LIC-15-0119, ISFSI - Quality Assurance Program Approval for Radioactive Material Packages

ISFSI - Quality Assurance Program Approval for Radioactive Material Packages

ML15349A873
Person / Time
Site:	Fort Calhoun, 07100256
Issue date:	12/09/2015
From:	Cortopassi L Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards
References
LIC-15-0119
Download: ML15349A873 (65)

Text

Omaha Public Power District December 9, 2015 LIC-15-01 19 10 CFR 50.71(e) 10 CFR 50.4(b)(6) 10 CFR 50.54(a) 10 CFR 50.59 10 CFR 54.37(b) 10 CFR 71.106 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Fort Calhoun Station, Unit No. 1 Renewed Facility Operating License No. DPR-40 NRC Docket No. 50-285 Fort Calhoun Station Independent Spent Fuel Storage Installation NRC Docket No.72-054 Omaha Public Power District - Fort Calhoun Quality Assurance Program Approval for Radioactive Material Packages NRC Docket No. 71-0256

Reference:

10 CFR 50.59 Report, Quality Assurance (QA) Program Changes, Technical Specification Basis Changes, 10 CFR 71.106 Quality Assurance Program Approval, Aging Management Review, Commitment Revisions, and Revision of Updated Safety Analysis Report Revision for Fort Calhoun Station (FCS), Unit No. 1

1. Letter from OPPD (L. P. Cortopassi) 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 17, 2014 (MLE141 76A236) (LIC-1 4-0077)

In accordance with 10 CFR 50.59(d)(2), the Omaha Public Power District (OPPD) submits as the report of changes, tests, and experiments performed pursuant to 10 CFR 50.59 for Fort Calhoun Station (FCS), Unit No. 1. Attachment 2 is provided to describe Quality Assurance (QA) Program changes as required by 10 CFR 50.54(a)(4)(i).

describes changes made to the quality assurance program approval for radioactive material packages. Attachment 4 contains a description of revised regulatory commitments that require Commission notification in accordance with NEI 99-04, "Guidelines for Managing NRC Commitment Changes." In accordance with FCS Technical Specification 5.20.d, Attachment 5 provides a brief summary of the Technical Specification Basis Changes (TSBCs) made since the previous submittal (Reference 1) and Attachment 6 includes a copy of the revised TSBC.

pages.

444 SOUTH 16TH STREET MALL

• OMAHA, NE 68102-2247 I',4.NSS

U. S. Nuclear Regulatory Commission LIC-15-01 19 Page 2 In accordance with 10 CFR 54.37(b), a review of structures, systems, and components (SSCs) newly identified that would have been subject to an aging management review or evaluation of time-limited aging analysis in accordance with 10 CFR 54.21 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.

This information covers the period of June 14, 2014 through December 4, 2015.

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 1) and includes changes made under the provisions of 10 CFR 50.59 but not previously submitted to the Commission.

The CD-ROM also contains Revision 4 of the Quality Assurance Topical Report (NO-FC-1 0),

incorporated by reference in the USAR. Attachment 7 contains a list of the files on the CD-ROM.

The Senior Resident Inspector is provided with an updated copy of the USAR by the FCS distribution process.

As required by 10 CFR 50.71(e)(2)i, I certify that the information in this submittal accurately presents changes made since the previous submittal, necessary to reflect information and analyses submitted to the Commission or prepared pursuant to Commission requirements, and identifies changes made under the provisions of 10 CFR 50.59 but not previously submitted to the Commission.

No commitments to the NRC are made in this letter.

If you should have any questions, please contact Mr. Bill Hansher at (402) 533-6894.

Respectuly Louis P. Cortopassi Site Vice President and CNO LPC/MLE/mle Attachments: 1. Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59

2. Quality Assurance Program Changes

3.

10 CFR 71.106 Quality Assurance Program Approval for Radioactive Material Package Changes

4. Regulatory Commitments Revised in Accordance with NEI 99-04

5.

Information Removed from the USAR

6. Summary of Technical Specification Basis Changes (TSBC)

7. TSBC Pages

8. List of Files on CD-ROM

Enclosure:

CD-ROM of USAR Sections and Figures c:

M. L. Dapas, NRC Regional Administrator, Region IV C. F. Lyon, NRC Senior Project Manager S. M. Schneider, NRC Senior Resident Inspector (w/o Enclosure)

LIC-15-01 19 Page 1 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59

LIC-1 5-0119 Page 2 Abbreviations and Acronyms:

AFW - Auxiliary Feedwater IST - In-service Testing ALARA - As Low as Reasonably Achievable LBLOCA - Large Break Loss of Coolant Accident ANSI - American National Standards Institute LCO - Limiting Conditions for Operation AOP - Abnormal Operating Procedure LOCA - Loss of Coolant Accident AOR - Analysis of Record LPSI - Low Pressure Safety Injection AR - Action Request LTOP - Low Temperature Overpressure Protection BAST - Boric Acid Storage Tank MCC - Motor Control Center BTP - Branch Technical Position MEW - Main Feedwater CCW - Component Cooling Water MH - Manhole CD-ROM - Compact Disk Read-Only Memory msl - Mean Sea Level CEA - Control Element Assembly NEI - Nuclear Energy Institute CEAPIS - CEA Position Indication System NLI - Nuclear Logistics Incorporated CFR - Code of Federal Regulations NRC - Nuclear Regulatory Commission CIV-Containment Isolation Valve NSRB - Nuclear Safety Review Board COLR - Core Operating Limits Report 01 - Operating Instruction CQE - Critical Quality Element OPPD - Omaha Public Power District CR - Condition Report PDIL - Power Dependent Insertion Limit CRS - Control Room Supervisor PORC - Plant Operations Review Committee CS -Containment Spray PRC - Plant Review Committee CW - Circulating Water PSAR - Preliminary Safety Evaluation Report DCS - Distributed Control System QA - Quality Assurance DG - Diesel Generator QATR - Quality Assurance Topical Report EA - Engineering Analysis QR - Qualified Reviewer EC - Engineering Change RCA - Root Cause Analysis EOP - Emergency Operating Procedure RCS - Reactor Coolant System ERFCS - Emergency Response Facility Computer System RFO - Refueling Outage FCS - Fort Calhoun Station, Unit No. 1 RG - Regulatory Guide FCSG - Fort Calhoun Station Guideline RPS - Reactor Protective System FSAR - Final Safety Analysis Report RSG - Replacement Steam Generators HEPA - High Efficiency Particulate Air RTD - Resistance Temperature Detector HZP - Hot Zero Power RVH - Reactor Vessel Head i&C -Instrumentation & Control RW - Raw Water IGSCC - Intergranular Stress Corrosion Cracking SARC - Safety Audit and Review Committee INPO - Institute of Nuclear Power Operation SDC - Shutdown Cooling

LIC-15-01 19 Page 3 SER - Safety Evaluation Report SSC - Structures, Systems and Components SG - Steam Generator ST - Surveillance Test SGBD - Steam Generator Blowdown TM - Temporary Modification SI - Safety Injection TS - Technical Specification SM - Shift Manager TSBC - Technical Specification Basis Change SIRWT - Safety Injection Refueling Water Tank UFSAR - Updated Final Safety Analysis Report SO - Standing Order USAR - Updated Safety Analysis Report SR - Surveillance Requirement VCT - Volume Control Tank SRP - Standard Review Plan WO - Work Order

LIC-I5-01 19 Page 4 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2014 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number EC 64651 Switchgear Supplemental Cooling Description of Activity (OI-VA-2 Compensatory Measures for R45)

Operability Evaluation 14-015 WHAT IS BEING EVALUATED BY THIS DOCUMENT During the July 2014 Inspection Manual Chapter (IMC) 0350 inspection, the NRC identified that a non-conservative design input was used for the electrical heat load in the switchgear room heat up analysis calculation, FC06102 Revision 2.

This was documented in NCV 2014009-009 in September 2014. On September 9, 2014, CR 2014-11223 was written to address immediate operability. The immediate operability determination (IOD) determined that existing supplemental cooling methods in OI-VA-2 Attachment 11 (Rev. 44) were adequate during cooler weather. These supplemental methods were already in OI-VA-2 Rev 44 specifically to address loss of normal switchgear room cooling.

To provide additional assurance of operability, Operability Evaluation 14-015 was generated. The OpEval has two compensatory measures. The compensatory measures are captured in OI-VA-2 Attachment 11 (Rev. 45) and an Adverse Condition Monitoring Plan (ACMP). The procedure change adds an improved supplemental method of ventilating the switchgear room after a loss of all normal cooling. This revised method utilizes portable fans that are staged just outside the room to blow turbine building air through the room after a loss of all cooling.

These fans are not large enough for the warmer weather/design conditions; however, they provide better flow rates and can be implemented faster than the existing supplemental methods in OI-VA-2 Revision 44. The ACMP performs daily monitoring to verify that the turbine building air temperature remains at or below 90°F. This ensures that the supply air temperature assumed for the portable fans in the OpEval remains valid.

This document evaluates the impact of the compensatory measures established

LIC-15-01 19 Page 5 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2014 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change

{Activity Title 150.59 Evaluation Summary Number_

by Operability Evaluation (OpEval)14-015. Specifically, it evaluates the impact of the changes that revision 45 makes to OI-VA-2 and the ACMP. The non-conforming condition and the adequacy of the compensatory measures to mitigate the non-conforming condition are described and evaluated in the OpEval.

BACKGROUND The following is provided as background to help set the context of this evaluation.

Brief Description of the Switchcqear Rooms and Room Coolin~q The switchgear rooms contain the electrical system normal and safety related components (e.g., 4160V buses, 480V buses, inverters, breakers, etc.) that feed power to normal power production equipment and all safety related equipment required by the Technical Specifications. Technical Specification 2.7 requires that the electrical system be operable whenever the reactor is above 300°F. USAR Section 9.10 states "The electronic equipment used in the plant safety related component can operate at 120F continuously."

The switchgear room is normally cooled by air conditioning units VA-87/89 and VA-88/90 and ventilation fans VA-41/45A/45B. The ventilation and cooling equipment associated with the switchgear room is not safety related, not completely protected from High Energy Line Breaks (HELBs), nor designed to withstand external events (e.g., seismic, flood, wind). Certain design and licensing events, like Seismic or HELB, can completely (Seismic) or partially (HELB) disable all normal switchgear room cooling. As a result, the use of temporary supplemental coolinq is discussed in USAR Section

LIC-15-O0119 Page 6 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2014 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change

[Activity Title

[.'/

';50.59 Evaluation Summary NumberI I__I_

9.10 and has been considered an acceptable backup method since the station was licensed. The electronic equipment used in the plant safety related components can operate at 1 20°F continuously per USAR Section 9.10. Therefore, the station would be outside of its design and licensing basis if supplemental cooling cannot maintain the switchgear room less than 1 20°F after a loss of all normal cooling.

Summary of Operability Evaluation 14-015 Operability Evaluation 14-015 shows that after a complete loss of all normal switchgear room ventilation and cooling, the portable fans and flow path used in the revised version of OI-VA-2 Attachment 11 (Rev. 45) are adequate to maintain the switchgear room below 120°F when supply air is at or below 90°F. The primary basis for the OpEval is the new analysis performed in FC06102 Rev.3. This calculation was revised to eliminate the nonconservative errors identified above. It provides the time available for the switchgear room to heatup under worst case design conditions and provides the minimum supplemental air flow requirements needed to keep the room below 120°F after loss of all normal cooling. In addition to the design case, the calculation assessed lesser cases (e.g., lower supplemental cooling supply air temperatures) for use in potential operability evaluations during cooler weather.

This compensatory measure added to OI-VA-2 Attachment 11 (Rev 45),

establishes air flow through the switchgear room by using portable fans to blow air from the south end of the turbine building and in through blocked open switchgear room door 1011-4. Air flow exits the north end of the room through blocked open door 1011-7. After leaving the room, a series of blocked open doors then allow the air to exhaust through the Auxiliary

LIC-15-01 19 Page 7 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2014 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number_

Building normal ventilation fans (VA-40A!B/C). If, these fans are unavailable, then the procedure provides an option to exhaust flow to the outside environment through the CARP building. Air sampling is established by Chemistry and RP if the flow is exhausted directly outside rather than through the monitored flow path provided by the VA-V-40s. The OpEval shows that after a loss of all switchgear room cooling, the time available to establish this supplemental method of cooling is 65 minutes. This is adequate time for the operators to perform the actions to implement cooling with the portable fans. The basis for the time available to perform the actions and an assessment of the time required is detailed in the Op Eval.

The procedure establishes the appropriate barrier permits (e.g., fire barrier).

However, a plant shutdown is required by the procedure since the turbine building-to-switchgear room door (1011-4) is blocked opened. This door is the High Energy Line Break (HELB) barrier between the switchgear room and the turbine building and is required as part of the design basis for HELB described in USAR Appendix M. Disabling a barriers is controlled by barrier control procedures CC-AA-201, Plant Barrier Control Program and SO-G-58, Control of Fire Protection System Impairments. The new revision to Ol-VA-2, this activity, establishes the appropriate "watches" using these barrier control procedures.

An Adverse Condition Monitoring Plan (ACMP) was described in and is being issued as part of the Operability Evaluation. It provides specific guidance for periodically checking turbine building temperature to ensure that it remains at or below 90°F. This verifies that the assumptions in the Operability Evaluation remain valid. Temperatures in this area of the turbine building are currently

('October 2014) around 75°F and typically well below 90°F during the late fall

LIC-15-01 19 Page 8 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2014 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number_

through early spring months.

This operability evaluation can be closed when corrective actions are completed to replace the portable fans with larger fans and Ol-VA-2 Rev 45 is revised to reflect the larger fans. The larger fans will provide enough airflow to cool the switchgear rooms under design basis conditions, Case 5D, of FC06102 Rev 3.

Reason for Activity This document evaluates the impact of the compensatory measures established by Operability Evaluation (OpEval)14-015. Specifically, it evaluates the impact of the changes that revision 45 makes to OI-VA-2 and the impact of the ACM P.

The procedure change adds an improved supplemental method of ventilating the switchgear room after a loss of normal cooling. This revised method utilizes portable fans that are staged just outside the room to blow turbine building air through the room after a loss of all cooling. These fans are not large enough for the warmer weather/design conditions; however, they are adequate for conditions where the supply air temperature (turbine building) is below 90°F. The new method provides better flow rates and can be implemented faster than the existing supplemental methods in OI-VA-2 Revision 44.

The ACMP adds a requirement for plant operations to check turbine building air temperature once per day, makes adjustments to allow more outside air into the turbine building if temperature rises above 85°F, and reassess

LIC-15-01 19 Page 9 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2014 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number operability of the components in the switchgear room if temperature rises above 900°F. The impact of these two compensatory measures on other SSCs and the plant design and licensing basis is reviewed under 50.59.

Effect of Activity This activity consists of (1) The procedure change, OI-VA-2 Rev. 45, which adds an improved supplemental method of ventilating the switchgear room after a loss of normal cooling and (2) the ACMP which monitors turbine building temperature to ensure it remains at or less than 90°F. These activities are evaluated for the impact that they have on plant operations, the design basis, or safety analysis described in the UFSAR. The specific item that potentially impact operations, the design bases, or safety analysis described in the UFSAR are addressed in the 50.59 and are as follows:

1. Defeating/Opening HELB and Fire Barrier Doors

2. Adding Load to the Diesel Generators

3.

Moving warm air at <120°F into the Aux Building from the Switchgear Room

4. The physical force/effect of blowing air into the switchgear room on switchgear room components

5.

Exhausting Auxiliary Building Air through the CARP Building (unmonitored release path)

6. Adding Operator Actions that are Time Critical Actions (TCAs)

Summary of Conclusion for the Activity's 50.59 Review This activity requires a 50.59 Screening. That screening indicated that a

LIC-15-0119 Page 10 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2014 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 EvalUation Summary Number.

