plant Operations EC 29642 EVp.lu~tW~;;$llrntnary

.,', /"'<,",~ "'"

apparent leakage to assess whether HCV-335 requires maintenance. Troubleshooting will involve routing leakage past HCV-335 through the SDC purification lineup to the VCT where it will be trended and quantified.

Because of this diverted flow and the bypassing of the SOC heat exchangers, the SOC system will be declared inoperable and Technical Specification 2.8.1 (3) 1 LCO entered during testing. The troubleshooting plan will contain steps to return the SOC to a normal lineup, if required.

The total volume of leakage and rate of leakage flow will be insignificant in comparison to the total volume and flow rate of the SOC system. Under FCS' current conditions (consistent with those expected during testing), SDC is experiencing only minimal cooling from the SOC heat exchangers. Bypassing the heat exchangers will not significantly alter the core cooling for the duration of this test. During testing, operators will monitor critical SOC parameters to ensure that there is no adverse impact to keeping the core cooled and covered and the troubleshooting plan includes steps to stop testing should any unexpected conditions appear. Therefore, this activity will not result in more than a minimal increase in the likelihood, frequency, or consequences of accidents or malfunctions of safety-related SSCs described in the UFSAR. Similarly, the activity will not result in an accident or SSC malfunction that has not been previously evaluated in the USAR. SDC will provide flow to the core during testing, but will be declared inoperable and Technical Specification 2.8.1 (3)1 LCO entered, as flow will be isolated from the SOC heat exchangers. This activity will not have more than a minimal impact on refueling cavity inventory and temperature increase. This will not result in altering or exceeding any deSign basis limits of fission product barriers.

This activity involves direct manipulation of FCS plant operating conditions and does not require or involve implementation of any methodology used in establishina the desian bases or in the safety analvses.

Revision 43 to operating instruction 01-AFW-4 was approved in March 2002 under Engineering Change (EC) 29642.

EC 29642 added a restriction to 01-AFW-4 to prevent opening of the auxiliary feedwater (AFW) to main feedwater (MFW) isolation valve HCV-1384 when the reactor coolant system (RCS) is at a temperature greater than 300 degrees F.

The purpose of this procedure change was to ensure operability of the AFW system without reliance on operator action to close HCV-1384 if it were to become necessary to isolate AFW from MFW during plant startup.

At the time, AFW was considered inoperable with HCV-1384 open because this configuration cross-tied the safety grade AFW system to the non-safety grade MFW system. Prior to revision 43, credit was taken for operator action to maintain AFW operability.

L1C-12-0076 Page 15 2012 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PRC. As a result, the language may be in future tense.

EC 29642 was implemented without consideration of the fact that requiring HCV-1384 to closed with the RCS greater than 300 degrees F effectively eliminates the use of the motor-driven pump (FW-6) and the steam-driven AFW pump (FW-10) as feedwater injection during normal startup or shutdown evolutions. Since the original FSAR and the current U identify the AFW system as capable of providing feedwater for normal plant startup operation, the procedure change should have been evaluated under 10 CFR 50.59 as identified in the screening performed for the procedure change. (Note: The original procedure change under EC 29642 did not have a 10 CFR 50.59 screening performed as part of the procedure change packaQe. A 50.59 screening was performed for revision 43 in January 2012 as part of the actions condition report (CR) 2011-9416.) The activity being evaluated under this 50.59 is the elimination of FW-6 and FW-10 as startup or shutdown pumps during normal plant operations, relying solely on the use of diesel-driven AFW pump FW-54 as a startup pump.

proposed activity to use the diesel-driven AFW pump FW-54 as the sole startup (or shutdown) pump does not adversely impact the capability of the AFW system to meet its design function to

'provide decay heat removal capability either during plant startup or in the event of a loss of the running startup pump. The USAR described analyses for a loss of feedwater or a main steam break remain bounding and the Technical Specification operability requirements for the Ic>\\/c>torn will be maintained.

EC 55904 I Add to OI-SC-1 to MI.JV--:j-:jO. the shutdown cooling (SOC) heat exchangers AC-4A and AC-4B inlet header isolatinnl Perform Seat Leakage Testing of HCV-valve, has shown evidence of leakage past its seat when closed. HCV-335's safety function in 335 closed position is to keep the low pressure safety injection (LPSI) and containment spray flow paths distinct during accident conditions so that each path retains adequate flow to perform I accident mitigation function. Therefore, a method is required to quantify the apparent lea assess HCV-335's condition. OI-SC-1, Attachment 4 has been proposed to collect leal\\Cl~tI HCV-335 and route it through the SOC purification lineup to the VCT where it will be trenrlorl quantified. OI-SC-1, Attachment 4 also provides for isolation of HCV-341 to preclude IleaKaQe past HCV-341 from affecting results of testing HCV-335.

The SOC system will inoperable and Technical Specification 2.8.1 (3)2's eight-hour LCO entered Inorformance of the attachment.

OI-SC-1, Attachment 4 also contains steps to ensu inistrative controls are in place to return the SOC to service if required.

flow will be Im::tlt'lnlTl('~n in comparison to the total

LlC-12-0076 Page 16 2012 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PRC. As a result, the language may be in future tense.

EC 35741 Rev. 1 Traveling Screen Replacement e and flow rate of the SOC system. During testing, operators will monitor critical parameters to ensure that there is no adverse impact to FCS' ability to keep the core cooled and IfmlArAn and the procedure includes steps to stop testing should any adverse conditions appear this activity will not result in more than a minimal increase in the likelihood, frequency, or consequences of accidents or malfunctions of safety-related SSCs described in the UFSAR.

ilarly, the activity will not result in an accident or SSC malfunction that has not been previously uated in the USAR. SOC will provide flow to the core during testing, but will be declared linoperable and Technical Specification 2.8.1 (3)2's LCO entered as flow will be isolated from the SOC heat exchangers. This activity will not have more than a minimal impact on refueling linventory and temperature increase. This will not result in altering or exceeding any design basis limits of fission product barriers. This activity involves direct manipulation of FCS plant operating and does not require or involve implementation of any methodology used in establishing the design bases or in the safety analyses.

adverse effects considered in this evaluation are:

1) This modification will transfer screen control functions from the screen wash panel AI-120 to Geiger control panels mounted on each individual screen as well as a Distributed Control (DCS) cabinet AI-339. Similar to the existing panel AI-120 screen controls, the Geiger control panel interface uses all analog hand switches, indications, and meters. The existing matic bubbler intake water level supervisory system will be replaced with digital radar level Itransmitters which will provide level input to the DCS cabinet. Existing River Level Loop L-1900 bubbler will be left in service as an automatic substitution via the DCS for L-2000 in the event of a alfunction associated with that loop. The DCS cabinet will consist of digital components which will process the radar level inputs, via redundant processors, and provide control inputs to the

'tr~\\lAling screens local control panels during the automatic mode of traveling screen operation.

change from an analog to a digital control and indication system may be considered as adverse effects have been considered and the following conclusions have been made as

1) The shift in control and indication from an analog system to a digital system will improve accuracy, reduce maintenance (pneumatic air gaskets failure and tube clogging is no longer an issue), allow for greater flexibility and quality of indications and alarms to the cell levels and hardware alarms will be available to the ooerator). The

LlC-12-0076 Page 17 2012 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PRC. As a result, the language may be in future tense.

use of software adds a new type of malfunction to the system but the utilization of fully redundant processors (with redundant power supplies) prevents the introduction of any single point failure as well as any result that differs from those described in the USAR. In the event the redundant DCS is completely disabled, the Geiger screens are still fully operational locally at the Geiger control panel. The traveling screens control system is non-safety related. None of the modified equipment is credited with accident prevention or mitigation or in mitigating the radiological consequences of a malfunction. No new types of accident are created, and fission product barriers are not affected. No methods of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses are affected by this modification.