50.59 Evaluation was required because the activity involves a change to an SSC that may adversely affect UFSAR described design functions.

Specifically, the compensatory measures required to enhance switchgear room supplementary cooling make changes to the Auxiliary Building door alignments that breach USAR described barriers and adds load to the diesels.

The conclusion of the 50.59 Evaluation was that NRC prior approval is not required and there are no required changes to the Technical Specifications.

This conclusion is based on a review of the key activities implemented by the compensatory measures in OI-VA-2 Rev 45 and the ACMP against the eight 50.59 Evaluation questions. Those key activities are:

1. Defeating/Opening HELB and Fire Barrier Doors.

2. Adding Load to the Diesel Generators

3. Moving warm air at <120°F into the Aux Building from the Switchgear Room

4. The physical force/effect of blowing air into the switchgear room on switchgear room components

5. Exhausting Auxiliary Building Air through the CARP Building (unmonitored release path)

6. Adding Operator Actions that are Time Critical Actions (TCAs)

EC 60821 Emergency RCS Fill Connections Activity Description

- FLEX Activity No. 1 This modification installs FLEX Emergency RCS Fill Connection lines into the CVCS system and SI system (Containment Spray (CS) and Shutdown

LIC-1 5-0119 Page 11 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2014 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change

[Activity Title

[50.59 Evaluation Summary Number

_I_

Cooling (SDC) piping) to allow fo netinit the RCS in accordance with the FLEX strategy. The connection line in the CH system is installed at the 1" blind flange located at the discharge of the CH-1A pump and the connection line for the SI system is installed at the 3" blind flange located on the Containment Spray (CS) header upstream of air-operated valve HCV-344.

Both connection lines consist of two manual gate valves, piping, a reducer, a Female National Pipe Thread (FNPT) to Female National Standard Thread (FNST) adapter, and a Male National Standard Thread (MNST) plug. The connection lines are prepared as spool pieces, which will be bolted on at the flange connection after the blind flanges are removed. This modification also installs a pipe support between the two gate valves for each connection line.

The new components installed as a result of this modification are CH-576, CH-577, CHSP-95, SI-51 8, SI-51 9, and SIS-243.

Activity No.2 The CVCS, CS, SDC and SI systems are Safety Class 2 systems. They were originally designed and analyzed to USAS B31.7 (Draft 1968 Edition, Code of Record). The design of the connection lines (excluding the pipe supports) conforms to the requirements of ASME Section III, 1980 edition, with Summer 1981 Addenda. Designing according to ASME Section III is consistent with current Fort Calhoun design practices described In PED-MSS-1 1 "Design Specification for Piping and Pipe Supports" as reconciled by Engineering Analysis EA-FC-91-054. This represents a methodology change from the USAR described analysis. A license amendment request (LAR 14-04, LIC-14-0043) has been submitted to allow future use of ASME Section III 1980 Edition ('no Addenda) as an alternative to B31.7, 1968 Draft for pipe

LIC-15-01 19 Page 12 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2014 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title

[50.59 Evaluation Summary Number stress analysis on non-reactor coolant safety systems, without additional reconciliation.

The 50.59 screening for the proposed activities concluded that only Activity No.2 requires a 50.59 evaluation because it represents a change in the USAR described methodology and is therefore subject to this evaluation.

Summary of Evaluation The use of ASME Section III 1980 Edition with 1981 Summer Addenda as reconciled to B31.7, 1968 Draft for pipe design and analysis is a change from the USAR methodology of using B31.7, 1968 Draft. This is a methodology change only; therefore, it does not impact the frequency of USAR analyzed accidents, the likelihood of USAR analyzed component malfunctions, the consequence of USAR analyzed accidents, or the consequence of USAR analyzed component malfunctions. In addition, the methodology change does not create new accidents types or malfunctions other than those analyzed in the USAR. The methodology change also does not change USAR design basis limits for fission product barriers.

Previous reconciliation documented in Engineering Analysis EA-FC-91-054 and additional reconciliation of applicable requirements not addressed in EA-FC-91-054 provide the basis for the conclusion that the change in methodology is equivalent. No USAR design functions are changed as a result of this modification. The conclusion of this evaluation is that a License Amendment is not required.

LIC-1 5-0119 Page 13 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change

[Activity Title 50.59 Evaluation Summary Number

[________________

EC 65174 Condensate Return Line Caps in Description of Activity Turbine Building The proposed activity temporarily modifies a portion of the Condensate Return piping in the Turbine building basement. This piping supports the return of auxiliary steam condensate to the drip and drain tank. The temporary modification is being completed to isolate the Auxiliary Building Condensate Return piping from the Condensate Return piping in the reminder of the Turbine Building which includes the Condensate Return line from the Intake Structure. Additionally a sample sink drain line attached to the piping being isolated requires extension so that it will drain into piping that will be in service after installation.

Reason for Activity The potential for High Energy Line Cracks (HELC's) in the Condensate Return piping poses a threat to the operation of Auxiliary Feedwater pumps FW-6 and FW-1 0 in Room 19 located in the Auxiliary Building at Elevation 989.' This activity temporarily isolates the Condensate Return line in the Auxiliary Building from the remainder of the system so Auxiliary Steam can be returned to service in the Intake Structure and CST without the HELC concern in Room 19.

Effect of Activity The Condensate Return line being modified in this activity is a high energy line and part of the Auxiliary Steam System (UFSAR Section 9.10). This modification will separate the Auxiliary Building and Turbine Building portions of the condensate return line. This will prevent the Auxiliary Building portion

LIC-15-01 19 Page 14 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change

[Activity Title T50.59 Evaluation Summary Number

_ I of the condensate return line from being used until a permanent tie in is made but will allow for the use of all portions of the Turbine Building condensate return line including the return from the intake structure. This result has an adverse effect on the ability of the Auxiliary Building HVAC system to heat desired spaces because Auxiliary Steam will not be available. Extension of the sample sink drain line has no impact on the plant other than installation of additional piping.

Summary of Conclusion for the Activity's 50.59 Review A 50.59 review was performed for the activities described in this modification.

The 50.59 applicability determination shows that the activities are not controlled by any of the processes considered in LS-AA-1 04-1002. As such, a 50.59 screen was performed. The 50.59 screen is attached (LS-AA-1 04-1003). The screen shows that the proposed activity adversely affects a USAR described design function and a procedure which performs or controls this design function. Therefore, a 50.59 Evaluation was performed to evaluate the effect of the proposed activity on accidents and malfunctions previously evaluated in the UFSAR and the potential to cause accidents or malfunctions whose effects are not bounded by previous analyses. The 50.59 Evaluation determined that the proposed activity could be implemented without prior NRC approval.

EC 66460 Install Gag on Relief Valve AC-168

-Description of Activity A gag is being installed on the Component Cooling Water (CCW) inlet thermal relief valve for Reactor Coolant Pump (RCP) RC-3C seal cooler, AC-168.

LIC-15-O0119 Page 15 Changes, Tests, and Experiments Performed Pursuant to 10 CER 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change

[Activity Title T-50.59 Evaluation Summary Number_

Reason for Activity AC-I168 is being gagged shut due to leakage concerns. Temporarily gagging AC-i168 will ensure its pressure boundary is maintained on its branch loop of the CCW system.

Effect of Activity The activity has been evaluated and found to have no adverse effect. The intent of the activity is to ensure that the USAR described functions of CCW system are maintained. Relief valve AC-i168 provides a thermal relief function for the protection of the seal cooler for RC-3C and its surrounding CCW piping. The thermal relief function is required should the cooler be isolated for maintenance, or other concern. AC-168 is isolated by closing manual valve AC-260 and control valve HCV -444. If a thermal transient occurs during the time when the relief valve is isolated, the CCW pipe and / or cooler could rupture. Loss of cooling to the seal would occur and a breach in the CCW system pressure boundary would exist. The breach would cause loss of CCW system inventory. If the branch loop containing AC-I168 were to be un-isolated by opening control valve HCV-444, or manual valve AC-260, a greater loss of CCW inventory would occur. That being said, it is not a reasonable concern to assume that the cooler would be isolated at a time where a thermal transient would occur that would rupture the CCW branch loop. It would mean that maintenance was being conducted during operation or the loop was isolated during operation. Neither scenario is likely, if ever, to occur.

Gagging AC-168 closed when AC-260 and HCV-444 are open is not a

LIC-15-01 19 Page 16 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number concern because if a transient occurs the resulting pressure would expand out and open another relief valve in the system.

There is no impact on plant operations, design bases, or safety analyses described in the UFSAR.

Summary of Conclusion for the Activity's 50.59 Review The 50.59 evaluation concludes that the proposed activity does not have any adverse impact on any USAR described functions or how any USAR described functions are controlled. Therefore, the proposed activity may be implemented without NRC prior approval.

The gagging of AC-i168 does not adversely affect the function, operation, or control of the CCW system. It does not affect any other plant system. No automatic system functions, other than the relief function of the valve, are being affected. No procedures are being changed. Therefore, UFSAR described design functions are not adversely affected, and, therefore, how UFSAR described design functions are performed or controlled are not adversely affected. The gagging of the valve does not involve a revision, or replacement, of an evaluation methodology. As such, neither the design basis nor the safety analyses are affected. The gagging of the valve does not involve a new test or experiment. As such, no SSC is utilized or controlled in an adverse manner. The activity is not controlled by the processes of the Applicability Review. The activity requires a 50.59 Evaluation under the 50.59 Screening.

LIC-15-01 19 Page 17 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized result, the language may be in future tense.

below are for the most part, unedited summaries as approved by the PORC. As a Change J Activity Title

'_50.59 EvalUation Summary EC 64574 Address EDG Load Shedding Concerns in Fire Area 36A by Rerouting Cables and Adding Coordinated Fuses and Isolation Switches (REC-1 18) (Ref NFPA-805 Transition LAR LIC-1 1-0099 Table S-2 Description of Activity Electrical isolation is required to preclude fire induced cable damage from causing a failure of breaker control circuits. These cable failures may introduce a direct short to ground (or hot short), with a potential to fail breaker I25VDC TRIP control power or cause uncontrolled, spurious operation of a breaker.

The scope of EC 64574 includes the following activities:

1. For the breaker 1 B4A-7 (480 VAC MCC-4A3) control circuit, cable 7700B will be re-routed from Tray Section 57S to Panel AI-109B in the existing Pyrocrete enclosure. The breaker TRIP control circuit will be electrically isolated from a fault on the Train A load shed input cable (cable 7700A to AI-109A) by installing a fuse and a blocking diode in Panel Al-i109B.

2. For the breaker 1 B4C-1 (480 VAC MCC-4C5) control circuit, cable Bill183 will be re-routed and a fuse will be installed on the conductor from Panel Al-i109B to Panel Al-i109A. The breaker TRIP control circuit will be electrically isolated from faults induced in the non-scheduled cable between Al-i109A and Al-i109B by installing a fuse and a blocking diode in Panel AlI-109B.

3.

For Containment Spray pump SI-3C breaker I1B3B-4B-3, an isolation switch will be installed in a new junction box mounted on the west wall within Fire Area 36B. This switch will isolate the breaker control circuit from all Train A automatic load shed and load sequencing signals to prevent a spurious closing of the breaker in response to fire induced faults in Fire Area 36A. The switch will be used to normally

LIC-15-01 19 Page 18 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number I ____________________________________________________________________

Change ActivitY Title 50.59 Evaluation Summary Number disconnect the Al-i108A conductors from the breaker trip circuit when 1 B3B-4B-1 is aligned to Train B.

4. For VA-i151 B, cable B2998 will be re-routed from Tray Section 57S to Panel Al-i109B in the existing Pyrocrete enclosure.

5.

For VA-i151 D, cable B3000 will be re-routed from Tray Section 57S to Panel AlI-109B in the existing Pyrocrete enclosure.

6. For EHC-3B, cable B8970 will be re-routed in a conduit from the Al-109B enclosure directly through the wall into the Turbine Building, rather than running the conduit through unprotected Tray Section 56S in Fire Area 36A.

7.

For ST-6B, cable B2630 will be re-routed in a conduit from the Al-1 09B enclosure directly through the wall into the Turbine Building, rather than running the conduit through unprotected Tray Section 56S in Fire Area 36A.

Reason for Activity In 2010, Omaha Public Power District (OPPD) Fort Calhoun Station (FCS),

Unit No. 1 submitted License Amendment Request (LAR) 10-07 (Ref. LIC-1 1-0099) to the Nuclear Regulatory Commission (NRC) to amend the Operating License, No. DPR-40, to adopt the National Fire Protection Association (NFPA) 805, Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition) standard.

As part of the NFPA 805 transition process, an outside, third-party was contracted to perform a Nuclear Safety Capability Assessment (NSCA), a Non-power Operations (NPO) Assessment, and a Fire Probabilistic Risk Assessment (Fire PRA) to evaluate the plant response to deterministic and

LIC-15-01 19 Page 19 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change 1Activity Title 150.59 Evaluation Summary Number realistic fire events. These assessments were documented in EPM Report R2008-004-002, early in 2015 (Ref. EPM Report EA-10-036).

The result of this report determined that a number of modifications are required to be performed to bring the plant configuration into compliance with the NFPA 805 standard. The report identified Recommended Engineering Change (REC) REC-1 18 which was committed to be performed in LIC-1 1-.

0099, Attachment S (Table S-2). It is the purpose of EC 64574 to implement REC-118.

On June 16th, 2014, the NRC approved the license amendment via letter NRC 14-0072, to be implemented on June 16th, 2015 with the contingency that the committed to engineering changes were implemented at the station (Ref. SE ML14098A092).

Effect of Activity The scope of EC 64574 involves the re-routing of control circuit cables, installation of an isolation switch in the control circuit for SI-3C and the installation of fuses and blocking diodes, in order to prevent spurious breaker operation or a loss of breaker TRIP capability in the event of a fire in Fire Area 36A. The addition of the diode to the circuit is a reduction in reliability because it is a semiconductor device with a limited shelf and service life. An open circuit of the diode will isolate one load shed train. A short circuit of the diode will have no impact on the design functions. There is no change in how any components are operated or controlled.

LIC-15-01 19 Page 20 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 150.59 Evaluation Summary Number

__r_

Summary of Conclusion for the Activity's 50.59 Review EC 64574 is being implemented per LAR 10-07 to adopt the NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition) standard. This configuration change is outside the previously performed 50.59 Screening, which was performed under EC 50741 in conjunction with transition to the NFPA 805 program. Due to the change in plant configuration as a result of this EC, a 50.59 Screening is performed for this activity in order to ensure that no adverse impacts to the plant USAR or technical specifications occur.

Results of the screening determined that the addition of the diode is an adverse activity due to impacting the reliability of the load shed circuit. The remainder of the activities are not adverse because design functions are not impacted and will be controlled in the same manner as they currently are.

The evaluation has determined that prior NRC approval is not required because the diode is significantly more reliable than the other components in the circuit. No accidents are impacted and all failure modes are bound by the existing malfunctions.

EC 58161 Modification to Provide Electrical Description of Activity Isolation for Breaker Trip Control Circuits Associated with 4kV Circuit The scope of EC 58161 includes installing a fuse both locally at each breaker Breakers (REC-1 12) (See Table 5-(see below), and within the control board console CB-1 0/11. The new fuses 2 in LIC-1 1-0099, NFPA-805 will protect and ensure continued availability of the 125VDC control power to Transition LAR) each breaker's TRIP circuit. Fusing the TRIP control cables will isolate any faults, allowing each breaker to be tripped. A blocking diode will be installed within each control circuit to prevent any shorts to ground from grounding the TRIP signals.