It is concluded that a License Amendment is not required prior to implementation of this modification.

EC 55306 I Temporary Modification - Install Plug fori This proposed Temporary Modification (TM) will remove the lower portion of the sheath of Pressurizer Heater #26 Pressurizer Heater #26, and bore out the sheath to remove the internals of the heater including the pressure boundary plug that formerly separated the heated and non-heated portions as well as the internal insulating materials. A small amount of the magnesium oxide insulation material may be left in place and will be exposed to reactor coolant. A restraining device will be installed inside the sheath to provide structural support to the sheath. The restraining device will then be bolted to a pressure boundary plug which will be welded to the heater sleeve (nozzle). This TM will initially be in effect only during Mode 5. The TM will then be incorporated into permanent EC 54959.

This evaluation is limited to the review of chemistry aspects of the TM in which the internal materials (heater element and insulation) are removed with the exception of no more than 0.2-inch of insulation (containing magnesium oxide) left at the top of the heater sheath. Given that the plug device permits reactor coolant in-flow and out-flow in the heater sheath containing up to a O.2-inch cylindrical piece of the insulation at the top of the heater sheath, this evaluation addresses the acceptability of this condition with respect to AREV A fuel chemistry guidelines and the FCS RCS chemistry limits prior to the RCS average temperature exceeding 250"F. This temperature is important because the magnesium will rapidly deposit out on heated fuel surfaces, so the important monitoring point for this element is prior to RCS heating by the fuel.

Temporary Modification (TM) EC 55306 proposes to remove the lower portion of the sheath of Pressurizer Heater #26 and bore out the sheath to remove the internals of the heater includina the

LlC-12-0076 Page 18 2012 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PRC. As a result, the language may be in future tense.

pressure boundary plug that formerly separated the heated and non-heated portions as well as the internal insulating materials. A small amount of the magnesium oxide insulation material will be left in place and will be exposed to reactor coolant. This evaluation shows that assuming all of the magnesium oxide goes into solution in the reactor coolant inside the subject pressurizer heater sheath and is transported into the RCS, the resultant RCS magnesium concentration will remain less than the allowable limits of both FCS Procedure CH-AD-0003 (Plant Systems Chemical Limits), Revision 88 and the AREVA chemistry guidelines for their fuel. Thus, the proposed TM is acceptable with respect to the concern of having magnesium deposit on the heated surfaces of the nuclear fuel and dearadina effective heat transfer from the fuel claddina.

LlC"' 12-0076 Page 1 Quality Assurance Program Changes and Regulatory Commitments Revised in Accordance with NEI 99-04

LlC-12-0076 Page 2 Revision 31 EC 48268

[02-23-10]

Revision 32 EC 50413

[06-03-11]

Revision 33 EC 55802

[05-25-12]

Revision 34 EC 56173

[06-07-12]

This change revises the SARC Composition listed in the USAR (including the QA Program, USAR Appendix A), removing specific position titles. This change ensures that there are at least five (5) SARC members, including the Chairman, to maintain compliance with ANSI N 18.7-1976 (Section 4.3.2.1) independent review program requirements.

ired SARC competencies are listed in USAR Appendix A, Section 2.6.1, and remain unchanged. Additionally, SARC function, authority, quorum, meeting frequency, and similar requirements remain unchanged.

('h~nn'" is in accordance with ANSI N18.7-1976 indeoendent review orooram reauirements and is not a reduction in commitment.

function of Manager - Nuclear Procurement Services is deleted. The function of approval of manufacturer or contractor selected supply the material, equipment, or services is approval by the Manager - Quality. The function of quality assurance applicable to manufacturer, contractor, or vendor such as audits, surveillances, and evaluation is assigned to the Quality Department.

function of vendor audits, surveillances, and evaluations is transferred to the Quality Department. All elements of the program transferred, including the content and the methods of the vendor qualification programs. Therefore this revision continues to the criteria of 10 CFR 50 Aooend ix B.

PRC review of administrative controls standing orders which do not require 10 CFR 50.59 or 72.48 evaluation is being removed.

addition, PRC review of Fire Protection Program Plan changes is being limited to changes which would adversely affect the ability achieve and maintain safe shutdown in the event of a fire. This QA Program change is a reduction in commitment but is permitted per 10 CFR 50.54(a)(3)(ii) and does not require prior NRC review and approval.

is QA Program change is consistent with the Exelon QATR changes (attached, reference letters RS-07-013 and RS-08-093 and related SERs) approved by the NRC in a Safety Evaluation Report (SER). The bases for the QA Program change are consistent with Exelon changes, and FCS continues to comply with 10 CFR 50 Appendix B, ANSI N18.7-1976, and RIS 99-002 controls for applicable changes.

These procedure change review and approval requirements continue to be satisfied by the Qualified Reviewer (QR) Prooram and olant administrative M Appendix A requires that the Plant Review Committee (PRC) review and approve corrective actions to resolve Iconditions that are determined to affect nuclear safety. This requirement is being deleted as it is redundant to the review performed by the Station Corrective Action Review Board (SCARB).

EC 56173 eliminates the redundant PRC review while preserving the commitment to ensure that the response to significantl conditions undergo internal independent review. The SCARB review includes the cause analysis and corrective actions. Both SCARB and PRC are chaired by a senior manager or a deSignated alternate.

The composition of both committees includes rl",..:::inn::lt",rl alternates and auorum reauirements are similar. The SCARB review of sianificant conditions

LlC-12-0076 PageS

AR 48576-01 AR 48576-02 AR 7991 specify that level A and B condition reports must be reviewed by the Corrective Action Review Board." The commitment has been revised to require that the Station Corrective Action Review Board review level A and Tier 1 level B condition reoorts.

During the 2011 Missouri River flood, the fire truck and trailer mounted pump used in a B.5.b scenario were temporarily brought inside the protected area. Thus, for a short period of time, the 100 yard distance from the power block noted in the B.5.b Mitigation Strategies SER (ADAMS Accession No. ML072040097) was not maintained. Due to flood waters blocking access to the site, this deviation from the B.5.b SER was necessary to increase the probability that the equipment would be available if needed. The fire truck and trailer mounted oumo are currentlv in excess of 100 yards from the oower block.

The B.5.b Mitigation Strategies SER notes that licensees who choose to conform to the NRC-approved resolution (NRC letter dated March 16,2006 (ML060690339)) are expected to include the following concept in procedures:

"Where feasible and practical, consistent with safe fuel handling practices, the licensee should make every attempt to pre-configure the spent fuel pool to enable direct placement of the expended assemblies from the vessel to the final distributed fuel pattern. Where this is not feasible or practical, licensees should distribute the fuel into the final pattern as soon as possible but no later than 60 days after subcriticality."

However, due to reactor shutdowns in December 2010 and February 2011, eight (8) Batch DD fuel assemblies did not meet the burnup requirement to be placed in Region " of the SFP as anticipated. With the fuel sequence for core offload already planned, FCS was not able to discharge those 8 fuel assemblies into the preferred pattern as the majority of the fuel assemblies in Region I were returning to the reactor and space in Region I remained full until May 31, 2011, when the reactor was loaded for Cycle 27. Due to preparations for the onset of Missouri River flooding, resource constraints prevented FCS from placing these 8 assemblies into the referred oattern until Julv 29, 2011.

This commitment was made in fetter LlC-88-1065 dated April 27, 1988. The commitment has been deleted. The letter stated:

"...changes to Fort Calhoun Station Standing Order G-43, Shift Technical Advisor, and Standing Order R-4, Station Incident Reports have been completed. The revised procedures instruct the Shift Technical Advisor to perform a safety assessment (10 CFR 50.59 evaluation) for any event that places the plant in an abnormal situation or for any plant parameter that affects or reflects an abnormal indication of a safety-related system. Upon completion of this assessment, the Shift Technical Advisor is instructed to initiate an Incident Report in conjunction with assisting the Shift Supervisor in determining one and four hour reportable events."

Per NEI-99-04, Guideline for Managing NRC Commitment Changes, a licensee may change or eliminate a commitment that has been determined to be unnecessary due to having been subsequently captured as part of an ongoing program or other administrative control that is subject to a revision review process.

LlC-12-0076 Page 4 Commitments Revised in Accordance with NEI 99-04 This commitment was made in letter L1C-08-0009 dated February 15, 2008. The commitment stated, "SO-R-2 will be revised to

~ -- -- --

LlC-12-0076 PageS

---~---~-.---------------------------------

The review of conditions adverse to quality is governed by SO-R-2, Condition Reporting and Corrective Action in conjunction with FCSG-24, Corrective Action Program Guideline. This procedure provides the necessary controls to ensure that all "conditions" are reviewed and assessed for operability/functionality and reportability (both verbal and written). The Condition Reporting System is computer-based and provides for a required review by the Shift Technical Advisor to ensure that operability/functionality and reportability are evaluated. Revisions to SO-R-2 are performed in accordance with SO-G-30, Procedure Changes and Generation.

The requirement to perform a "safety assessmenf' for "conditions" is antiquated in that risk assessments are now performed for "conditions" due to changes to 10 CFR 50.59 and 10 CFR 50.65. The Condition Reporting program (Le., SO-R-2 & FCSG-24) addresses the intent of the oriainal commitment and therefore, this commitment is no lonaer necessarv.

In the previous 50.59/USAR update (LlC-10-0040) it was noted that a revision to AR 10563 had been made to allow either a caution statement in the ST (as originally committed) or a new form (FC-68P, Pending Important Change) be used to flag ST acceptance criteria that were less conservative than an SAO. However, compliance with the original commitment has been reinstated such that a caution statement and not a form is used to flag ST acceptance criteria that are less conservative than an SAO.

The first is that the identification of human performance problems is addressed through the RCA program. The second is that management reviews or problem reports will focus on ensuring that root causes are accurately identified. In accordance with NEI 99-04, the definition of a commitment has changed from one of process orientation to one of outcomes orientation. Therefore, the method (Le., process) used by a licensee to restore compliance will usually not be considered a commitment. As a result, with the exception of the two commitments noted above, the portion of Safety Enhancement Program No. 15 tracked by AR 14040 has been deleted as it described the orocess for achievina comoliance.

This commitment was made in letter LlC-08-0044 dated April 10, 2008. The commitment was to implement and maintain the newly created HEPA filter replacement criteria in Procedure PE-RR-VA-0209 unless superseded by other criteria via the license amendment process. With the approval of Amendment No. 260, which instituted a new SR to replace the HEPA filters at intervals not to exceed 10 Years, this commitment is deleted.

This commitment was made in letter LlC-90-0419 dated May 9, 1990 in which OPPD stated:

Except for short (72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or less) maintenance outages with appropriate precautionary measures, the system will be insulated during the period of October through April to preclude rapid temperature changes causing boric acid precipitation. During the period of MaYI through September maintenance work may require the long term removal of insulation, however, the ambient daily average temperatures are sufficient that prolonged temperature changes be/ow 43OF are a low probability.

The commitment is deleted. The requirements of TS 2.2 and TS 3.1, Table 3-2, Item 16a are to ensure that the BAST temperature is sufficient to ensure that the solubility requirements of TS Figure 2-12 are maintained. Temperatures are monitored and documented on OP-ST-SHIFT-0001.

Procedures OI-VA-1 and OI-EW-1 also ensure that eauioment ooerabilitv is not affected bvextreme AR 10563 AR 42092 AR 43475 1-1-A-cR-1-4-0-40-----+fThiS commitment was reviewed and it was determined that there are only two (2) actual commitments in this AR.

LlC-12-0076 Page 6 AR 16848 Commitments Revised in Accordance with NEI 99-04 This commitment was made in letter LlC-95-0174 dated September 22,1995 in which OPPD states: The Control Room Operators required to verbalize pertinent label information, as an additional self-checking technique, prior to operating equipment from the Control Room.

commitment is deleted. Per NEI-99-04, Guideline for Managing NRC Commitment Changes, a licensee may change or eliminate commitment that has been determined to be unnecessary due to having been subsequently captured as part of an ongoing izing self-checking at Fort Calhoun Station has become culturally engrained due to continual reinforcement during training and requirement to perform a verbalization during self-checking is institutionalized (both procedurally and culturally) and audited intprnj:lflv j:lnn P1ltprnj:lfl\\l\\ Therefore. this commitment is no Ion program or other administrative control that is subject to a revision review process.

The use of Human Performance Tools at Fort Calhoun Station is governed by FCSG-7, Human Performance. This document is a reflection of the industry best practices as determined by INPO and incorporates most items from the Human Performance Reference Manual, INPO 06-003. FCSG-7 and INPO 06-003 both utilize verbalization during a self-check as the behavior standard.

Programmatically, Fort Calhoun Station Guidelines (FCSGs) are reviewed by and either approved or rejected by the Plant Manager.

Human Performance tool use is assessed by INPO during assist visits and during the biannual INPO Plant Evaluation. The use

LlC-12-0076 Page 1 Summary of Technical Specification Basis Changes (TSBC)

--- -- ~.---------------------

LlC-12-0076 Page 2 Technical Specification (TS) Bases Sections for TS Sections 2.0.1 and 2.7 were revised as part of Amendment No. 264 10-002-0 implementation.

Amendment No. 264 deleted TS limiting condition for operation (LCO) 2.0.1 (2), resulting in the 2.0 - Page 2, 3 associated Bases Section information being deleted and LCO 2.0.1 (3) being renumbered so the associated Bases 2.7 - Page 7, B, 9 information was renumbered accordingly. In addition, TS LCO 2.7 was revised to incorporate guidance that was deleted

[09-09-10]

from 2.0.1 (2) resulting in additional guidance being added to the TS Basis Section for 2.7.

USAR Section 2.7, 2.11, 9.B, USAR Figure 9.B - 1 & Technical Specification 2.16 Basis were updated to clarify the 10-003-0 protection required at various flood levels.