LIC-15-01 19 Page 21 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number Fault protection will be provided by installing fuses, blocking diodes, and performing circuit re-wiring to the following 4160 V breaker control circuits:

1A1-0, Fire Pump FP-1A 1A1-1, Heater Drain Pump, FW-5A 1A1-2, S/G Feed Pump, FW-4A 1A1-3, Condensate Pump, FW-2A 1A1-4, Circulating Water Pump, CW-IA 1A2-6, Circulating Water Pump, CW-1B 1A2-7, Condensate Pump, FW-2B 1A2-8, S/G Feed Pump, FW-4B 1A2-9, Heater Drain Pump, FW-5B 1A4-3, Circulating Water Pump, CW-lC 1A4-4, Heater Drain Pump, FW-5C 1A4-5, S/G Feed Pump, FW-4C 1A4-6, Condensate Pump, FW-2C Reason for Activity In 2010, Omaha Public Power District (OPPD) Fort Calhoun Station (FCS),

Unit No. 1 submitted License Amendment Request (LAR) 10-07 (Ref. LIC-1 1-0099) to the Nuclear Regulatory Commission (NRC) to amend the Operating License, No. DPR-40, to adopt the National Fire Protection Association (NFPA) 805, Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition) standard.

As part of the NFPA 805 transition process, an outside, third-party was

___________________________________contracted to perform a Nuclear Safety Capability Assessment (NSCA), a

LIC-15-01 19 Page 22 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title, 50.59 Evaluation Summary Number Non-power Operations (NPO) Assessment, and a Fire Probabilistic Risk Assessment (Fire PRA) to evaluate the plant response to deterministic and realistic fire events. These assessments were documented in EPM Report R2008-004-002, early in 2015 (Ref. EPM Report EA-10-036).

The result of this report determined that a number of modifications are required to be performed to bring the plant configuration into compliance with the NFPA 805 standard. The report identified Recommended Engineering Change (REC) REC-1 12; which was committed to be performed in LIC-1 1-0099, Attachment S (Table S-2). It is the purpose of EC 58161 to implement REC-1 12.

On June 16th, 2014, the NRC approved the license amendment via letter NRC 14-0072, to be implemented on June 16th, 2015 with the contingency that the committed to engineering changes were implemented at the station (Ref. SE ML14098A092).

Effect of Activity The fuses have been evaluated and determined to meet electrical requirements for use in the identified 4160 V breaker TRIP circuits. The fuses were also determined to meet seismic requirements for installation in control board panel CB-1 0/11. The circuit re-wiring will result in no change to pump control or operation. The introduction of the diode introduces a new failure mode of the circuit where local trip control can be lost without tripping the breaker. The EC will prevent a loss of each breaker's TRIP capability in the event of a fire in Fire Areas 31, 46 and 47.

LIC-15-0119 Page 23 Changes, Tests, and Experiments Performed Pursuant to 10 CER 50.59 201 5 Evaluations Chnotg heA10cFRit 5059evluationsEvasummarizedar reultbterlnug a

ei uuetne below are for the most part, unedited summaries as approved by the PORC. As a Change Activity Title J 50.59 Evaluation Summary Number I

Summary of Conclusion for the Activity's 50.59 Review EC 58161 is being implemented per LAR 10-07 to adopt the NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition) standard. This configuration change is outside the previously performed 50.59 Screening, which was performed under EC 50741 in conjunction with transition to the NFPA 805 program. Due to the change in plant configuration as a result of this EC, a 50.59 Screening is performed for this activity in order to ensure that no adverse impacts to the plant USAR or technical specifications occur.

The screening determined that the SSC design functions are adversely affected due to the installation of the blocking diode creating a new failure mechanism and mode. An evaluation was performed and determined that prior NRC approval is not required. The impact on design function in not significant because the reliability of the diode far exceeds the other limiting components in the loop and all new failure modes are bound by the existing evaluations in the USAR.

EC 66825 Inadvertent Digital Upgrade of Description of Activity Degraded Voltage Relays (OPLS)

Fort Calhoun Station (FCS) was notified that Allen-Bradley 700-RTC Time Delay Relays had been transitioned from a solid-state to a complex digital based device without part number change or notification from the manufacturer. These digital relays were previously installed as like-for-like replacements on three of the four Engineered Safeguard Features (ESF) channels for the Offsite Power Low Signal (OPLS). Because the relays were thought to be like-for-like; they were installed as a maintenance activity without a 10 CFR 50.59 review. The station has decided to correct this non-

LIC-15-01 19 Page 24 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change 1Activity Title 50.59 Evaluation Summary.

Number conformance by accepting the Condition and incorporating the necessary changes into the design and licensing basis. This 50.59 review will determine if a license amendment is required to accept this change.

Reason for Activity In August 2014, United Controls supplied four (4) time delay relays to Omaha Public Power-Fort Calhoun Station. The subject relay was qualified in accordance with IEEE 323-74/83, IEEE 344-1 975/1 987, and IEEE C37.98-1987, for use in mild environment safety related applications.

Per NRC Part 21 notifications, UCI was informed that the Allen Bradley relays base model 700RTC contain a Complex Programmable Logic (CPLD) which was unpublished. This design change could not be noticed since the external appearance of the relay and the relay part number remained the same.

Hence, UCI has qualified the subject relay as solid state relay whereas the presence of the CPLD device elects the item as a digital device which can be affected by EMI/RFI noises.

Due to the event described above, the station has declared a non-conformance on the installed relays.

Effect of Activity The affected relays are 27-T1/OPLS-A, 27-TI/OPLS-C, and 27-TI/IOPLS-D.

These relays provide the time delay function for the degraded voltage control relays. The OPLS signal is used to ensure offsite power voltage levels are sufficient to start and operate all reauired electrical loads in the event of a

LIC-15-O0119 Page 25 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change 1

Activity 'Title J 50.59 EvalUation Summary Number design basis accident (DBA). The three affected relays are located on the stations 4160 VAC electrical distribution buses. If an undervoltage condition is detected, then the control relays actuate; which in turn actuates the time delay relays. If voltage does not recover within the time delay period, then the time delay relays contacts are actuated and the electrical buses are transferred to the emergency diesel generators (EDG).

With the exception of being digital devices, the installed relays are functionally equivalent to the solid-state relays that were replaced. Therefore, the effect of the activity is limited to the software considerations as described in NEI 01-01, Guideline on Licensing Digital Upgrades EPRI TR-1 02348 Revision 1.

Summary of Conclusion for the Activity's 50.59 Review No other processes where identified by the Applicability Determination. The Screening determined that there is an adverse impact due to the possibility of software common cause failure. Since the physical replacement of the relay was previously screened and there are no human interface concerns with a relay; none of the other criteria were applicable. The Evaluation determined that prior NRC approval is not required because the function of the relay is extremely simple and has a strong pedigree to support reliable operation.

EC 66490 Add a Varistor Across HCV-1 041A-Description of Activity 20A & HCV-1042A-20A This EC will install a varistor across the terminals of HCV-1041A-20A (Main Steam Valve HCV-1041A Pilot Solenoid Valve) and HCV-1042A-20A (Main Steam Valve HCV-1042A Pilot Solenoid Valve).

LIC-15-01 19 Page 26 Changes, Tests, and Experiments Performed Pursuant to 10 CER 50.59 2015 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number Reason for Activity Varistors will be installed across the pilot (opening) solenoid coils for HCV-1041A and HCV-1042A to arrest induced voltage spikes. Based on troubleshooting associated with CR 2015-07217, spikes greater than the PC-11 5A and PC-I118A setpoint of 17.5 MA for the trip of shutdown cooling were discovered during the cycling of both HCV-1041A and HCV-1042A. Installing varistors for the purpose of arresting voltage spikes is a common industry practice.

Effect of Activity HCV-1041A and HCV-1042A are Main Steam Isolation Valves (MS IVs) used to isolate the steam generators from the main steam header. The valves are provided to isolate the steam generators during normal and accident conditions. While the addition of the varistor to the control loop will improve performance of the external loops affected by the EMI spike, it also reduces the reliability of the 1041A and 1042A control loops by increasing the probability of an existing failure mechanism (i.e. shorted varistor, shorted solenoid). In general, failure of HCV-1041A and HCV-1042A, regardless of the initiating failure mechanism, could result in the inability to control an excessive reactor coolant system cooldown rate and resultant reactivity insertion following a main steam break incident.

Summary of Conclusion for the Activity's 50.59 Review The Applicability Review determined that no other regulatory processes are applicable. The screening has determined that there may be an adverse

LIC-15-01 19 Page 27 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number impact to the MSIV design functions due to the addition of components to the control loops. The evaluation determined that prior NRC approval is not required because the addition of varistors does not fulfill any of the 10 CFR 50.59 criteria because the reduction in reliability is minimal and no new failure modes are created (i.e. no new accidents or malfunctions result).

EC 85749 increasing Portable Fan Size and Description of Activity (OI-VA-2 Changing the Flow Path for R47)

Switchgear Room Supplemental WHAT IS BEING EVALUATED BY THIS DOCUMENT Cooling During the July 2014 Inspection Manual Chapter (IMC) 0350 inspection, the NRC identified that a non-conservative design input was used for the electrical heat load in the switchgear room heat up analysis calculation, FC06102 Revision 2. This was documented in NCV 2014009-009 in September 2014. On September 9, 2014, CR 2014-11223 was written to address immediate operability. The immediate operability determination (IOD) determined that existing supplemental cooling methods in Ol-VA-2 1 (Rev. 44) were adequate during cooler weather, but inadequate during warmer weather/design conditions. These supplemental methods were already in OI-VA-2 Rev 44 specifically to address loss of all normal switchgear room cooling.

To provide additional assurance of operability at higher ambient temperatures, Operability Evaluation 14-015 was generated and Procedure OI-VA-2 Revision 45 was issued. OI-VA-2 Revision 45 provided additional detail on using existing portable fans to supply air from the turbine building through the switchgear room. The OpEval showed that these fans could maintain the room below 1200F after a loss of all normal cooling at supply air temperatures below 900F. Use of these fans was an interim solution until a larger portable fan could be obtained.

LIC-15-01 19 Page 28 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORO. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number This document evaluates changes to Procedure Ol-VA-2 Attachment 11, Calculation FC061 02, and a USAR change to Section 9.10. These documents are being issued under EC 65749. Specifically, the changes are:

OI-VA-2 Attachment 11 Rev 45 is being revised to o

Utilize a new large capacity portable fan to provide switchgear room supplemental cooling in the event that all normal cooling is lost. The new portable fan replaces the smaller existing portable fans and is stored in an existing storage cage in the switchgear room.

o Change the supplemental cooling flow path to take air from the south end of the turbine building and return it to the North end.

The existing procedure revision discharges air out through the RCA portion of the Auxiliary Building. This new path is preferable since it eliminates breaching the RCA barrier.

FC06102 Rev 3 is being revised to:

o Evaluate the larger capacity portable fan and revised air flow path to show that it will maintain switchgear room temperature below the USAR required temperature of 120°F using supply air sources (turbine building or outside air) that are at or below the USAR design temperatures. The USAR design temperature is 95°F for outside air (USAR Section 9.10.1) and 1 05°F for turbine building air (USAR Section 9.10, Table 9.10-1).

o Eliminate GOTHIC runs/cases from the calculation that are not utilized (e.g., supplying air from the diesel rooms).

USAR Section 9.10 page 7 is being revised to clarify what "supplementary cooling" entails. This has always consisted restarting existing fans or use of portable fans; however, that was never clearly defined in the USAR nor

LIC-15-01 19 Page 29 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change 1Activity Title 1

"50.59 Evaluation Summary Number in any previous design basis documents or NRC correspondence. The changes clarify that supplementary cooling includes a portable fan sized to maintain the room below 120°F after losing all normal cooling.

After implementation of these documents through EC 65749, OpEval 14-15 will be closed.

BACKGROUND The following is provided as background to help set the context of this evaluation.

Brief Description of the Switchgear Rooms and Room Coolincq The switchgear rooms contain the electrical system normal and safety related components (e.g., 4160V buses, 480V buses, inverters, breakers, etc.) that feed power to normal power production equipment and all safety related equipment required by the Technical Specifications. Technical Specification 2.7 requires that the electrical system be operable whenever the reactor is above 300°F. USAR Section 9.10 states "The electronic equipment used in the plant safety related component can operate at 120F continuously."

The switchgear room is normally cooled by air conditioning units VA-87/89 and VA-88/90 and ventilation fans VA-41/45A/45B. The ventilation and cooling equipment associated with the switchgear room is not safety related, not completely protected from High Energy Line Breaks (HELBs), nor designed to withstand external events (e.g., seismic, flood, wind). Certain design and licensing events, like Seismic or HELB, can completely (Seismic) or partially (HELB) disable all normal switchgear room cooling. As a result,

LIC-15-01 19 Page 30 Changes, Tests, and Experiments Performed Pursuant to 10 CER 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change

[Activity Title 50.59 Evaluation Summary Number L

the use of temporary supplemental cooling is discussed in USAR Section 9.10 and has been considered an acceptable backup method since the station was licensed. The electronic equipment used in the plant safety related components can operate at 120°F continuously per USAR Section 9.10. Therefore, the station would be outside of its design and licensing basis if supplemental cooling cannot maintain the switchgear room less than 1200F after a loss of all normal cooling.

Summary of Operability Evaluation 14-015 Operability Evaluation 14-015 shows that after a complete loss of all normal switchgear room ventilation and cooling, the portable fans and flow path used in the revised version of OI-VA-2 Attachment 11 (Rev. 45) are adequate to maintain the switchgear room below 120°F when supply air is at or below 90°F (later revised to 98°F). The primary basis for the OpEval is the analysis performed in FC06102 Rev.3. This calculation was revised to eliminate the non-conservative errors. It provides the time available for the switchgear room to heatup under worst case design conditions and provides the minimum supplemental air flow requirements needed to keep the room below 1 20°F after loss of all normal cooling. In addition to the design case, the calculation assessed lesser cases (e.g., lower supplemental cooling supply air temperatures) for use in potential operability evaluations during cooler weather. The OpEval was written because the portable fan used in OI-VA-2 1 is undersized and therefore, may not have maintain the room below 1 20°F with supply air from the turbine building at the USAR design temperature of 105°F.

I ________________________________________

L

LIC-15-01 19 Page 31 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number Sum mary of the Existincq Version of OI-VA-2 Attachment 11 ( Rev 45)

OI-VA-2 Attachment 11 Rev 45, previously evaluated under 50.59, establishes air flow through the switchgear room by using portable fans to blow air from the south end of the turbine building and in through blocked open switchgear room door 1011-4. Airflow exits the north end of the room through blocked open door 1011-7. After leaving the room, a series of blocked open doors then allow the air to exhaust into the north end of the turbine building. Calculation FC06102 Rev. 4 shows that after a loss of all switchgear room cooling, the time available to establish this supplemental method of cooling is 65 minutes. This is adequate time for the operators to perform the actions to implement cooling with the portable fans. The basis for the time available to perform the actions and an assessment of the time required was previously evaluated.

The procedure establishes the appropriate barrier permits. However, a plant shutdown is required by the procedure since the turbine building-to-switchgear room door is blocked opened. These doors are High Energy Line Break (HELB) barriers between the switchgear room and the turbine building and required as part of the design basis for HELB described in USAR Appendix M. Disabling barriers is controlled by barrier control procedures CC-AA-201, Plant Barrier Control Program and SO-G-58, Control of Fire Protection System Impairments.

Reason for Activity This activity is being performed to complete the corrective actions necessary to close Operability Evaluation 14-015. The actions include changes to Procedure OI-VA-2 Attachment 11, Calculation FC061 02, and a USAR

LIC-1 5-0119 Page 32 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change 1Activity Title 50.59 Evaluation Summary Number change to Section 9.10. These documents are being issued under EC 65749. Specifically, the changes are:

OI-VA-2 Attachment 11 Rev 45 is being revised to o

Utilize a new large capacity portable fan to provide switchgear room supplemental cooling in the event that all normal cooling is lost. The new portable fan replaces the smaller existing portable fans and is stored in an existing storage cage in the switchgear room.

o Change the supplemental cooling flow path to take air from the south end of the turbine building and return it to the North end.