The USAR and Technical Specification Basis changes were made in 2.16-Page 1 conjunction with maintenance procedure changes which administratively implement new flood barriers. The new flood

[OB-1 0-1 0]

barriers are being implemented to enhance the ability to protect vital structures to a flood elevation of 1014 feet. A separate 50.59 screen has been prepared to address the procedure changes. EC 4950B(MCC), Implement Catastrophic Floodgates for the Intake Structure, and EC 49739 (MCC), Implement Catastrophic Floodgates for the Auxiliary Building, were completed to incorporate the new flood barriers into the permanent plant configuration. These ECs permanently installed anchors and components to expedite the installation of the new flood barriers in the event of a flood. The included USAR and TS Basis chances reflect the as-built desian of EC 4950B and EC 49739.

A statement in the basis for Technical Specification 2.1.6 was removed. The statement "Analysis of loss of load case 10-004-0 involving elevated PSV opening pressures indicated that RCS pressures remained below the 2750 psia Safety Limit with 2.1 - Page 21 PSV opening pressures up to 6% above nominal setpoints* is no longer being followed and is no longer required. This

[10-19-10]

change was performed to address the issue identified in CR 2010-1955.

Added clarifying statement in Bases of Technical Specification 2.6 that locked closed manual containment isolation 10-005-0 valves inside containment cannot be opened under administrative control when containment integrity is required.

2.6-Page 3 Clarification was added as a corrective action from CR 2010-1664. Clarifying step is verbatim from Standing Order 0-44

[09-30-10]

that was incorporated into procedure as part of the implementation of TS Amendment 151, which implemented the administrative controls.

LlC-12-0076 Page 3 11-001-0 Canceled CR 2011-0125 identified an error in TDB-VIII, Equipment Operability Guidance, related to auxiliary feedwater pump FW-6 inoperable concurrent with diesel generator DG-2 inoperable, and the correct Technical Specification to enter. As a 11-002-0 2.7 - Page 7 result, changes were made to USAR Sections 8.3.1.2 and 8.4.1.1 to provide clarification re: FW-10 not requiring AC

[10-25-11]

power and it being the redundant pump to FW-6 which is powered by bus 1A3. The TS Basis Section for TS 2.7 was also revised to add a clarification statement with FW-6 & DG-2 inoperable, coincident with a single failure of house service transformer T1A-4, would not result in a complete loss of a safety function since FW-10 would still be operable.

These changes are administrative clarifications as defined in FCSG-23.

The Technical Specification Bases were changed to remove the specific time-to-boil times listed for certain conditions.

11-003-0 This is considered extraneous information and does not need to be listed in the technical specifications bases. One 2.8 - Page 18 specific time point does not change the fact that there is a low probability of the time-to-boil being greater than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

[04-28-11]

with the refueling cavity flooded 23 feet above the core. Many cases show that the time-to-boil with the reactor cavity filled to greater than 23 feet will have a time-to-boil greater than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The discussion in the technical specification bases about the time-to-boil at a temperature of 200 0 F is greater than 25 minutes is no longer true. This was changed due to errors in the analysis that provided that basis. However, this 25-minute time does not provide a basis for this technical specification and does not need to be listed.

11-004-0 Canceled 11-005-0 Canceled

LlC-12-0076 Page 4 DescriptiQTI 11-006-0 2.15-Page6 3.1 - Page 2

[02-24-12]

12-001-0 2.10 - Page 18

[04-12-12]

The Bases of TS 2.15 and TS 3.1 are revised to incorporate changes approved by the NRC in Amendment No. 267. The changes note that when the CEAs are fully inserted or fully withdrawn, DCS core mimic can be used to verify primary CEAPIS data.

However, since DCS core mimic indication is not fully independent of secondary CEAPIS, primary CEAPIS is still required to verify secondary CEAPIS. CEA positions are now required to be verified each shift and within 15 minutes of CEA motion when a CEAPIS channel is inoperable.

The Basis of Technical Specification 2.10.4 contained typographical errors regarding where the CEA insertion limits are located.

Prior to Amendment No. 117, these limits were correctly identified as TS 2.10.2(6) and TS 2.10.2(7).

Amendment No. 117 incorrectly changed the designations to TS 2.10.1(6) and TS 2.10.1(7), which do not exist. TSBC 12-001-0 corrected the designations to be as they were prior to Amendment No. 117.

LlC-12-0076 Page 1 Technical Specification Basis Change (TSBC) Pages

TECHNICAl SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.0.1 General Requirements (Continued)

(1 )

This specification delineates corrective measures to be taken for circumstances.

not directly provided for in the system specific specifications and whose occurrence would violate the intent of the specification. For example, Specification 2.3 requires each Low Pressure Safety Injection (lPSI) pump to be operable and provides explicit corrective measures to be followed if one pump is inoperable.

Under the terms of Specification 2.0.1 (1), if more than one lPSI pump is inoperable, the unit must be placed in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , in at least subcritical and < 300°F within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in at least COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , unless at least one lPSI pump were restored to operability. It is assumed that the unit is brought to the required mode within the required times by promptly initiating and carrying out the appropriate measures, required by the specification.

I (2) lCO 2.0.1 (2) establishes conditions under which systems are considered to I

remain capable of performing their intended safety function when associated snubbers are not capable of providing their associated support function(s). This lCO states that the supported system is not considered to be inoperable solely due to one or more snubbers not capable of performing their associated support function(s). This is appropriate because a limited length of time is allowed for maintenance, testing, or repair of one or more snubbers not capable of performing their associated support function(s) and appropriate compensatory measures are specified in the snubber requirements, which are located outside of the Technical Specifications (TS) under licensee control. The snubber requirements do not meet the criteria in 10 CFR 50.36( c)(2)(ii), and. as such, are appropriate for control by the licensee.

If the allowed time expires and the snubber(s) are unable to perform their associated support function(s), the affected supported system's lCO(s) must be declared not met and the Conditions and Required Actions entered.

lCO 2.0.1 (2)a applies when one or more snubbers are not capable of providing I

their associated support function(s) to a single train or subsystem of a multiple train or subsystem supported system or to a single train or subsystem supported system. lCO 2.0.1 (2)a allows 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore the snubber(s) before declaring I the supported system inoperable. The 72-hour Completion Time is reasonable based on the low probability of a seismic event concurrent with an event that would require operation of the supported system occurring while the snubber(s) are not capable of performing their associated support function and due to the availability of the redundant train of the supported system.

2.0 - Page 2 Amendment No. 52, 238, 264 TSBC-10-002-0

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.0.1 General Requirements (Continued)

LCO 2.0.1 (2)b applies when one or more snubbers are not capable of providing I

their associated support function(s) to more than one train or subsystem of a multiple train or subsystem supported system. LCO 2.0.1 (2)b allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to I restore the snubber(s) before declaring the supported system inoperable. The 12-hour Completion Time is reasonable based on the low probability of a seismic event concurrent with an event that would require operation of the supported system occurring while the snubber(s) are not capable of performing their associated support function.

LCO 2.0.1 (2) requires that risk be assessed and managed. Industry and NRC I

guidance on the implementation of 10 CFR 50.65(a)(4) (the Maintenance Rule) does not address seismic risk. However, use of LCO 2.0.1 (2) should be I

considered with respect to other plant maintenance activities, and integrated into the existing Maintenance Rule process to the extent possible so that maintenance on any unaffected train or subsystem is properly controlled, and emergent issues are properly addressed. The risk assessment need not be quantified, but may be a qualitative awareness of the vulnerability of systems and components when one or more snubbers are not able to perform their associated support function.