The existing procedure revision discharges air out through the RCA portion of the Auxiliary Building. This new path is preferable since it eliminates breaching the RCA barrier.

FC061 02 Rev 3 is being revised to:

o Evaluate the larger capacity portable fan and revised air flow path to show that it will maintain switchgear room temperature below the USAR required temperature of 120°F using supply air sources (turbine building or outside air) that are at or below the USAR design temperatures. The USAR design temperature is 95°F for outside air (USAR Section 9.10.1) and I105°F for turbine building air (USAR Section 9.10, Table 9.10-1).

oEliminate GOTHIC runs/cases from the calculation that are not utilized (e.g., supplying air from the diesel rooms).

USAR Section 9.10 page 7 is being revised to clarify what "supplementary cooling" entails. This has always consisted restarting existing fans or use of portable fans; however, that was never clearly defined in the USAR nor in any previous design basis documents or NRC correspondence. The changes clarify that supplementary cooling includes a portable fan sized

LIC-15-01 19 Page 33 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title

• 50.59 Evaluation Summary Number to maintain the room below 120°F after losing all normal cooling.

Effect of Activity This activity consists of a change to the supplementary cooling method used in OI-VA-2 Attachment 11. The revised procedure will utilize a larger portable fan and a different air flow path. The change in flow path exhausts air back to the north end of the turbine building instead of outside through the RCA portion of the auxiliary building. The larger fan applies a larger load (20 hp) on the diesels. The analysis that supports this work is FC06102 Rev 4 and EC 65749.

USAR section 9.10 is revised to add clarity to the discussion on switchgear room supplementary cooling; however, the USAR change does not affect the safety analyses described in the USAR.

This activity is evaluated for the impact that it has on plant operations, the design basis, or safety analysis described in the UFSAR. The specific items that potentially impact operations, the design bases, or safety analysis described in the UFSAR are addressed in the 50.59 and are as follows:

1. Defeating/Opening HELB and Fire Barrier Doors

2. Adding Load to the Diesel Generators

3. The physical force/effect of blowing air into the switchgear room on switchgear room components

4. Adding Operator Actions that are Time Critical Actions (TCAs)

LIC-15-01 19 Page 34 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number Summary of Conclusion for the Activity's 50.59 Review This activity requires a 50.59 Screening. That screening indicated that a 50.59 Evaluation was required because the activity involves a change to an SSC that may adversely affect UFSAR described design functions.

Specifically, the switchgear room supplementary cooling makes changes to the Auxiliary Building door alignments that breach USAR described barriers and adds load to the diesels.

The conclusion of the 50.59 Evaluation was that NRC prior approval is not required and there are no required changes to the Technical Specifications.

This conclusion is based on a review of the proposed changes to Ol-VA-2 against the eight 50.59 Evaluation questions. Those key activities are:

1. Defeating/Opening HELB and Fire Barrier Doors.

2. Adding Load to the Diesel Generators

3. The physical force/effect of blowing air into the switchgear room on switchgear room components

4. Adding Operator Actions that are Time Critical Actions (TCAs)

EC 63785 Replace Pressurizer Heaters Description of Activity The purpose of EC 63785 is to replace all 36 safety-related pressurizer heater elements with new Westinghouse heater elements of an improved design.

The pressurizer is a vertically standing vessel designed to operate with the top half full of steam and the bottom half full of water. The pressurize heaters

___________________________________enter the pressurizer vessel from the bottom. The configuration of the bottom

LIC-15-01 19 Page 35 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 150.59 Evaluation Summary Number_

_ I_

head and support plate result in two configurations of pressurizer heaters.

The new 'inner' pressurizer heaters will be component number WATLOW STE 69LX-4590; the new 'outer' pressurizer heaters will be component number WATLOW STE 69LX-4591.

The replacement Westinghouse pressurizer heaters elements are longer, eliminating the internal heater elements in the zone adjacent to the support plate. This minimizes operating differential temperature stresses in the heater sheath. In addition, the heater sheaths will be annealed following installation to reduce residual stresses and shot peened to induce compressive stresses on its outer surface region/layer. These improvements will minimize the potential for Inter Granular Stress Corrosion Cracking (IGSCC) of the pressurizer heater sheath.

The other design change is that the pressurizer heater will be connected to the pressurizer heater nozzles with a filet weld rather than a full penetration weld. The weld stresses will remain within ASME Code allowable limits (FC07276). The replacement pressurizer heater electrical connection and power requirements are the same as the existing pressurizer heaters.

The pressurizer heaters are designed for submerged operation. In the existing design, during a system upset condition where the pressurizer water level decreases below 32 percent, the pressurizer heaters are deenergized.

The 'cut out' set point [of] the [pressurizer heaters] will be raised to 35% (Ref.

FC071 80 Ri) to accommodate the longer heaters. The new set point is still within the limits for the ECS transients and Safety Analysis (calculation FC0840 1).

The Ioncier heaters will reduce the RCS volume by a maximum of 0.237 ft3.

LIC-15-01 19 Attachment I Page 36 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title

{ 50.59 Evaluation Summary Number I

I__________________________________________________________________

Change Activity Title 50.59 Evaluation Summary, Number The total Reactor Coolant System (RCS) volume (6616 ft3) will be reduced by 0.004%. In addition, the weight of the heaters will be increased by 85 lbs.

This results in an overall pressurizer weight increase.

Reason for Activity Fort Calhoun Station (ECS) replaced its pressurizer in 2006. In May of 2010, pressurizer heater # 26 failed followed by pressurizer heater # 16. The heaters were replaced with identical heaters and the Unit restarted. The root cause analysis (2012-04327) identified that the heaters failed due to IGSCC in the outside diameter of the heater sheath tube in the vicinity of the support plate. Heater No. 26 had visible signs of IGSCC which lead to a failure of the ASME Class pressure boundary. Inspections of heater No. 16 identified no signs of IGSCC or pressure boundary cracking supporting the cause of failure being electrical. Similar pressurizer heater sheath failures have also been reported at Palo Verde, St Lucie, Sizewell, and other plants. Replacing the pressurizer heaters will reduce the vulnerability of IGSCC related failures.

Effect of Activity Replacing the heaters will improve the reliability of the pressurizer and the ASME Class I pressure boundary. Replacing the pressurizer heaters will require a new pressurizer low water level 'cut out' set point. The 'cut out' set point of the [pressurizer heaters] will be raised from 32% to 35% (FC07180 RI) to accommodate the longer heaters. The new set point is still within the limits for the ECS transients and Safety Analysis (calculation FC08401). The longer heaters will take up slightly more space in the pressurizer reducing the pressurizer volume by 0.237 ft3. The combined longier heaters weiqht is

LIC-i5-01 19 Page 37 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

.Change

• ActivityTitle

-50.59 Evaluation Summaryi N um ber.

approximately 85 lbs., increasing the overall pressurizer weight.

Summary of Conclusion for the Activity's 50.59 Review:

The conclusions of the 50.59 screening is that all aspects of the design change do not involve a change to an System, Structure or Component (SSC) that adversely affects an UFSAR described design function with the exception that the heaters are larger and take up volume displacing reactor coolant volume inventory. The reduction in ROS volume was determined to be an adverse change resulting in the need for a 50.59 evaluation.

The conclusion of the 50.59 evaluation is that the small reduction in RCS volume (0.237 ft3) has a negligible effect on the frequency, likelihood, or consequences of Design Basis Accidents (DBAs). The conclusion of the 50.59 Screen and Evaluation is that the change can be made without prior NRC approval.

EC 64310 Add a Varistor Across FCV-269Y-Description of Activity 20 This EC will install a varistor across the terminals of FCV-269Y-20, Blending Tee CH-13 Boric Acid Inlet Valve Solenoid Valve.

Reason for Activity Internal QE has found that this type of valve has the potential to induce a voltage spike when its solenoid is de-energized causing spurious indications on nearby equipment. A varistor is being installed to suppress this spike; as recommended by Op-Eval 14-013; which documents EMI impacts on A/JI-

LIC-15-0119 Page 38 Changes, Tests, and Experiments Performed Pursuant to 10 CER 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number 007Y (Al-N I), Power Margin and Setpoint Dual Meter (Channel A) as a result Change Activity Title 50.59 Evaluation Summary Number of the solenoid operating.

Effect of Activity FCV-269Y is the boric acid flow control valve used for RCS makeup in the Chemical Volume Control System (CVCS). While the addition of the varistor to the control loop will improve performance of the external loops affected by the EMI spike, it also reduces the reliability of the 269Y control loop by creating a new failure mechanism (i.e. failed varistor). Failure of FCV-269Y could result in the inability to control reactivity; which could result in a plant shut down. Because A/JI-007Y also interfaces with the Reactor Protective System (RPS), the change also indirectly impacts the design functions of the RPS.

Summary of Conclusion for the Activity's 50.59 Review The Applicability Review has determined that no other regulatory processes are applicable. The screening has determined that there is an adverse impact to the CVCS design functions due to additional components being added to the control loop. The evaluation had determined that prior NRC approval is not required because the addition of the varistor does not fulfill any of the 10 CER 50.59 criteria because the reduction in reliability is minimal and no new failure modes are created (i.e., no new accidents or malfunction results).

J ______________________________________

.1

LIC-15-01 19 Page 39 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 20)15 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number EC 58237 Containment Internal Structure Description of Activity:

RVH Stand Area EC 58237 scope is to replace the existing Reactor Vessel Head (RVH) Stand with a new stand that changes the load distribution to existing Containment Internal Structure (CIS). The existing stand is 4 concrete piers that sit on the 1045'-0" slab. The new stand will be a metal frame above the slab and transfer the loads to vertical walls and columns. This is a Class I Structure and all the load combinations as defined in USAR 5.11 are applicable.

The new frame will be coated with an approved coating for the Containment Building.

The additional steel reduces the volume and adds surface area to the heat sink analysis.

There is a modification to the curb around the reactor cavity and hand rails to allow base plates to be installed for the new RVH stand.

The method for seismically restraining the RVH on the stand is changing from controlling tipping of the RVH to allowing the RVH to slide.

A steel extension is added to the top of the shield ring to address ALARA where the location of the head is approximately 3 feet higher in the new RVH stand.

The existing seismic brace that restricted movement in the south west direction (which put the shutdown cooling at risk) is no longer required. Most of the steel structural shapes are removed.

The new reactor vessel head is anchored to the Containment Internal Structure with through bolt anchors.

The use of slip critical connections in AISC 13th edition is used to

LIC-15-01 19 Page 40 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title.

015 Evlato Summary...

Num ber evaluate composite beams.

Reason for Activity During reactor core refueling the Reactor Vessel Head is removed using the overhead crane and rested on a designated Reactor Vessel (RV) Head Stand that is mounted to the top of the Operating Floor at Elevation 1 045'-0" inside Containment. The RVH load on the stand is supported by the Containment Internal Structure at Fort Calhoun Station.

An over stress condition of the existing Containment Internal Structure, including this RV head stand area, was documented by OPPD in CRs 2014-04219 and 2012-04392. OPPD committed to resolving the reactor head stand load and margin issue in their 12/2/13 letter to the NRC titled "Integrated Report to Support Restart of Fort Calhoun Station and Post Restart Commitments for Sustained Improvement." In addition, the NRC issued a letter on May 14, 2014, Fort Calhoun - NRC Integrated Inspection Report

1. NRC-14-0053 that also identified this condition and requires completion prior to the next use of the head stand.

The head stand is being replaced to restore the Containment Internal Structure to design basis requirements of UFSAR Section 5.11.

Effect of Activity The replacement of the existing head stand with a new head stand frame restores the head stand and containment internal structure to design basis for a Class I structure per the criteria defined in USAR Section 5.11. The replacement head stand is qualified for OBE (Operating Basis Earthquake)

LIC-15-01 19 Page 41 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title T50.59 Evaluation Summary Number and SSE (Safe Shutdown Earthquake) loading with the reactor vessel head resting on the head stand. The RVH has no safety functions when placed on the stand.

The new structure adds qualified coatings into containment and does not increase the debris generated inside containment following a LOCA.

The new head stand structure increases the volume of components inside containment by approximately 100 ft3. This decreases the containment free volume by approximately 0.01% which would increase accident containment pressure by less than 0.01 psid. Since the peak accident pressure is more than 2 psi less than the allowable design pressure of 60 psig the change has no impact and is acceptable.

The modification to the curbs and handrails meet design basis. The function of the concrete curb is replaced with metal kick plates.

The RVH stand meets seismic design basis and methodology as defined in USAR Appendix F.

Modifying the shield ring (steel extension) is a passive change to minimize shine from the bottom of the reactor vessel head to the workers that maintain the RVH when it is located in the stand. The new stand is approximately 3 feet higher than the existing stand. The shield ring does not have any safety related functions.

The existing seismic brace near the entrance to the RVH laydown area and the new RVH stand was installed to protect safety related equipment during Mode 5 (shut down cooling). The new RVH stand is designed so that

LIC-15-01 19 Page 42 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change TActivity Title,

{..

50.59 Evalu~ation Summary

,Number the reactor vessel head will see minimal controlled sliding during a seismic event and will not tip. Therefore the seismic brace is no longer required.

The design of the RVH anchorage is through bolts. The original code of record is silent on this type of anchors. All post installed anchor designs do not address through bolt design. The design basis for the through bolt is a methodology that is different than the design of record.

Summary of Conclusion for the Activity's 50.59 Review The conclusions of the 50.59 screening is that all aspects of the design change screen out with two exceptions. These are that the design and analysis methods utilized for 1). anchoring the new Reactor Vessel Head (RVH) stand to the Class I containment structure and 2). for the connection between the pedestal the RVH sits on and the girder, are a change in USAR described evaluation methodology warranting a Safety Evaluation. The changes do not involve a change to an System, Structure or Component (SSC) that adversely affects an UFSAR described design function. These activities do not adversely alter how the station is controlled and does not involve a test or experiment. The methodology for seismic design, from tipping to sliding is within the current ECS design basis and does not add a new method as described in FSAR/USAR Appendix F. Sliding was approved in the FSAR (original construction) for the polar crane where friction was credited to limit movement during a seismic event.

The methodology for anchor design is not within the design basis and will be evaluated.

LIC-15-0119 Page 43 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change fActivity Title

[50.59 Evaluation Summary Number

__Th deincieifoth ne recoveslha stn mette requirements for a Class I Structure defined in the USAR Section 5.11.

There is no change to the Technical Specifications.

The conclusion of the 50.59 Screen is that the methodology evaluation is required. All other activities are covered by the 50.59 screening process.

The conclusion of the 50.59 Evaluation is that the methodology is not adverse to the original code of record and does not require NRC approval.

LIC-15-O1 19 Page 1 Quality Assurance Program Changes Fort Calhoun Station, Unit No. 1 Renewed Facility Operating License No. DPR-40 NRC Docket No. 50-285 And Fort Calhoun Station Independent Spent Fuel Storage Installation NRC Docket No.72-054

LIC-15-0119 Page 2 QA Program Description Change Number

[Date of Change]

Revision 1 /

This Quality Assurance Topical Report (QATR) revision includes:

9/30/2014 changing the title Corporate Health Physicist to Manager-Radiation Protection; deleting the Division Manager-Human Resources function for the Fitness for Duty Program as this function has transitioned under the Manager-Site Security; modifying the title of the Manager-Emergency Planning and Administration to Manager-Emergency Planning; substantial changes made to the Nuclear Safety Review Board (NSRB) description; revised form numbers for 10 CFR 50.59 and 10 CER 72.48 evaluations; and removal of reference to the Corrective Action Review Board (CARB).

Revision 2 I This QATR revision includes:

12/2/2014 Revised Plant Review Committee (PRC) to Plant Operations Committee (PORC) and associated responsibilities to reflect the Exelon Quality Assurance Topical Report (QATR) description; Eliminated reference to fire protection systems and radioactive material packaging for transportation from the definition of Safety-Related.

Revision 3 /

The QATR has been revised to include applicable Exelon QATR changes made in revisions 88 and 89, as 03/26/2015 follows:

Clarify reporting structure for Document Control and Records Management. Also, clarified requirements for biennial procedure reviews.

Clarified the practice for post installation testing.

Clarified requirements for dispositioning non-conforming items in alignment with NQA-I.

Clarify NSRB audit requirements.

In addition, changes were made to reflect the Station's adoption of NFPA 805 fire protection standards per issuance of Technical SpeCification Amendment 275, to eliminate reference to ANSI standards that are addressed by NQA-1-1 994, and to clarify Manager-Systems Engineering and Manager-Engineering Programs responsibilities.

Revision 4/

This QATR revision:

11/19/2015 incorporates changes due to a shift of responsibilities for the Fort Calhoun Nuclear Oversight Department from the Vice-President Energy Delivery and Chief Compliance Officer to the Vice-President Energy Production & Marketing.

includes commitments that were inadvertently revised or were removed in the initial issue of the QATR.

LIC-15-O1 19 Page 1 10 CFR 71.106 Quality Assurance Program Approval for Radioactive Material Package Changes

LIC-15-01 19 Page 2 QA Program Description Change Number

[Date of change]

N/A No changes have been made to the quality assurance program approval for radioactive material package changes since August 13, 2015 when NRC Form 311, Quality Assurance Program Approval for Radioactive Material Packages was approved

_____________(QA Program Approval No. 71-0256, Rev. No. 8). See NRC letter dated August 13, 2015 (NRC-15-075) (ML15231A598).

LIC-1 5-0119 Page 1 Regulatory Commitments Revised in Accordance with NEI 99-04

LIC-1 5-0119 Page 2 Regulatory Commitments Revised in Accordance with NEI 99-04 Commitment

• Description Number AR 10237 This commitment tracked OPPD's commitment to implement a test program for verifying the heat transfer capability of individual heat exchangers in the CCW/RW systems and specified that CCW heat exchangers AC-IA, AC-1B, AC-IC, and AC-i1D would be tested annually. The commitment to test the heat exchangers was revised to complete inspection and cleaning on an 18-month frequency (cleaning is a license renewal commitment).

Since 1991, heat exchanger cleaning and testing has demonstrated that an 18-month cleaning interval is sufficient to ensure that the heat exchangers are capable of handling the heat load from a DBA. Testing after such cleaning has been determined to be unnecessary.

AR 9366 GL 88-17 required the implementation of procedures and administrative controls that generally avoid operations that deliberately or knowingly lead to perturbations to the RCS and/or to systems that are necessary to maintain the RCS in a stable and controlled condition while the RCS is in a reduced inventory condition.

In its response (LIC-88-1 106) dated January 4, 1989, OPPD committed to revise procedures such that if RCS water level was above the top of the hot leg nozzle prior to removing the RVH, the water level would be decreased until the steam generator U-tubes dump. This was intended to ensure that no unexpected rise in RCS water level would take place should vacuum inadvertently be lost in the steam generator U-tubes, In April 2015, the commitment was revised to clarify that the requirement to lower RCS water level until the steam generator U-tubes dump is necessary when draining the RCS to a reduced inventory, which allows for RVH removal prior to dumping the SG tubes. The RVH flange is removed at a level of 1012.5 feet, which is above reduced inventory conditions at 1010 feet (i.e., 2.5-foot margin). In a lowered inventory condition, operating experience has shown that the SG tubes dump when RCS level is below the top inner diameter of the hot leg (i.e., 1007.708 feet). Had the original commitment been maintained, it would cause entry into RCS reduced inventory conditions (i.e., < 1010 feet) when not required to remove the RVH.

AR 13723 In Amendment No. 155, the NRC approved OPPD's request to increase the spent fuel pool storage capacity to 1083 fuel assemblies. To ensure that the radiation dose to the divers was maintained ALARA, OPPD committed to follow draft RG DG-8006 ("Control of Access to High and Very High Radiation Areas in Nuclear Power Plants") if divers were used in the process of increasing the SFP storage capacity.

This one-time commitment applicable during re-racking of the SFP in the 1990's was treated as a programmatic commitment with no expiration date. The institutionalization of ALARA diving practices is demonstrated by OPPD's adoption of Exelon procedure RP-AA-461, "Radiological Controls for Contaminated Water Diving Operations."

Furthermore, no increase in the storage capacity of the SFP is planned. Thus, the commitment is no longer necessary and has been deleted.

LIC-15-01 19 Page 3 Regulatory Commitments Revised in Accordance with NEI 99-04 Commitment Description Number AR 14443 In response to a 1993 Notice of Violation, OPPO committed to develop a tag-out preparation guideline for reference during tag-out preparation to ensure that various plant configurations are considered prior to generating the tag-out. As a result of integration into the Exelon fleet, OPPD adopted Exelon procedure OP-FC-109-101, "Clearance and Tagging" that reflects current industry standard practices to ensure that plant configurations are considered when generating clearances and tag-outs. The previous OPPD procedures that contained the commitment have been superseded and

______________this commitment has been deleted as it is considered institutionalized by adoption of the Exelon procedure.

LIC-1 5-0119 Page 1 Information Removed from the USAR

LIC-15-O0119 Page 2 Information Removed from the USAR EC Number Description' EC 65695 During the Cycle 28 Core Reload Design, Note 5 was removed from Table 14.1-2, "Typical Operating Parameter Values Used in the Analysis of the Fort Calhoun Station." Note 5 previously stated "All events evaluated up to 545°F in Reference 14.1-4." Note 5 was incorporated in Table 14.1-2 in anticipation of increasing plant efficiency by raising core inlet temperature to 545°F. Although analyses were performed up to 545°F for Cycles 26 and 27, not all events were evaluated up to 545°F for Cycle 28. The note was deleted as there is currently no intention to change the core inlet

______________temperature to 5450F for Cycle 28 or any future cycle.

LIC-1 5-0119 Page 1 Summary of Technical Specification Basis Changes (TSBC)

LIC-1 5-0119 Page 2 TSBC No.

Description TS Page(s)

[Date]

14-002-0 TSBC 14-002-0 describes a normal minimum submergence level for the Raw Water pumps during a design 11-06-14basis low river event. TSBC 14-002-0 provides the basis for minimum submergence requirements of the safety related raw water pumps as low as 976'8".

14-003-0 Amendment No. 277 issued on November 6, 2014 changed footnote designations in Table 3-5 from asterisks to 12-04-14numbered footnotes. TSBC 14-003-0 made corresponding changes to footnote designations in the Basis of TS 3.2.

LIC-1 5-0119 Page 1 Technical Specification Basis Change (TSBC) Pages

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.16 River Level (Continued)

Basis.(Continued)

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 normal RW pump minimum submergence level (MSL)(2). The RW pumps can perform their design function during a design low river event of 976'9", where up to 1" of head loss can occur across the traveling screens, resulting in a cell level of 976 feet 8 inches(1,2).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 cell 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 monitoring river level to assure no sudden loss of water supply occurs, the level of the intake cells is monitored.

Intake cell levels are also adversely affected by the flows associated with the non-safety related circulating water (CW) pumps since the large flow rates associated with the CW pumps create significant head losses even with relatively clean intake cell conditions.

However, the CW pumps have a much higher MSL requirement (983 feet 0 inches) and would become unstable and trip or be manually shutdown well before intake cell levels decrease to the RW pump MSL. The head loss associated with CW pump flow would then be recovered and intake cell levels would rise.

References (1)

USAR, Section 2.7.1.2

• (2)

USAR, Section 9.8 2.16 - Page 2 Amendment No. 274

,TSBC 07-002-0 TSBC-1 4-001-0 TSBC-1 4-002-0

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.2 Equipment and Sampling Tests (continued)

The spent fuel storage-decontamination areas air treatment system is designed to filter the building atmosphere to the auxiliary building vent during refueling operations. The charcoal adsorbers are installed to reduce the potential release of radioiodine to the environment. In-place testing is performed to confirm the integrity of the filter system. The charcoal adsorbers are periodically sampled to insure capability for the removal of radioactive iodine.

The Safety Injection (SI) pump room air treatment system consists of charcoal adsorbers which are installed in normally bypassed ducts. This system is designed to reduce the potential release of radioiodine in SI pump rooms during the recirculation period following a DBA. The in-place and laboratory testing of charcoal adsorbers will assure system integrity and performance.

Pressure drops across the combined HEPA filters and charcoal adsorbers, of less than 9 inches of water for the control room filters (VA-64A & VA-64B) and of less than 6 inches of water for each of the other air treatment systems will indicate that the filters and adsorbers are not clogged by amounts of foreign matter that would interfere with performance to established levels.

If significant painting, fire or chemical release occurs such that the HEPA filters or charcoal adsorbers could become contaminated from the fumes, chemicals or foreign materials, testing will be performed to confirm system performance.

Demonstration of the automatic and/or manual initiation capability will assure the system's availability.

Verifying Reactor Coolant System (RCS) leakage to be within the LCO limits ensures the integrity of the Reactor Coolant Pressure Boundary (RCPB) is maintained. Pressure boundary leakage would at first appear as unidentified leakage and can only be positively identified by inspection. Unidentified leakage is determined by performance of an RCS water inventory balance. Identified leakage is then determined by isolation and/or inspection. Since Primary to Secondary Leakage of 150 gallons per day cannot be measured accurately by an RCS water inventory balance, footnote 3 for line item 8a on Table 3-5 states that the Reactor Coolant System Leakage surveillance is not applicable to Primary to Secondary Leakage. Primary to secondary leakage is measured by performance of effluent monitoring within the secondary steam and feedwater systems.

3.2 - Page 2 Amendment No. 15,67,128,138,169,216., 257 TSBC-0-7-003-0!, 14-003-0

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.0 SURVEILLANCE REQUIREMENTS 3.2 Equipment and Samplinci Tests (continued)

Table 3-5, Item 8b verifies that primary to secondary LEAKAGE is less or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this surveillance requirement is not met, compliance with LCO 3.17, "Steam Generator Tube Integrity," should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference 5.

The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.

The Surveillance is modified by a footnote which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

The Surveillance Frequency of daily is a reasonable interval to trend primary to secondary LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref. 5).

Table 3-5, Item 25 verifies adequate measurements are taken to ensure that facility protective actions will be taken (and power operation will be terminated) in the event of high and/or low river level conditions. The high river level limit of less than 1004 feet mean sea level is based on the maximum elevation at which facility flood control measures provide protection to safety related equipment (i.e., due to restricted access/egress to the intake structure veranda once the flood barriers are installed prior to river level reaching 1004 feet msl). A continuous watch will be established at 1002 feet mean sea level to provide adequate response time for rising river levels in accordance with the abnormal operating procedure. The river level surveillance requirement specified also ensures sufficient net positive suction head is available for operating the RW pumps. The minimum river level of 976 feet 9 inches provides adequate suction to the RW pumps for cooling plant components. The surveillance frequency of "Daily" is a reasonable interval and models guidance provided in NUREG-0212, Revision 2, "Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors," Section 4.7.6. This surveillance requirement verifies that the Missouri River water level is maintained at a level greater than or equal to 976 feet 9 inches mean sea level. A continuous watch is established to monitor the river level when the river level reaches 980 feet mean sea level to assure no sudden loss of water supply occurs.

Table 3-5, Item 26 verifies the proper position of stops on high pressure safety injection system valves.

The valves have stops to position them properly so that flow is restricted to a ruptured cold leg, ensuring that the other cold legs receive at least the required minimum flow. The refueling frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for unplanned transients if the Surveillances were performed with the reactor at power.

References

1)

USAR, Section 9.10

2)

ASTM 04057, ASTM D975, ASTM D4176, ASTM 02622, ASTM D287, ASTM 6217, ASTM D2709

3)

ASTM 0975, Table 1

4)

Regulatory Guide 1.137

5)

EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines."

3.2 - Page 5 Amendment No. 229,.24g,2.5-7, 274, 280 TSBC-09-003-0,1 !!.001 -0,14-003-0

LIC-1 5-0119 Page 1 List of Files on CD-ROM

1. File Name Size Sensitivity Location Folder Level 001 Index.pdf 35 KB Publicly Available CD-ROM 1-USAR 002 USAR 01-01.pdf 177 KB Publicly Available CD-ROM 1-USAR 003 USAR 01-02.pdf 229 KB Publiciy Availabie CD-ROM 1-USAR 004 USAR 01-03. df 192 KB Publicly Available CD-ROM 1-USAR 005 USAR 01-04.pdf 263 KB Publicly Available CD-ROM 1-USAR 006 USAR 01-05.pdf 211 KB Publicly Available CD-ROM 1-USAR 007 USAR 01-06.pdf 127 KB Publicly Available CD-ROM 1-USAR 008 USAR 01-07.pdf 109 KB Publicly Available CD-ROM 1-USAR 009 USAR 01-08.pdf 106 KB Publicly Available CD-ROM 1-USAR 010 USAR 01-09.pdf 158 KB Publicly Available CD-ROM 1-USAR 011 USAR 01-10.pdf 93 KB Publicly Available CD-ROM 1-USAR 012 USARO01-11.pdf 178 KB Publicly Available CD-ROM 1-USAR 013 USAR 01-12.pdf 172 KB Publicly Available CD-ROM 1-USAR 014 USAR 02-01.pdf 115 KB Publicly Available CD-ROM 1-USAR 015 USAR 02-02.pdf 132 KB Publicly Available CD-ROM 1-USAR 016 USAR 02-03.pdf 97 KB Publicly Available CD-ROM 1-USAR 017 USAR 02-04.pdf 124 KB Publicly Available CD-ROM 1-USAR 018 USAR 02-05.pdf 3,538 KB Publicly Available CD-ROM 1 -USAR 019 USAR 02-06.pdf 185 KB Publicly Available CD-ROM 1-USAR 020 USAR 02-07.pdf 194 KB Publicly Available CD-ROM 1-USAR 021 USAR 02-08.pdf 209 KB Publicly Available CD-ROM 1-USAR 022 USAR 02-09.pdf 201 KB Publicly Available CD-ROM 1-USAR 023 USAR 02-10.pdf 148 KB Publicly Available CD-ROM 1-USAR 024 USAR 02-11.pdf 129 KB Publicly Available CD-ROM 1-USAR 025 USAR 03-01.pdf 185 KB Publicly Available CD-ROM 1-USAR 026 USAR 03-02.pdf 222 KB Publicly Available CD-ROM 1-USAR 027 USAR 03-03.pdf 98 KB Publicly Available CD-ROM 1-USAR 028 USAR 03-04.pdf 403 KB Publicly Available CD-ROM 1-USAR 029 USAR 03-05.pdf 228 KB Publicly Available CD-ROM 1-USAR 030 USAR 03-06.pdf 212 KB Publicly Available CD-ROM 1-USAR 031 USAR 03-07.pdf 214 KB Publicly Available CD-ROM 1-USAR 032 USAR 03-08.pdf 238 KB Publicly Available CD-ROM 1-USAR 033 USAR 03-09.pdf 215 KB Publicly Available CD-ROM 1-USAR 034 USAR 03-10.pdf 103 KB Publicly Available CD-ROM 1-USAR 035 USAR 04-01.pdf 174 KB Publicly Available CD-ROM 1-USAR 036 USAR 04-02.pdf 248 KB Publicly Available CD-ROM 1-USAR 037 USAR 04-03.pdf 330 KB Publicly Available CD-ROM 1-USAR 038 USAR 04-04.pdf 192 KB Publicly Available CD-ROM 1-USAR 039 USAR 04-05.pdf 509 KB Publicly Available CD-ROM 1-USAR 040 USAR 04-06.pdf 130 KB Publicly Available CD-ROM 1-USAR 041 USAR 04-07.pdf 183 KB Publicly Available CD-ROM 1 -USAR 042 USAR 05-01.pdf 182 KB Publicly Available CD-ROM 1 -USAR 043 USAR 05-02.pdf 212 KB Publicly Available CD-ROM 1-USAR 044 USAR 05-03.pdf 214 KB Publicly Available CD-ROM 1-USAR 045 USAR 05-04.pdf 190 KB Publicly Available CD-ROM 1-USAR 046 USAR 05-05.pdf 233 KB Publicly Available CD-ROM 1-USAR 047 USAR 05-06.pdf 245 KB Publicly Available CD-ROM 1-USAR 048 USAR 05-07.pdf 222 KB Publicly Available CD-ROM 1-USAR 049 USAR 05-08.pdf 212 KB Publicly Available CD-ROM 1-USAR 050 USAR 05-09.pdf 271 KB Publicly Available CD-ROM 1-USAR