2.0 - Page 3 Amendment No. ~ 264 TSBC-10-002-0

TECHNICAl SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (continued) 2.1.6 Pressurizer and Main Steam Safety Valves (continued)

d. With both PORVs inoperable in Modes 4 or 5, depressurize and vent the RCS through at least a 0.94 square inch or larger vent within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

(5)

Two power-operated relief valves (PORVs) and their associated block valves shall be operable in Modes 1, 2, and 3.

a. With one or both PORV(s) inoperable because of excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to operable status or close the associated block valve(s) with power maintained to the block valve(s); otherwise, be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

b. With one PORV inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV to operable status or close its associated block valve and remove I power from the block valve; restore the PORV to operable status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

c. With both PORVs inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore at least one PORV to operable status or close both block valves, remove power from the block valves, and be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

d. With one or both block valve(s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore the block valve(s) to operable status or place the associated PORV(s) in the closed position. Restore at least one block valve to operable status within the next hour if both block valves are inoperable; restore the remaining inoperable block valve to operable within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Otherwise, be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Basis The purpose of the two spring-loaded Pressurizer Safety Valves (PSV's) is to provide Reactor Coolant System (RCS) overpressure protection and thereby ensure that the Safety Limit for RCS pressure (i.e., 2750 psia) is not exceeded for analyzed accidents. The maximum RCS pressure transient for an analyzed accident is associated with a Loss of Load event(2).

The TS 2.1.6(1) lift settings are determined during Surveillance Testing in accordance with ASME Code test methods. The ASME Code requires that valves in steam service use steam as the test medium for establishing the setpoint. The +1 %/-3% tolerance range speCified in TS 2.1.6(1) applies to opening pressures determined during Surveillance Testing. When the valves are installed in the system, the presence of a water-filled loop seal at the valve inlets may result in in-situ actuation at a pressure that differs from the actuation pressure with steam at the inlet.

Comparative testing and analysis indicates that with a loop seal present, the opening pressure of these valves may be up to 1 % lower than the opening pressure under normal test conditions.

Opening pressures below the specified setpoints are not a concern with respect to the safety limit for RCS pressure. The valves are set to a tolerance of +/-1 % of setpoint using ASME Code II test methods before being returned to service after testing. This allows for some setpoint variance over the surveillance interval.

2.1 - Page 21 Amendment No. 54,146,157,161,189,219,237 TSBC-10-004-0

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.6 Containment System (Continued)

The reactor coolant system conditions of cold shutdown assure that no steam will be formed and, hence, there would be no pressure buildup in the containment if the reactor coolant system ruptures. The shutdown margins are selected based on the type of activities that are being carried out.

Regarding internal pressure limitations, the containment design pressure of 60 psig would not be exceeded if the internal pressure before a major loss-of-coolant accident were as much as 3 psig.(1) The opening of locked or sealed closed containment isolation valves on an intermittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves (valve is accessible under accident conditions, e.g., not in Containment) and that this action will prevent the release of radioactivity outside the containment. Operation of the purge isolation valves is prevented during normal operations due to the size of the valves (42 inches) and a concern about their ability to close against the differential pressure that could result from a LOCA or MSLB.

Specification 2.6(1)a applies when both doors of the PAL are declared inoperable, or the entire air lock assembly leakage exceeds the requirements of Specification 5.19.

Specification 2.6(1 )b(i) applies when a PAL door, or mechanisms attached thereto, such as the inner PAL door equalizing valve, are declared inoperable. If the inner PAL door equalizing valve is inoperable, a removable cap can be placed on the equalizing valve to provide containment isolation when necessary.

Specification 2.6(1 )b(ii) applies when mechanisms other than a door are declared inoperable.

When one PAL door is inoperable, the ability to open the operable door to perform repairs of the affected air lock components even if it means the containment boundary is not intact (during access through the outer door) is acceptable because of the low probability of an event that could pressurize the containment during the short time in which the operable door is expected to be open. After each entry and exit, the operable door must be locked immediately.

The containment integrity will be protected by ensuring the penetration valves VA-280 and VA-289 are "locked closed" while HCV-881 and HCV-882 are normally closed during power operation.

References (1)

USAR, Section 14.16; Figure 14.16-2 2.6 - Page 3 Amendment No. 138,151, 188 TSBC 06-003-0 TSBC 07-001-1 TSBC-10-005-0 I

I I

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.7 Electrical Systems (Continued)

One battery charger on each battery shall be operating so that the batteries will always be at full charge; this ensures that adequate d-c power will be available for all emergency uses. Each battery has one battery charger permanently connected with a third charger capable of being connected to either battery bus. The chargers are each rated for 400 amperes at 130 volts. Following a DBA the batteries and the chargers will handle all required loads. Each of the reactor protective channels instrumentation channels is supplied by one of the safety-related a-c instrument buses. The removal of one of the safety-related a-c instrument buses is permitted as the 2-of-4 logic may be manually changed to a 2-of-3 logic without compromising safety.

The engineered safeguards instrument channels use safety-related a-c instrument buses (one redundant bus for each channel) and d-c buses (one redundant bus for each logic circuit). The removal of one of the safety-related a-c instrument buses is permitted as the two of four logic automatically becomes a two of three logic.

The requirement in SpeCification 2.7(2)j, todeclare required redundant feature(s) inoperable, is intended to provide assurance that a loss of offsite power, during the period that a DG is inoperable, does not result in a complete loss of safety function of critical systems. These features are designed with redundant safety related components.

Redundant required feature failures consist of inoperable features with a component redundant to the component that has an inoperable DG. The steam driven auxiliary feedwater pump FW-10 is required to be considered a redundant required feature to motor driven auxiliary feedwater pump FW-6, and, is therefore, required to be determined OPERABLE, since there are only two safety-related AFW pumps. With FW-10 and DG-1 INOPERABLE, coincident with a single failure of house service transformer T1A-3, would result in a complete loss of a safety function. With FW-6 and DG-2 INOPERABLE, I

coincident with a single failure of house service transformer T1A-4, would not result in a complete loss of a safety function since FW-10 would still be OPERABLE.

Redundant required features for an inoperable DG do not include components powered from 125 VDC or 120 VAC sources, since a loss of function would not occur with an inoperable DG coincident with a single failure of its associated house service transformer.

Radiation Monitors RM-051, RM-052, and RM-062 are required to be considered redundant features since the monitors are contained on a skid assembly which is powered from 4BOVAC.

2.7 - Page 7 Amendment No. ~, 251 TSBC-OB-OO1-0 TSBC-10-002-0 TSBC-11-002-0

TECHNICAl SPEC IFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.7 Electrical Systems (Continued)

Basis (continued)

The time allowed for declaring a redundant required feature(s) inoperable is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This also allows for an exception to the normal beginning for the limiting condition for operation time. In this required action, the time only begins upon discovery that both:

a. An inoperable DG exists and

b. A required feature associated with the other 4160V bus is inoperable.

If at any time during the existence of this Condition (one DG inoperable) a required feature subsequently becomes inoperable, this time begins to be tracked. Discovering one required DG inoperable coincident with one or more inoperable required support or supported features, or both, that are associated with the OPERABLE DG, results in starting the time for the required action.

Four hours from the discovery of these events existing concurrently, is acceptable because it minimizes risk while allowing time for restoration before subjecting the unit to transients associated with shutdown.