1. File Name Size Sensitivity Location Folder Level 051 USAR 05-10.pdf 227 KB Publicly Available CD-ROM 1-USAR 052 USAR 05-11.pdf 183 KB Publicly Available CD-ROM 1-USAR 053 USAR 05-12.pdf 121 KB Publicly Available CD-ROM 1-USAR 054 USAR 05-13.pdf 115 KB Publicly Available CD-ROM 1-USAR 055 USAR 06-01.pdf 200 KB Publicly Available CD-ROM 1-USAR 056 USAR 06-02.pdf 310 KB Publicly Available CD-ROM 1-USAR 057 USAR 06-03.pdf 202 KB Publicly Available CD-ROM 1-USAR 058 USAR 06-04.pdf 283 KB Publicly Available CD-ROM 1-USAR 059 USAR 06-05.pdf 176 KB Publicly Available CD-ROM 1-USAR 060 USAR 06-06.pdf 178 KB Publicly Available CD-ROM 1-USAR 061 USAR 07-01.pdf 178 KB Publicly Available CD-ROM 1-USAR 062 USAR 07-02.pdf 1,152 KB Publicly Available CD-ROM 1-USAR 063 USAR 07-03.pdf 307 KB Publicly Available CD-ROM 1-USAR 064 USAR 07-04.pdf 212 KB Publicly Available CD-ROM 1-USAR 065 USAR 07-05.pdf 298 KB Publicly Available CD-ROM 1-USAR 066 USAR 07-06.pdf 216 KB Publicly Available CD-ROM 1-USAR 067 USAR 07-07.pdf 175 KB Publicly Available CD-ROM 1-USAR 068 USAR 08-01.pdf 190 KB Publicly Available CD-ROM 1-USAR 069 USAR 08-02.pdf 207 KB Publicly Available CD-ROM 1-USAR 070 USAR 08-03.pdf 216 KB Publicly Available CD-ROM 1-USAR 071 USAR 08-04.pdf 213 KB Publicly Available CD-ROM 1-USAR 072 USAR 08-05.pdf 211 KB Publicly Available CD-ROM 1-USAR 073 USAR 08-06.pdf 100 KB Publicly Available CD-ROM 1-USAR 074 USAR 08-07.pdf 168 KB Publicly Available CD-ROM 1-USAR 075 USAR 09-01.pdf 184 KB Publicly Available CD-ROM 1-USAR 076 USAR 09-02.pdf 321 KB Publicly Available CD-ROM 1-USAR 077 USAR 09-03.pdf 216 KB Publicly Available CD-ROM 1-USAR 078 USAR 09-04.pdf 227 KB Publicly Available CD-ROM 1-USAR 079 USAR 09-05.pdf 204 KB Publicly Available CD-ROM 1-USAR 080 USAR 09-06.pdf 196 KB Publicly Available CD-ROM 1-USAR 081 USAR 09-07.pdf 162 KB Publicly Available CD-ROM 1-USAR 082 USAR 09-08.pdf 168 KB Publicly Available CD-ROM 1-USAR 083 USAR 09-09.pdf 181 KB Publicly Available CD-ROM 1-USAR 084 USAR 09-10.pdf 318 KB Publicly Available CD-ROM 1-USAR 085 USAR 09-11.pdf 474 KB Publicly Available CD-ROM 1-USAR 086 USAR 09-12.pdf 160 KB Publicly Available CD-ROM 1-USAR 087 USAR 09-13.pdf 202 KB Publicly Available CD-ROM 1-USAR 088 USAR 09-14.pdf 311 KB Publicly Available CD-ROM 1-USAR 089 USAR 10-01.pdf 177 KB Publicly Available CD-ROM 1-USAR 090 USAR 10-02.pdf 202 KB Publicly Available CD-ROM 1-USAR 091 USAR 10-03.pdf 179 KB Publicly Available CD-ROM 1-USAR 092 USAR 10-04.pdf 95 KB Publicly Available CD-ROM 1-USAR 093 USAR 10-05.pdf 91 KB Publicly Available CD-ROM 1-USAR 094 USAR 10-06.pdf 92 KB Publicly Available CD-ROM 1-USAR 095 USAR 11-01.pdf 2,729 KB Publicly Available CD-ROM 1-USAR 096 USAR 11-02.pdf 344 KB Publicly Available CD-ROM 1-USAR 097 USAR 11-03.pdf 234 KB Publicly Available CD-ROM 1-USAR 098 USAR 11-04.pdf 179 KB Publicly Available CD-ROM 1-USAR 099 USAR 11-05.pdf 174 KB Publicly Available CD-ROM 1-USAR 100 USAR 12-01.pdf 131 KB Publicly Available CD-ROM 1-USAR

1. File Name Size Sensitivity Location Folder Level 101 USAR 12-02.pdf 199 KB Publicly Available CD-ROM 1-USAR 102 USAR 12-03.pdf 180 KB Publicly Available CD-ROM 1-USAR 103 USAR 12-04.pdf 168 KB Publicly Available CD-ROM 1-USAR 104 USAR 12-05.pdf 179 KB Publicly Available CD-ROM 1-USAR 105 USAR 12-06.pdf 167 KB Publicly Available CD-ROM 1-USAR 106 USAR 12-07.pdf 166 KB Publicly Available CD-ROM 1-USAR 107 USAR 12-06.pdf 112 KB Publicly Available CD-ROM 1-USAR 108 USAR 13-01.pdf 177 KB Publicly Available CD-ROM 1-USAR 109 USAR 13-02.pdf 174 KB Publicly Available CD-ROM 1-USAR 110 USAR 13-03.pdf 169 KB Publicly Available CD-ROM 1-USAR 111 USAR 13-04.pdf 172 KB Publicly Available CD-ROM 1-USAR 112 USAR 13-05.pdf 169 KB Publicly Available CD-ROM 1-USAR 113 USAR 14-01.pdf 496 KB Publicly Available CD-ROM 1-USAR 114 USAR 14-02.pdf 232 KB Publicly Available CD-ROM 1-USAR 115 USAR 14-03.pdf 222 KB Publicly Available CD-ROM 1-USAR 116 USAR 14-04.pdf 210 KB Publicly Available CD-ROM 1-USAR 117 USAR 14-05.pdf 178 KB Publicly Available CD-ROM 1-USAR 118 USAR 14-06.pdf 254 KB Publicly Available CD-ROM 1-USAR 119 USAR 14*-07.pdf 181 KB Publicly Available CD-ROM 1-USAR 120 USAR 14-08.pdf 243 KB Publicly Available CD-ROM 1-USAR 121 USAR 14-09.pdf 251 KB Publicly Available CD-ROM 1-USAR 122 USAR 14-10.pdf 258 KB Publicly Available CD-ROM 1-USAR 123 USAR 14-11.pdf 208 KB Publicly Available CD-ROM 1-USAR 124 USAR 14-12.pdf 315 KB Publicly Available CD-ROM 1-USAR 125 USAR 14-13.pdf 246 KB Publicly Available CD-ROM 1-USAR 126 USAR 14-14.pdf 141 KB Publicly Available CD-ROM 1-USAR 127 USAR 14-15.pdf 382 KB Publicly Available CD-ROM 1-USAR 128 USAR 14-16.pdf 302 KB Publicly Available CD-ROM 1-USAR 129 USAR 14-17.pdf 181 KB Publicly Available CD-ROM 1-USAR 130 USAR 14-18.pdf 228 KB Publicly Available CD-ROM 1-USAR 131 USAR 14-19.pdf 211 KB Publicly Available CD-ROM 1-USAR 132 USAR 14-20.pdf 203 KB Publicly Available CD-ROM 1-USAR 133 USAR 14-21.pdf 169 KB Publicly Available CD-ROM 1-USAR 134 USAR 14-22.pdf 202 KB Publicly Available CD-ROM 1-USAR 135 USAR 14-23.pdf 168 KB Publicly Available CD-ROM 1-USAR 136 USAR 14-24.pdf 265 KB Publicly Available CD-ROM 1-USAR 137 USAR 15-01.pdf 93 KB Publicly Available CD-ROM 1-USAR 138 USAR 15-02.pdf 218 KB Publicly Available CD-ROM 1-USAR 139 USAR 15-03.pdf 227 KB Publicly Available CD-ROM 1-USAR 140 USAR 15-04.pdf 282 KB Publicly Available CD-ROM 1-USAR 141 USAR Appendix A.pdf 113 KB Publicly Available CD-ROM 1-USAR 142 USAR Appendix B.pdf 101 KB Publicly Available CD-ROM 1-USAR 143 USAR Appendix C.pdf 14,630 KB Publicly Available CD-ROM 1-USAR 144 USAR Appendix D.pdf 12,090 KB Publicly Available CD-ROM 1-USAR 145 USAR Appendix E.pdf 87 KB Publicly Available CD-ROM 1-USAR 146 USAR Appendix F.pdf 396 KB Publicly Available CD-ROM 1-USAR 147 USAR Appendix G.pdf 446 KB Publicly Available CD-ROM 1-USAR 148 USAR Appendix H.pdf 769 KB Publicly Available CD-ROM 1-USAR 149 USAR Appendix I.pdf 371 KB Publicly Available CD-ROM 1-USAR 150 USAR Appendix J.pdf 91 KB Publicly Available CD-ROM 1-USAR

1. File Name Size Sensitivity Location Folder Level 151 USAR Appendix K.pdf 92 KB Publicly Available CD-ROM 1-USAR 152 USAR Appendix L.pdf 98 KB Publicly Available CD-ROM 1-USAR 153 USAR Appendix M.pdf 331 KB Publicly Available CD-ROM 1-USAR 154 USARA pendix N.pdf 260 KB Publicly Available CD-ROM 1-USAR USAR Figure Section-1-USAR 155 02.pdf 11,495KB Publicly Available CD-ROM 16USAR Figure Section-1,9 BPbil vial DRM 1-USAR 5703.pdf 64K ulcyAalbeC-O USAR Figure Section-1-USAR 157 04.pdf 664 KB Publicly Available CD-ROM USAR Figure Section-1-USAR 158 05.pdf 7590 KB Publicly Available CD-ROM USAR Figure Section-1-USAR 160 06.pdf 430 KB Publicly Available CD-ROM USAR Figure Section-1-USAR 161

.d 12,955 KB Publicly Available CD-ROM 6307.pdf 477K ulcyAalbeC-O 162 USAR Figure Section-1,9 BPbil vial DRM 1-USAR 11.pdf46K ullyvilleC-M USAR Figure Section-1-USAR 163 12.pdf 4,74 KB Publicly Available CD-ROM USAR Figure Section-1-USAR 164 14.pdf 4,64KB Publicly Available CD-ROM USAR Figures Spection-1-USAR 165 A2.pdf 482 KB Publicly Available CD-ROM 16USAR Figures Section-1-USAR 168 4.pdf 1,641 KB Publicly Available CD-ROM USAR Figures Appendix-4KBPbilAvlaeCDRM 1-USAR USAR~pd FgrsApni-2310 KB Publicly Available CD-ROM 1UA 10USAR Figures Appendix-10 BPbil vial DRM 1-USAR USARdfiueApedx 360 KB Publicly Available CD-ROM 1UA NO-FC-10, Quality 2-QATR 171 Assurance Topical Report 482 KB Publicly Available CD-ROM (QATR), Revision 4

Omaha Public Power District December 9, 2015 LIC-15-01 19 10 CFR 50.71(e) 10 CFR 50.4(b)(6) 10 CFR 50.54(a) 10 CFR 50.59 10 CFR 54.37(b) 10 CFR 71.106 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Fort Calhoun Station, Unit No. 1 Renewed Facility Operating License No. DPR-40 NRC Docket No. 50-285 Fort Calhoun Station Independent Spent Fuel Storage Installation NRC Docket No.72-054 Omaha Public Power District - Fort Calhoun Quality Assurance Program Approval for Radioactive Material Packages NRC Docket No. 71-0256

Reference:

10 CFR 50.59 Report, Quality Assurance (QA) Program Changes, Technical Specification Basis Changes, 10 CFR 71.106 Quality Assurance Program Approval, Aging Management Review, Commitment Revisions, and Revision of Updated Safety Analysis Report Revision for Fort Calhoun Station (FCS), Unit No. 1

1. Letter from OPPD (L. P. Cortopassi) 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 17, 2014 (MLE141 76A236) (LIC-1 4-0077)

In accordance with 10 CFR 50.59(d)(2), the Omaha Public Power District (OPPD) submits as the report of changes, tests, and experiments performed pursuant to 10 CFR 50.59 for Fort Calhoun Station (FCS), Unit No. 1. Attachment 2 is provided to describe Quality Assurance (QA) Program changes as required by 10 CFR 50.54(a)(4)(i).

describes changes made to the quality assurance program approval for radioactive material packages. Attachment 4 contains a description of revised regulatory commitments that require Commission notification in accordance with NEI 99-04, "Guidelines for Managing NRC Commitment Changes." In accordance with FCS Technical Specification 5.20.d, Attachment 5 provides a brief summary of the Technical Specification Basis Changes (TSBCs) made since the previous submittal (Reference 1) and Attachment 6 includes a copy of the revised TSBC.

pages.

444 SOUTH 16TH STREET MALL

• OMAHA, NE 68102-2247 I',4.NSS

U. S. Nuclear Regulatory Commission LIC-15-01 19 Page 2 In accordance with 10 CFR 54.37(b), a review of structures, systems, and components (SSCs) newly identified that would have been subject to an aging management review or evaluation of time-limited aging analysis in accordance with 10 CFR 54.21 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.

This information covers the period of June 14, 2014 through December 4, 2015.

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 1) and includes changes made under the provisions of 10 CFR 50.59 but not previously submitted to the Commission.

The CD-ROM also contains Revision 4 of the Quality Assurance Topical Report (NO-FC-1 0),

incorporated by reference in the USAR. Attachment 7 contains a list of the files on the CD-ROM.

The Senior Resident Inspector is provided with an updated copy of the USAR by the FCS distribution process.

As required by 10 CFR 50.71(e)(2)i, I certify that the information in this submittal accurately presents changes made since the previous submittal, necessary to reflect information and analyses submitted to the Commission or prepared pursuant to Commission requirements, and identifies changes made under the provisions of 10 CFR 50.59 but not previously submitted to the Commission.

No commitments to the NRC are made in this letter.

If you should have any questions, please contact Mr. Bill Hansher at (402) 533-6894.

Respectuly Louis P. Cortopassi Site Vice President and CNO LPC/MLE/mle Attachments: 1. Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59

2. Quality Assurance Program Changes

3.

10 CFR 71.106 Quality Assurance Program Approval for Radioactive Material Package Changes

4. Regulatory Commitments Revised in Accordance with NEI 99-04

5.

Information Removed from the USAR

6. Summary of Technical Specification Basis Changes (TSBC)

7. TSBC Pages

8. List of Files on CD-ROM

Enclosure:

CD-ROM of USAR Sections and Figures c:

M. L. Dapas, NRC Regional Administrator, Region IV C. F. Lyon, NRC Senior Project Manager S. M. Schneider, NRC Senior Resident Inspector (w/o Enclosure)

LIC-15-01 19 Page 1 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59

LIC-1 5-0119 Page 2 Abbreviations and Acronyms:

AFW - Auxiliary Feedwater IST - In-service Testing ALARA - As Low as Reasonably Achievable LBLOCA - Large Break Loss of Coolant Accident ANSI - American National Standards Institute LCO - Limiting Conditions for Operation AOP - Abnormal Operating Procedure LOCA - Loss of Coolant Accident AOR - Analysis of Record LPSI - Low Pressure Safety Injection AR - Action Request LTOP - Low Temperature Overpressure Protection BAST - Boric Acid Storage Tank MCC - Motor Control Center BTP - Branch Technical Position MEW - Main Feedwater CCW - Component Cooling Water MH - Manhole CD-ROM - Compact Disk Read-Only Memory msl - Mean Sea Level CEA - Control Element Assembly NEI - Nuclear Energy Institute CEAPIS - CEA Position Indication System NLI - Nuclear Logistics Incorporated CFR - Code of Federal Regulations NRC - Nuclear Regulatory Commission CIV-Containment Isolation Valve NSRB - Nuclear Safety Review Board COLR - Core Operating Limits Report 01 - Operating Instruction CQE - Critical Quality Element OPPD - Omaha Public Power District CR - Condition Report PDIL - Power Dependent Insertion Limit CRS - Control Room Supervisor PORC - Plant Operations Review Committee CS -Containment Spray PRC - Plant Review Committee CW - Circulating Water PSAR - Preliminary Safety Evaluation Report DCS - Distributed Control System QA - Quality Assurance DG - Diesel Generator QATR - Quality Assurance Topical Report EA - Engineering Analysis QR - Qualified Reviewer EC - Engineering Change RCA - Root Cause Analysis EOP - Emergency Operating Procedure RCS - Reactor Coolant System ERFCS - Emergency Response Facility Computer System RFO - Refueling Outage FCS - Fort Calhoun Station, Unit No. 1 RG - Regulatory Guide FCSG - Fort Calhoun Station Guideline RPS - Reactor Protective System FSAR - Final Safety Analysis Report RSG - Replacement Steam Generators HEPA - High Efficiency Particulate Air RTD - Resistance Temperature Detector HZP - Hot Zero Power RVH - Reactor Vessel Head i&C -Instrumentation & Control RW - Raw Water IGSCC - Intergranular Stress Corrosion Cracking SARC - Safety Audit and Review Committee INPO - Institute of Nuclear Power Operation SDC - Shutdown Cooling

LIC-15-01 19 Page 3 SER - Safety Evaluation Report SSC - Structures, Systems and Components SG - Steam Generator ST - Surveillance Test SGBD - Steam Generator Blowdown TM - Temporary Modification SI - Safety Injection TS - Technical Specification SM - Shift Manager TSBC - Technical Specification Basis Change SIRWT - Safety Injection Refueling Water Tank UFSAR - Updated Final Safety Analysis Report SO - Standing Order USAR - Updated Safety Analysis Report SR - Surveillance Requirement VCT - Volume Control Tank SRP - Standard Review Plan WO - Work Order

LIC-I5-01 19 Page 4 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2014 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number EC 64651 Switchgear Supplemental Cooling Description of Activity (OI-VA-2 Compensatory Measures for R45)

Operability Evaluation 14-015 WHAT IS BEING EVALUATED BY THIS DOCUMENT During the July 2014 Inspection Manual Chapter (IMC) 0350 inspection, the NRC identified that a non-conservative design input was used for the electrical heat load in the switchgear room heat up analysis calculation, FC06102 Revision 2.

This was documented in NCV 2014009-009 in September 2014. On September 9, 2014, CR 2014-11223 was written to address immediate operability. The immediate operability determination (IOD) determined that existing supplemental cooling methods in OI-VA-2 Attachment 11 (Rev. 44) were adequate during cooler weather. These supplemental methods were already in OI-VA-2 Rev 44 specifically to address loss of normal switchgear room cooling.

To provide additional assurance of operability, Operability Evaluation 14-015 was generated. The OpEval has two compensatory measures. The compensatory measures are captured in OI-VA-2 Attachment 11 (Rev. 45) and an Adverse Condition Monitoring Plan (ACMP). The procedure change adds an improved supplemental method of ventilating the switchgear room after a loss of all normal cooling. This revised method utilizes portable fans that are staged just outside the room to blow turbine building air through the room after a loss of all cooling.

These fans are not large enough for the warmer weather/design conditions; however, they provide better flow rates and can be implemented faster than the existing supplemental methods in OI-VA-2 Revision 44. The ACMP performs daily monitoring to verify that the turbine building air temperature remains at or below 90°F. This ensures that the supply air temperature assumed for the portable fans in the OpEval remains valid.

This document evaluates the impact of the compensatory measures established

LIC-15-01 19 Page 5 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2014 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change

{Activity Title 150.59 Evaluation Summary Number_

by Operability Evaluation (OpEval)14-015. Specifically, it evaluates the impact of the changes that revision 45 makes to OI-VA-2 and the ACMP. The non-conforming condition and the adequacy of the compensatory measures to mitigate the non-conforming condition are described and evaluated in the OpEval.

BACKGROUND The following is provided as background to help set the context of this evaluation.

Brief Description of the Switchcqear Rooms and Room Coolin~q The switchgear rooms contain the electrical system normal and safety related components (e.g., 4160V buses, 480V buses, inverters, breakers, etc.) that feed power to normal power production equipment and all safety related equipment required by the Technical Specifications. Technical Specification 2.7 requires that the electrical system be operable whenever the reactor is above 300°F. USAR Section 9.10 states "The electronic equipment used in the plant safety related component can operate at 120F continuously."

The switchgear room is normally cooled by air conditioning units VA-87/89 and VA-88/90 and ventilation fans VA-41/45A/45B. The ventilation and cooling equipment associated with the switchgear room is not safety related, not completely protected from High Energy Line Breaks (HELBs), nor designed to withstand external events (e.g., seismic, flood, wind). Certain design and licensing events, like Seismic or HELB, can completely (Seismic) or partially (HELB) disable all normal switchgear room cooling. As a result, the use of temporary supplemental coolinq is discussed in USAR Section

LIC-15-O0119 Page 6 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2014 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change

[Activity Title

[.'/

';50.59 Evaluation Summary NumberI I__I_

9.10 and has been considered an acceptable backup method since the station was licensed. The electronic equipment used in the plant safety related components can operate at 1 20°F continuously per USAR Section 9.10. Therefore, the station would be outside of its design and licensing basis if supplemental cooling cannot maintain the switchgear room less than 1 20°F after a loss of all normal cooling.

Summary of Operability Evaluation 14-015 Operability Evaluation 14-015 shows that after a complete loss of all normal switchgear room ventilation and cooling, the portable fans and flow path used in the revised version of OI-VA-2 Attachment 11 (Rev. 45) are adequate to maintain the switchgear room below 120°F when supply air is at or below 90°F. The primary basis for the OpEval is the new analysis performed in FC06102 Rev.3. This calculation was revised to eliminate the nonconservative errors identified above. It provides the time available for the switchgear room to heatup under worst case design conditions and provides the minimum supplemental air flow requirements needed to keep the room below 120°F after loss of all normal cooling. In addition to the design case, the calculation assessed lesser cases (e.g., lower supplemental cooling supply air temperatures) for use in potential operability evaluations during cooler weather.

This compensatory measure added to OI-VA-2 Attachment 11 (Rev 45),

establishes air flow through the switchgear room by using portable fans to blow air from the south end of the turbine building and in through blocked open switchgear room door 1011-4. Air flow exits the north end of the room through blocked open door 1011-7. After leaving the room, a series of blocked open doors then allow the air to exhaust through the Auxiliary

LIC-15-01 19 Page 7 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2014 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number_

Building normal ventilation fans (VA-40A!B/C). If, these fans are unavailable, then the procedure provides an option to exhaust flow to the outside environment through the CARP building. Air sampling is established by Chemistry and RP if the flow is exhausted directly outside rather than through the monitored flow path provided by the VA-V-40s. The OpEval shows that after a loss of all switchgear room cooling, the time available to establish this supplemental method of cooling is 65 minutes. This is adequate time for the operators to perform the actions to implement cooling with the portable fans. The basis for the time available to perform the actions and an assessment of the time required is detailed in the Op Eval.

The procedure establishes the appropriate barrier permits (e.g., fire barrier).

However, a plant shutdown is required by the procedure since the turbine building-to-switchgear room door (1011-4) is blocked opened. This door is the High Energy Line Break (HELB) barrier between the switchgear room and the turbine building and is required as part of the design basis for HELB described in USAR Appendix M. Disabling a barriers is controlled by barrier control procedures CC-AA-201, Plant Barrier Control Program and SO-G-58, Control of Fire Protection System Impairments. The new revision to Ol-VA-2, this activity, establishes the appropriate "watches" using these barrier control procedures.

An Adverse Condition Monitoring Plan (ACMP) was described in and is being issued as part of the Operability Evaluation. It provides specific guidance for periodically checking turbine building temperature to ensure that it remains at or below 90°F. This verifies that the assumptions in the Operability Evaluation remain valid. Temperatures in this area of the turbine building are currently

('October 2014) around 75°F and typically well below 90°F during the late fall

LIC-15-01 19 Page 8 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2014 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number_

through early spring months.

This operability evaluation can be closed when corrective actions are completed to replace the portable fans with larger fans and Ol-VA-2 Rev 45 is revised to reflect the larger fans. The larger fans will provide enough airflow to cool the switchgear rooms under design basis conditions, Case 5D, of FC06102 Rev 3.

Reason for Activity This document evaluates the impact of the compensatory measures established by Operability Evaluation (OpEval)14-015. Specifically, it evaluates the impact of the changes that revision 45 makes to OI-VA-2 and the impact of the ACM P.

The procedure change adds an improved supplemental method of ventilating the switchgear room after a loss of normal cooling. This revised method utilizes portable fans that are staged just outside the room to blow turbine building air through the room after a loss of all cooling. These fans are not large enough for the warmer weather/design conditions; however, they are adequate for conditions where the supply air temperature (turbine building) is below 90°F. The new method provides better flow rates and can be implemented faster than the existing supplemental methods in OI-VA-2 Revision 44.

The ACMP adds a requirement for plant operations to check turbine building air temperature once per day, makes adjustments to allow more outside air into the turbine building if temperature rises above 85°F, and reassess

LIC-15-01 19 Page 9 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2014 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number operability of the components in the switchgear room if temperature rises above 900°F. The impact of these two compensatory measures on other SSCs and the plant design and licensing basis is reviewed under 50.59.

Effect of Activity This activity consists of (1) The procedure change, OI-VA-2 Rev. 45, which adds an improved supplemental method of ventilating the switchgear room after a loss of normal cooling and (2) the ACMP which monitors turbine building temperature to ensure it remains at or less than 90°F. These activities are evaluated for the impact that they have on plant operations, the design basis, or safety analysis described in the UFSAR. The specific item that potentially impact operations, the design bases, or safety analysis described in the UFSAR are addressed in the 50.59 and are as follows:

1. Defeating/Opening HELB and Fire Barrier Doors

2. Adding Load to the Diesel Generators

3.

Moving warm air at <120°F into the Aux Building from the Switchgear Room

4. The physical force/effect of blowing air into the switchgear room on switchgear room components

5.

Exhausting Auxiliary Building Air through the CARP Building (unmonitored release path)

6. Adding Operator Actions that are Time Critical Actions (TCAs)

Summary of Conclusion for the Activity's 50.59 Review This activity requires a 50.59 Screening. That screening indicated that a

LIC-15-0119 Page 10 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2014 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 EvalUation Summary Number.

50.59 Evaluation was required because the activity involves a change to an SSC that may adversely affect UFSAR described design functions.

Specifically, the compensatory measures required to enhance switchgear room supplementary cooling make changes to the Auxiliary Building door alignments that breach USAR described barriers and adds load to the diesels.

The conclusion of the 50.59 Evaluation was that NRC prior approval is not required and there are no required changes to the Technical Specifications.

This conclusion is based on a review of the key activities implemented by the compensatory measures in OI-VA-2 Rev 45 and the ACMP against the eight 50.59 Evaluation questions. Those key activities are:

1. Defeating/Opening HELB and Fire Barrier Doors.

2. Adding Load to the Diesel Generators

3. Moving warm air at <120°F into the Aux Building from the Switchgear Room

4. The physical force/effect of blowing air into the switchgear room on switchgear room components

5. Exhausting Auxiliary Building Air through the CARP Building (unmonitored release path)

6. Adding Operator Actions that are Time Critical Actions (TCAs)

EC 60821 Emergency RCS Fill Connections Activity Description

- FLEX Activity No. 1 This modification installs FLEX Emergency RCS Fill Connection lines into the CVCS system and SI system (Containment Spray (CS) and Shutdown

LIC-1 5-0119 Page 11 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2014 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change

[Activity Title

[50.59 Evaluation Summary Number

_I_

Cooling (SDC) piping) to allow fo netinit the RCS in accordance with the FLEX strategy. The connection line in the CH system is installed at the 1" blind flange located at the discharge of the CH-1A pump and the connection line for the SI system is installed at the 3" blind flange located on the Containment Spray (CS) header upstream of air-operated valve HCV-344.

Both connection lines consist of two manual gate valves, piping, a reducer, a Female National Pipe Thread (FNPT) to Female National Standard Thread (FNST) adapter, and a Male National Standard Thread (MNST) plug. The connection lines are prepared as spool pieces, which will be bolted on at the flange connection after the blind flanges are removed. This modification also installs a pipe support between the two gate valves for each connection line.

The new components installed as a result of this modification are CH-576, CH-577, CHSP-95, SI-51 8, SI-51 9, and SIS-243.

Activity No.2 The CVCS, CS, SDC and SI systems are Safety Class 2 systems. They were originally designed and analyzed to USAS B31.7 (Draft 1968 Edition, Code of Record). The design of the connection lines (excluding the pipe supports) conforms to the requirements of ASME Section III, 1980 edition, with Summer 1981 Addenda. Designing according to ASME Section III is consistent with current Fort Calhoun design practices described In PED-MSS-1 1 "Design Specification for Piping and Pipe Supports" as reconciled by Engineering Analysis EA-FC-91-054. This represents a methodology change from the USAR described analysis. A license amendment request (LAR 14-04, LIC-14-0043) has been submitted to allow future use of ASME Section III 1980 Edition ('no Addenda) as an alternative to B31.7, 1968 Draft for pipe

LIC-15-01 19 Page 12 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2014 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title

[50.59 Evaluation Summary Number stress analysis on non-reactor coolant safety systems, without additional reconciliation.

The 50.59 screening for the proposed activities concluded that only Activity No.2 requires a 50.59 evaluation because it represents a change in the USAR described methodology and is therefore subject to this evaluation.

Summary of Evaluation The use of ASME Section III 1980 Edition with 1981 Summer Addenda as reconciled to B31.7, 1968 Draft for pipe design and analysis is a change from the USAR methodology of using B31.7, 1968 Draft. This is a methodology change only; therefore, it does not impact the frequency of USAR analyzed accidents, the likelihood of USAR analyzed component malfunctions, the consequence of USAR analyzed accidents, or the consequence of USAR analyzed component malfunctions. In addition, the methodology change does not create new accidents types or malfunctions other than those analyzed in the USAR. The methodology change also does not change USAR design basis limits for fission product barriers.