In this modified Condition (one DG inoperable and loss of required component on the oppOSite DG),

the remaining OPERABLE DG and offsite circuits are adequate to supply electrical power to the onsite electrical distribution system. Thus, on a component basis, single failure protection for the required feature's function may have been lost; however, the function has not been lost. The 4-hour allowed time takes into account the operability of the redundant counterpart to the inoperable required feature. Additionally, the 4-hour allowed time takes into account the capacity and capability of the remaining AC sources, a reasonable time for repairs, and the low probability of a DBA occurring during this period.

When a system has installed spare components, the spare component is not required to be OPERABLE to meet required feature operability. As an example, there are three installed 100%

capacity high pressure safety injection (HPSI) pumps, one (SI-2B) associated with 4160V bus 1A4, and two (SI-2A and SI-2C) associated with 4160V bus 1A3. Specification 2.3(1) Minimum Requirements are that there be one HPSI pump on each associated 4160V bus and each safety injection refueling water tank-containment sump header. This requires that SI-2A OR SI-2C be OPERABLE, not both.

The DG lubrication system is designed to provide sufficient lubrication to permit proper operation of its associated DG under all loading conditions. The system is required to circulate the lube oil to the diesel engine working surfaces and to remove excess heat generated by friction during operation.

The onsite storage in addition to the engine oil sump is sufficient to ensure 7 days of continuous operation. This supply is sufficient supply to allow the operator to replenish lube oil from outside sources. With lube oil inventory < 500 gallons, sufficient lubricating oil to support 7 days of continuous DG operation at full load conditions may not be available. However, the Condition is restricted to lube oil volume reductions that maintain at least a 6 day supply. This restriction allows sufficient time to obtain the requisite replacement volume. A period of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is considered sufficient to complete restoration of the required volume prior to declaring the DG inoperable. This period is acceptable based on the remaining capacity (> 6 days), the low rate of usage, the fact that procedures will be initiated to obtain replenishment, and the low probability of an event during this brief period.

2.7 - Page 8 Amendment No. ~. 251 TSBC-08-001*0 TSBC-10-002-0

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueling Bases (Continued) 2.8.1 (3)

Shutdown Cooling System - High Water Level (Continued)

Specification 2.8.1 (3) is modified by an exception to allow both trains of SOC out of service for up to eight hours provided, in part, that at least one SOC train is available under administrative controls. This allows evolutions such as Engineered Safety Feature testing to be completed when the SOC system is not fully OPERABLE but is considered available since only minor operator actions are required to restore the SOC system to OPERABLE status and place it IN OPERATION. A SOC loop is considered available under administrative controls if there are: (1) approved procedures, (2) a dedicated operator stationed at the controls if they are outside of the control room, and (3) direct communication between the dedicated operator and the control room. Similarly, the SOC system is considered available under administrative controls when an operator is not at the location of the controls provided: (1) procedural guidance is consulted prior to removing SOC from service to determine the time-to-boil, and (2) there is sufficient time for the operator to travel to the local controls and perform the required actions.

With the water level ~ 23 feet above the top of the core, only one SOC loop is required for decay heat removal. Only one is required because the volume of water above the top of the core provides backup decay heat removal capability.

The 23 ft level was selected because it ensures that adequate time is available to initiate emergency procedures to cool the core.

I If the SOC loop requirements are not met, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Therefore, actions that reduce boron concentration are required to be suspended immediately.

Additionally, suspending any operation that would increase the decay heat load, such as loading a fuel assembly, is a prudent action under this condition. Closing the containment penetrations that provide direct access to the outside environment prevents fission products, if released from a loss of decay heat removal event, from escaping the containment. A completion time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is reasonable because most SOC problems can be repaired within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and because there is a low probability of the cooling boiling in that time.

When "immediately" is used as a completion time, the required action should be pursued without delay and in a controlled manner.

2.8 - Page 18 Amendment No..:J-3S, 239 TSBC-11-003-0

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) 2.10A Power Distribution Limits (Continued)

(5)

DNBR Margin During Power Operation Above 15% of Rated Power (a) The following limits on DNB-related parameters shall be maintained:

(i)

Cold Leg Temperature as specified in the COLR (Core Inlet Temperature)

(ii)

Pressurizer Pressure

~ 2075 pSia(1)

(iii)

Reactor Coolant Flow rate

~ 202,500 gpm indicated (iv)

Axial Shape Index as specified in the COLR (b) With any of the above parameters exceeding the limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce power to less than 15% of rated power within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200°F.

Either of the two core power distribution monitoring systems, the Excore Detector Monitoring System or the Incore Detector Monitoring System, provides adequate monitoring of the core power distribution and is capable of verifying that the linear heat rate does not exceed its Iirnit. The Excore Detector Monitoring System periorms this function by continuously monitoring the axial shape index (ASI) with the operable quadrant symmetric excore neutron flux detectors. The axial shape index is maintained within the allowable limits of the Limiting Condition for Operation for Excore Monitoring of LHR Figure provided in the COLR. This ASI is adjusted by Specification 2.10A(1)(c) for the allowed linear heat rate of the Allowable Peak Linear Heat Rate vs.

Burnup Figure provided in the COLR and the FRT and Core Power Limitations Figure provided in the COLR. In conjunction with the use of the excore monitoring system and in establishing the axial shape index limits, the following assumptions are made:

(1) the CEA insertion limits of Specification 2.10.2(6) and long term insertion limits of I Specification 2.10.2(7) are satisfied, and (2) the flux peaking augmentation factors are as shown in Figure 2-8.

(1)

Limit not applicable during either a thermal power ramp in excess of 5% of rated thermal power per minute or a thermal power step of greater than 10% of rated thermal power.

SEE TDB-VIII 2.10 - Page 18 Amendment No. 32,43,57,70, 77,92,109,117,141,156, 193,196,209,249 TSBC 12-001-0 I

~-----------------

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERArlON

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 ofTS 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 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 (AI-185), 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).

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.

SEE TOB-VIII 2.15 - Page 6 Amendment No. ~. 249 I TSBC-04-00 1-0 TSBC-08-003-0 TSBC 08-008-0 TSBC-11-006-0

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.16 River Level Applicability Applied to Missouri River level as measured at the intake structure at the Fort Calhoun Station.

Objective To specify maximum and minimum Missouri River levels which must be present to assure safe reactor operation.

Specifications

1)

If the Missouri River level exceeds 1009(1) feet the reactor will be placed in a cold shutdown condition using normal operating procedures. When the river level reaches elevation 1004.2 feet and rising, the emergency plan to protect the plant will be instituted.

(2)

If the Missouri River level is less than 976 feet 9 inches the reactor will be placed in a cold shutdown condition using normal operating procedures. At river levels less than 980 feet a continuous watch will be maintained to assure no sudden loss of water supply occurs.

At the Fort Calhoun Station (FCS) site, the probable maximum flood that might occur as a result of runoff from a probable maximum rainstorm over the area below the Gavins Point dam coupled with an assumed outflow of 50,000 cubic feet per second from Gavins Point reservoir is 1009.3 feet. In the unlikely event that the Oahe or Fort Randall dams fail at that time, the Corps of Engineers has estimated that the flood level could be as high as 1014 feet(1).

The intake structure can be protected from these Missouri River floods using removable flood gates on doorways and the screen wash discharge trough. The water level inside the intake cells can be controlled by positioning the exterior sluice gates to restrict the flow into the cells.

The auxiliary building can be protected to 1009.5 feet using its installed flood gates. Protection of the auxiliary building to 1014 feet requires the installation of removable flood barriers and sandbagging at the 1013 foot elevation of the equipment hatch room (Room 66).

The minimum river level of 976 feet 9 inches provides adequate suction to the raw water (RW) pumps for cooling plant components. The minimum elevation of the RW pump suction is 973 feet 9 inches. An intake cell level of 976 feet 9 inches is required for RW pump minimum submergence level (MSL)(2). The partial loss of this supply is considered highly unlikely.

However, provisions for low water levels during winter and spring ice conditions are considered necessary. When river level is low, head loss from debris and/or ice on the traveling screens and/or trash racks could reduce intake cel/levels such that the required RW pump MSL is not achieved. This could lead to pump degradation from the formation of vortices at the free water surface. Thus, when the continuous watch requirement is in effect, in addition to river level, the level of the intake cells is monitored.'"

See TOB-VIII 2.16 - Page 1 TSBC-07 -002-0 TSBC-10-001-0 TSBC-10-003-0

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.1 Instrumentation and Control (Continued)

Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures.

The minimum calibration frequencies of once-per-day (heat balance adjustment only) for the power range safety channels, and once each refueling shutdown for the process system channels, are considered adequate.

The minimum testing frequency for those instrument channels connected to the Reactor Protective System and Engineered Safety Features is based on ABB/CE probabilistic risk analyses and the accumulation of specific operating history. The quarterly frequency for the channel functional tests for these systems is based on the analyses presented in the NRC approved topical report CEN-327 -A, "RPS/ESFAS Extended Test Interval Evaluation," as supplemented, and OPPD's Engineering Analysis EA-FC-93-064, "RPS/ESF Functional Test Drift Analysis."

The low temperature setpoint power operated relief valve (PORV) CHANNEL FUNCTIONAL TEST verifies operability of the actuation circuitry using the installed test switches. PORV actuation could depressurize the reactor coolant system and is not required.

OPERABILITY of two CEA position indication system (CEAPIS) channels is required to determine CEA positions, and thereby ensure compliance with the CEA alignment and insertion limits. Limit switches on the regulating CEAs and reed switches on the shutdown CEAs provide an additional means of determining CEA positions when the CEAs are at either their fully inserted or fully withdrawn positions. This indication displayed on the distributed control system (DCS) flat-panel touch monitors is known as DCS core mimic.

Performance of a CHANNEL CHECK on the primary CEAPIS channel includes comparison with the secondary CEAPIS channel and DCS core mimic (when available). If secondary CEAPIS is inoperable, DCS core mimic alone can verify primary CEAPIS, however, because DCS core mimic is not fully independent of secondary CEAPIS, it can not be used to verify secondary CEAPIS data when primary CEAPIS is inoperable. The primary CEAPIS channel must be OPERABLE in order to verify secondary CEAPIS data. The frequency of each shift takes into consideration other information continuously available to the operator in the control room, so that during CEA movement deviations can be detected.

Verification that individual CEA positions are within 12 inches of all other CEAs in the group at a frequency of each shift allows the operator to detect a CEA that is beginning to deviate from its expected position.

The specified frequency takes into account other CEA position information that is continuously available to the operator in the control room, so that during CEA movement, deviations can be detected. Protection is also provided by the CEA deviation alarm.

Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is calibrated.

Calculation of the Reactor Coolant System (RCS) total flow rate by performance of a precision calorimetric heat balance once every 18 months verifies that the actual RCS flow rate is greater than or equal to the minimum required RCS flow rate (Table 3-3, Item 15, Reactor Coolant Flow).

The frequency of 18 months reflects the importance of verifying flow after a refueling outage when the core has been altered, Steam Generator tubes plugged or repaired, or other activities, which may have caused an alteration of flow resistance.

This requirement is modified by a footnote that requires the surveillance to be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after ~ 95% reactor thermal power (RTP) following power escalation from a refueling outage. The footnote is necessary to allow measurement of the flow rate at normal operating conditions at power in MODE 1.

3.1 - Page 2 Amendment No. 163,182, 193, 195 TSBC-08-008-0 TSBC-11-006-0

LlC-12-0076 Page 1 List of Files on CD-ROM

LIC-12-0076 Page 2 File Name Size Sensitivity Level Location 001 LIC-12-0076.pdf 2.98 MB Publicly Available CD-ROM 002 USAR Index.pdf 116 KB Publicly Available CD-ROM 003 USAR 01-01.pdf 28 KB Publicly Available CD-ROM 004 USAR 01-02.pdf 85 KB Publicly Available CD-ROM 005 USAR 01-03.pdf 36 KB Publicly Available CD-ROM 006 USAR 01-04.pdf 113 KB Publicly Available CD-ROM 007 USAR 01-05.pdf 50 KB Publicly Available CD-ROM 008 USAR 01-06.pdf 31 KB Publicly Available CD-ROM 009 USAR 01-07.pdf 17 KB Publicly Available CD-ROM 010 USAR 01-08.pdf 17 KB Publicly Available CD-ROM 011USAR 01-09.pdf 84 KB Publicly Available CD-ROM 012 USAR 01-10.pdf 13 KB Publicly Available CD-ROM 013 USAR 01-11.pdf 28 KB Publicly Available CD-ROM 014 USAR 01-12.pdf 27 KB Publicly Available CD-ROM 015 USAR 02-01.pdf 14 KB Publicly Available CD-ROM 016 USAR 02-02.pdf 31 KB Publicly Available CD-ROM 017 USAR 02-03.pdf 14 KB Publicly Available CD-ROM 018 USAR 02-04.pdf 33 KB Publicly Available CD-ROM 019 USAR 02-05.pdf 276 KB Publicly Available CD-ROM 020 USAR 02-06.pdf 33 KB Publicly Available CD-ROM 021 USAR 02-07.pdf 88 KB Publicly Available CD-ROM 022 USAR 02-08.pdf 39 KB Publicly Available CD-ROM 023 USAR 02-09.pdf 37 KB Publicly Available CD-ROM 024 USAR 02-10.pdf 62 KB Publicly Available CD-ROM 025 USAR 02-11.pdf 16 KB Publicly Available CD-ROM 026 USAR 03-01.pdf 34 KB Publicly Available CD-ROM 027 USAR 03-02.pdf 79 KB Publicly Available CD-ROM 028 USAR 03-03.pdf 14 KB Publicly Available CD-ROM 029 USAR 03-04.pdf 177 KB Publicly Available CD-ROM 030 USAR 03-05.pdf 57 KB Publicly Available CD-ROM 031 USAR 03-06.pdf 85 KB Publicly Available CD-ROM 032 USAR 03-07.pdf 76 KB Publicly Available CD-ROM 033 USAR 03-08.pdf 98 KB Publicly Available CD-ROM 034 USAR 03-09.pdf 40KB Publicly Available CD-ROM 035 USAR 03-10.pdf 14 KB Publicly Available CD-ROM 036 USAR 04-01.pdf 28 KB Publicly Available CD-ROM 037 USAR 04-02.pdf 88 KB Publicly Available CD-ROM 038 USAR 04-03.pdf 176 KB Publicly Available CD-ROM 039 USAR 04-04.pdf 37 KB Publicly Available CD-ROM 040 USAR04-05.pdf 168 KB Publicly Available CD-ROM 041 USAR 04-06.pdf 54 KB Publicly Available CD-ROM 042 USAR 04-07.pdf 29 KB Publicly Available CD-ROM 043 USAR 05-01.pdf 32 KB Publicly Available CD-ROM

LlC-12-0076 Page 3 File Name Size 044 USAR 05-02.pdf 045 USAR 05-03.pdf 046 USAR 05-04.pdf 047 USAR 05-05.pdf 048 USAR 05-06.pdf 049 USAR 05-07.pdf 050 USAR 05-08.pdf 051 USAR 05-09.pdf 052 USAR 05-10.pdf 053 USAR 05-11.pdf 054 USAR 05-12.pdf 055 USAR 05-13.pdf 056 USAR 06-01.pdf 057 USAR 06-02.pdf 058 USAR 06-03.pdf 059 USAR 06-04.pdf 060 USAR 06-05.pdf 061 USAR 06-06.pdf 062 USAR 07 -01.pdf 063 USAR 07 -02.pdf 064 USAR 07 -03.pdf 065 USAR 07 -04.pdf 066 USAR 07 -05.pdf 067 USAR 07 -06.pdf 068 USAR 07-07.pdf 069 USAR 08-01.pdf 070 USAR 08-02.pdf 071 USAR 08-03.pdf 072 USAR 08-04.pdf 073 USAR 08-05.pdf 074 USAR 08-06.pdf 075 USAR 08-07.pdf 076 USAR 09-01.pdf 077 USAR 09-02.pdf 078 USAR 09-03.pdf 079 USAR 09-04.pdf 080 USAR 09-05.pdf 081 USAR 09-06.pdf 082 USAR 09-07.pdf 083 USAR 09-08.pdf 084 USAR 09-09.pdf 085 USAR 09-10.pdf 086 USAR 09-11. pdf 46 KB 49KB 35 KB 121 KB 71 KB 88 KB 76 KB 69KB 60 KB 86 KB 15 KB 14 KB 39 KB 96 KB 40 KB 74 KB 28 KB 29KB 30 KB 169 KB 93 KB 81 KB 89KB 53 KB 28KB 35 KB 36 KB 49KB 57 KB 46 KB 14 KB 26 KB 31 KB 136 KB 45 KB 65 KB 69 KB 42KB 87 KB 88 KB 30 KB 111 KB 138 KB Sensitivity Level Location Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM

LlC-12-0076 Page 4 File Name Size Sensitivity Level Location 087 USAR 09-12.pdf 088 USAR 09-13.pdf 089 USAR 10-01.pdf 090 USAR 10-02. pdf 091 USAR 10-03.pdf 092 USAR 10-04.pdf 093 USAR 10-05.pdf 094 USAR 10-06.pdf 095 USAR 11-01.pdf 096 USAR 11-02.pdf 097 USAR 11-03.pdf 098 USAR 11-04.pdf 099 USAR 11-05.pdf 100 USAR 12-01.pdf

. 101 USAR 12-02.pdf 102 USAR 12-03.pdf 103 USAR 12-04.pdf 104 USAR 12-05.pdf 105 USAR 12-06.pdf 106 USAR 12-07.pdf 107 USAR 12-08.pdf 108 USAR 13-01.pdf 109 USAR 13-02.pdf 110 USAR 13-03.pdf 111 USAR 13-04.pdf 112 USAR 13-05.pdf 113 USAR 14-01.pdf 114 USAR 14-02.pdf 115 USAR 14-03.pdf 116 USAR 14-04.pdf 117 USAR 14-05.pdf 118 USAR 14-06.pdf 119 USAR 14-07.pdf 120 USAR 14-08.pdf 121 USAR 14-09.pdf 122 USAR 14-10.pdf 123 USAR 14-11.pdf 124 USAR 14-12.pdf 125 USAR 14-13.pdf 126 USAR 14-14.pdf 127 USAR 14-15.pdf 128 USAR 14-16.pdf 129 USAR 14-17.pdf 48 KB 79 KB 28 KB 42 KB 30 KB 13 KB 12 KB 12 KB 348 KB 120 KB 88 KB 29KB 27 KB 32 KB 40 KB 30KB 27 KB 30 KB 27 KB 27 KB 14 KB 30 KB 28 KB 27 KB 28 KB 27 KB 117 KB 60 KB 118 KB 48 KB 30 KB 109 KB 31 KB 90 KB 103 KB 102 KB 47 KB 118 KB 98 KB 84 KB 164 KB 122 KB 30 KB Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM

LlC-12-0076 Page 5 File Name Size 130 USAR 14-18.pdf 131 USAR 14-19.pdf 132 USAR 14-20.pdf 133 USAR 14-21.pdf 134 USAR 14-22.pdf 135 USAR 14-23.pdf 136 USAR 14-24.pdf 137 USAR 15-01.pdf 138 USAR 15-02.pdf 139 USAR 15-03.pdf 140 USAR 15-04.pdf 141 USAR Appendix A.pdf 142 USAR Appendix B.pdf 143 USAR Appendix C.pdf 144 USAR Appendix O.pdf 145 USAR Appendix E.pdf 146 USAR Appendix F.pdf 147 USAR Appendix G.pdf 148 USAR Appendix H.pdf 149 USAR Appendix I.pdf 150 USAR Appendix J.pdf 151 USAR Appendix K.pdf 152 USAR Appendix Lpdf 153 USAR Appendix M.pdf 154 USAR Appendix N.pdf 155 USAR Fig. 12-1-1.pdf 156 USAR Fig. 12-1-2.pdf 157 USAR Fig. 12-1-3.pdf 158 USAR Fig. 12-1-4.pdf 159 USAR Figures Append ix-A. pdf 160 USAR Figures Appendix-F.pdf 161 USAR Figures Appendix-I.pdf 162 USAR Figures Appendix-M.pdf 163 USAR Figures Section-01.pdf 164 USAR Figures Section-02.pdf 165 USAR Figures Section-03.pdf 166 USAR Figures Section-04.pdf 167 USAR Figures Section-05.pdf 168 USAR Figures Section-06.pdf 169 USAR Figures Section-07.pdf 170 USAR Figures Section-08.pdf 171 USAR Figures Section-09.pdf 172 USAR Figures Section-11. pdf 91 KB 36 KB 41 KB 27 KB 81 KB 51 KB 101 KB 12KB 56 KB 50KB 64KB 124 KB 19 KB 9.367 KB 9.228 KB 12 KB 172 KB 243 KB 248 KB 139 KB 12 KB 12 KB 13 KB 148 KB 71 KB 15 KB 11 KB 11 KB 14 KB 33 KB 2,343 KB 103 KB 386 KB 1,940 KB 1,897 KB 662KB 756 KB 6,050 KB 519 KB 13,672 KB 854 KB 1,472 KB 44KB Sensitivity Level Location Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM Publicly Available CD-ROM

LlC-12-0076 Page 6 File Name Size Sensitivity Level Location 173 USAR Figures Section-12.pdf 401 KB Publicly Available CD-ROM 174 USAR Figures Section-14.pdf 3,080 KB Publicly Available CD-ROM

L1C-12-0076 Enclosure CD-ROM of USAR Sections and Figures Fort Calhoun Station Updated Safety Analysis Report June 2012

ML12226A171

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