Previous reconciliation documented in Engineering Analysis EA-FC-91-054 and additional reconciliation of applicable requirements not addressed in EA-FC-91-054 provide the basis for the conclusion that the change in methodology is equivalent. No USAR design functions are changed as a result of this modification. The conclusion of this evaluation is that a License Amendment is not required.

LIC-1 5-0119 Page 13 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change

[Activity Title 50.59 Evaluation Summary Number

[________________

EC 65174 Condensate Return Line Caps in Description of Activity Turbine Building The proposed activity temporarily modifies a portion of the Condensate Return piping in the Turbine building basement. This piping supports the return of auxiliary steam condensate to the drip and drain tank. The temporary modification is being completed to isolate the Auxiliary Building Condensate Return piping from the Condensate Return piping in the reminder of the Turbine Building which includes the Condensate Return line from the Intake Structure. Additionally a sample sink drain line attached to the piping being isolated requires extension so that it will drain into piping that will be in service after installation.

Reason for Activity The potential for High Energy Line Cracks (HELC's) in the Condensate Return piping poses a threat to the operation of Auxiliary Feedwater pumps FW-6 and FW-1 0 in Room 19 located in the Auxiliary Building at Elevation 989.' This activity temporarily isolates the Condensate Return line in the Auxiliary Building from the remainder of the system so Auxiliary Steam can be returned to service in the Intake Structure and CST without the HELC concern in Room 19.

Effect of Activity The Condensate Return line being modified in this activity is a high energy line and part of the Auxiliary Steam System (UFSAR Section 9.10). This modification will separate the Auxiliary Building and Turbine Building portions of the condensate return line. This will prevent the Auxiliary Building portion

LIC-15-01 19 Page 14 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change

[Activity Title T50.59 Evaluation Summary Number

_ I of the condensate return line from being used until a permanent tie in is made but will allow for the use of all portions of the Turbine Building condensate return line including the return from the intake structure. This result has an adverse effect on the ability of the Auxiliary Building HVAC system to heat desired spaces because Auxiliary Steam will not be available. Extension of the sample sink drain line has no impact on the plant other than installation of additional piping.

Summary of Conclusion for the Activity's 50.59 Review A 50.59 review was performed for the activities described in this modification.

The 50.59 applicability determination shows that the activities are not controlled by any of the processes considered in LS-AA-1 04-1002. As such, a 50.59 screen was performed. The 50.59 screen is attached (LS-AA-1 04-1003). The screen shows that the proposed activity adversely affects a USAR described design function and a procedure which performs or controls this design function. Therefore, a 50.59 Evaluation was performed to evaluate the effect of the proposed activity on accidents and malfunctions previously evaluated in the UFSAR and the potential to cause accidents or malfunctions whose effects are not bounded by previous analyses. The 50.59 Evaluation determined that the proposed activity could be implemented without prior NRC approval.

EC 66460 Install Gag on Relief Valve AC-168

-Description of Activity A gag is being installed on the Component Cooling Water (CCW) inlet thermal relief valve for Reactor Coolant Pump (RCP) RC-3C seal cooler, AC-168.

LIC-15-O0119 Page 15 Changes, Tests, and Experiments Performed Pursuant to 10 CER 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change

[Activity Title T-50.59 Evaluation Summary Number_

Reason for Activity AC-I168 is being gagged shut due to leakage concerns. Temporarily gagging AC-i168 will ensure its pressure boundary is maintained on its branch loop of the CCW system.

Effect of Activity The activity has been evaluated and found to have no adverse effect. The intent of the activity is to ensure that the USAR described functions of CCW system are maintained. Relief valve AC-i168 provides a thermal relief function for the protection of the seal cooler for RC-3C and its surrounding CCW piping. The thermal relief function is required should the cooler be isolated for maintenance, or other concern. AC-168 is isolated by closing manual valve AC-260 and control valve HCV -444. If a thermal transient occurs during the time when the relief valve is isolated, the CCW pipe and / or cooler could rupture. Loss of cooling to the seal would occur and a breach in the CCW system pressure boundary would exist. The breach would cause loss of CCW system inventory. If the branch loop containing AC-I168 were to be un-isolated by opening control valve HCV-444, or manual valve AC-260, a greater loss of CCW inventory would occur. That being said, it is not a reasonable concern to assume that the cooler would be isolated at a time where a thermal transient would occur that would rupture the CCW branch loop. It would mean that maintenance was being conducted during operation or the loop was isolated during operation. Neither scenario is likely, if ever, to occur.

Gagging AC-168 closed when AC-260 and HCV-444 are open is not a

LIC-15-01 19 Page 16 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number concern because if a transient occurs the resulting pressure would expand out and open another relief valve in the system.

There is no impact on plant operations, design bases, or safety analyses described in the UFSAR.

Summary of Conclusion for the Activity's 50.59 Review The 50.59 evaluation concludes that the proposed activity does not have any adverse impact on any USAR described functions or how any USAR described functions are controlled. Therefore, the proposed activity may be implemented without NRC prior approval.

The gagging of AC-i168 does not adversely affect the function, operation, or control of the CCW system. It does not affect any other plant system. No automatic system functions, other than the relief function of the valve, are being affected. No procedures are being changed. Therefore, UFSAR described design functions are not adversely affected, and, therefore, how UFSAR described design functions are performed or controlled are not adversely affected. The gagging of the valve does not involve a revision, or replacement, of an evaluation methodology. As such, neither the design basis nor the safety analyses are affected. The gagging of the valve does not involve a new test or experiment. As such, no SSC is utilized or controlled in an adverse manner. The activity is not controlled by the processes of the Applicability Review. The activity requires a 50.59 Evaluation under the 50.59 Screening.

LIC-15-01 19 Page 17 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized result, the language may be in future tense.

below are for the most part, unedited summaries as approved by the PORC. As a Change J Activity Title

'_50.59 EvalUation Summary EC 64574 Address EDG Load Shedding Concerns in Fire Area 36A by Rerouting Cables and Adding Coordinated Fuses and Isolation Switches (REC-1 18) (Ref NFPA-805 Transition LAR LIC-1 1-0099 Table S-2 Description of Activity Electrical isolation is required to preclude fire induced cable damage from causing a failure of breaker control circuits. These cable failures may introduce a direct short to ground (or hot short), with a potential to fail breaker I25VDC TRIP control power or cause uncontrolled, spurious operation of a breaker.

The scope of EC 64574 includes the following activities:

1. For the breaker 1 B4A-7 (480 VAC MCC-4A3) control circuit, cable 7700B will be re-routed from Tray Section 57S to Panel AI-109B in the existing Pyrocrete enclosure. The breaker TRIP control circuit will be electrically isolated from a fault on the Train A load shed input cable (cable 7700A to AI-109A) by installing a fuse and a blocking diode in Panel Al-i109B.

2. For the breaker 1 B4C-1 (480 VAC MCC-4C5) control circuit, cable Bill183 will be re-routed and a fuse will be installed on the conductor from Panel Al-i109B to Panel Al-i109A. The breaker TRIP control circuit will be electrically isolated from faults induced in the non-scheduled cable between Al-i109A and Al-i109B by installing a fuse and a blocking diode in Panel AlI-109B.

3.

For Containment Spray pump SI-3C breaker I1B3B-4B-3, an isolation switch will be installed in a new junction box mounted on the west wall within Fire Area 36B. This switch will isolate the breaker control circuit from all Train A automatic load shed and load sequencing signals to prevent a spurious closing of the breaker in response to fire induced faults in Fire Area 36A. The switch will be used to normally

LIC-15-01 19 Page 18 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number I ____________________________________________________________________

Change ActivitY Title 50.59 Evaluation Summary Number disconnect the Al-i108A conductors from the breaker trip circuit when 1 B3B-4B-1 is aligned to Train B.

4. For VA-i151 B, cable B2998 will be re-routed from Tray Section 57S to Panel Al-i109B in the existing Pyrocrete enclosure.

5.

For VA-i151 D, cable B3000 will be re-routed from Tray Section 57S to Panel AlI-109B in the existing Pyrocrete enclosure.

6. For EHC-3B, cable B8970 will be re-routed in a conduit from the Al-109B enclosure directly through the wall into the Turbine Building, rather than running the conduit through unprotected Tray Section 56S in Fire Area 36A.

7.

For ST-6B, cable B2630 will be re-routed in a conduit from the Al-1 09B enclosure directly through the wall into the Turbine Building, rather than running the conduit through unprotected Tray Section 56S in Fire Area 36A.

Reason for Activity In 2010, Omaha Public Power District (OPPD) Fort Calhoun Station (FCS),

Unit No. 1 submitted License Amendment Request (LAR) 10-07 (Ref. LIC-1 1-0099) to the Nuclear Regulatory Commission (NRC) to amend the Operating License, No. DPR-40, to adopt the National Fire Protection Association (NFPA) 805, Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition) standard.

As part of the NFPA 805 transition process, an outside, third-party was contracted to perform a Nuclear Safety Capability Assessment (NSCA), a Non-power Operations (NPO) Assessment, and a Fire Probabilistic Risk Assessment (Fire PRA) to evaluate the plant response to deterministic and

LIC-15-01 19 Page 19 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change 1Activity Title 150.59 Evaluation Summary Number realistic fire events. These assessments were documented in EPM Report R2008-004-002, early in 2015 (Ref. EPM Report EA-10-036).

The result of this report determined that a number of modifications are required to be performed to bring the plant configuration into compliance with the NFPA 805 standard. The report identified Recommended Engineering Change (REC) REC-1 18 which was committed to be performed in LIC-1 1-.

0099, Attachment S (Table S-2). It is the purpose of EC 64574 to implement REC-118.

On June 16th, 2014, the NRC approved the license amendment via letter NRC 14-0072, to be implemented on June 16th, 2015 with the contingency that the committed to engineering changes were implemented at the station (Ref. SE ML14098A092).

Effect of Activity The scope of EC 64574 involves the re-routing of control circuit cables, installation of an isolation switch in the control circuit for SI-3C and the installation of fuses and blocking diodes, in order to prevent spurious breaker operation or a loss of breaker TRIP capability in the event of a fire in Fire Area 36A. The addition of the diode to the circuit is a reduction in reliability because it is a semiconductor device with a limited shelf and service life. An open circuit of the diode will isolate one load shed train. A short circuit of the diode will have no impact on the design functions. There is no change in how any components are operated or controlled.

LIC-15-01 19 Page 20 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 150.59 Evaluation Summary Number

__r_

Summary of Conclusion for the Activity's 50.59 Review EC 64574 is being implemented per LAR 10-07 to adopt the NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition) standard. This configuration change is outside the previously performed 50.59 Screening, which was performed under EC 50741 in conjunction with transition to the NFPA 805 program. Due to the change in plant configuration as a result of this EC, a 50.59 Screening is performed for this activity in order to ensure that no adverse impacts to the plant USAR or technical specifications occur.

Results of the screening determined that the addition of the diode is an adverse activity due to impacting the reliability of the load shed circuit. The remainder of the activities are not adverse because design functions are not impacted and will be controlled in the same manner as they currently are.

The evaluation has determined that prior NRC approval is not required because the diode is significantly more reliable than the other components in the circuit. No accidents are impacted and all failure modes are bound by the existing malfunctions.

EC 58161 Modification to Provide Electrical Description of Activity Isolation for Breaker Trip Control Circuits Associated with 4kV Circuit The scope of EC 58161 includes installing a fuse both locally at each breaker Breakers (REC-1 12) (See Table 5-(see below), and within the control board console CB-1 0/11. The new fuses 2 in LIC-1 1-0099, NFPA-805 will protect and ensure continued availability of the 125VDC control power to Transition LAR) each breaker's TRIP circuit. Fusing the TRIP control cables will isolate any faults, allowing each breaker to be tripped. A blocking diode will be installed within each control circuit to prevent any shorts to ground from grounding the TRIP signals.

LIC-15-01 19 Page 21 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number Fault protection will be provided by installing fuses, blocking diodes, and performing circuit re-wiring to the following 4160 V breaker control circuits:

1A1-0, Fire Pump FP-1A 1A1-1, Heater Drain Pump, FW-5A 1A1-2, S/G Feed Pump, FW-4A 1A1-3, Condensate Pump, FW-2A 1A1-4, Circulating Water Pump, CW-IA 1A2-6, Circulating Water Pump, CW-1B 1A2-7, Condensate Pump, FW-2B 1A2-8, S/G Feed Pump, FW-4B 1A2-9, Heater Drain Pump, FW-5B 1A4-3, Circulating Water Pump, CW-lC 1A4-4, Heater Drain Pump, FW-5C 1A4-5, S/G Feed Pump, FW-4C 1A4-6, Condensate Pump, FW-2C Reason for Activity In 2010, Omaha Public Power District (OPPD) Fort Calhoun Station (FCS),

Unit No. 1 submitted License Amendment Request (LAR) 10-07 (Ref. LIC-1 1-0099) to the Nuclear Regulatory Commission (NRC) to amend the Operating License, No. DPR-40, to adopt the National Fire Protection Association (NFPA) 805, Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition) standard.

As part of the NFPA 805 transition process, an outside, third-party was

___________________________________contracted to perform a Nuclear Safety Capability Assessment (NSCA), a

LIC-15-01 19 Page 22 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title, 50.59 Evaluation Summary Number Non-power Operations (NPO) Assessment, and a Fire Probabilistic Risk Assessment (Fire PRA) to evaluate the plant response to deterministic and realistic fire events. These assessments were documented in EPM Report R2008-004-002, early in 2015 (Ref. EPM Report EA-10-036).

The result of this report determined that a number of modifications are required to be performed to bring the plant configuration into compliance with the NFPA 805 standard. The report identified Recommended Engineering Change (REC) REC-1 12; which was committed to be performed in LIC-1 1-0099, Attachment S (Table S-2). It is the purpose of EC 58161 to implement REC-1 12.

On June 16th, 2014, the NRC approved the license amendment via letter NRC 14-0072, to be implemented on June 16th, 2015 with the contingency that the committed to engineering changes were implemented at the station (Ref. SE ML14098A092).

Effect of Activity The fuses have been evaluated and determined to meet electrical requirements for use in the identified 4160 V breaker TRIP circuits. The fuses were also determined to meet seismic requirements for installation in control board panel CB-1 0/11. The circuit re-wiring will result in no change to pump control or operation. The introduction of the diode introduces a new failure mode of the circuit where local trip control can be lost without tripping the breaker. The EC will prevent a loss of each breaker's TRIP capability in the event of a fire in Fire Areas 31, 46 and 47.

LIC-15-0119 Page 23 Changes, Tests, and Experiments Performed Pursuant to 10 CER 50.59 201 5 Evaluations Chnotg heA10cFRit 5059evluationsEvasummarizedar reultbterlnug a

ei uuetne below are for the most part, unedited summaries as approved by the PORC. As a Change Activity Title J 50.59 Evaluation Summary Number I

Summary of Conclusion for the Activity's 50.59 Review EC 58161 is being implemented per LAR 10-07 to adopt the NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition) standard. This configuration change is outside the previously performed 50.59 Screening, which was performed under EC 50741 in conjunction with transition to the NFPA 805 program. Due to the change in plant configuration as a result of this EC, a 50.59 Screening is performed for this activity in order to ensure that no adverse impacts to the plant USAR or technical specifications occur.

The screening determined that the SSC design functions are adversely affected due to the installation of the blocking diode creating a new failure mechanism and mode. An evaluation was performed and determined that prior NRC approval is not required. The impact on design function in not significant because the reliability of the diode far exceeds the other limiting components in the loop and all new failure modes are bound by the existing evaluations in the USAR.

EC 66825 Inadvertent Digital Upgrade of Description of Activity Degraded Voltage Relays (OPLS)

Fort Calhoun Station (FCS) was notified that Allen-Bradley 700-RTC Time Delay Relays had been transitioned from a solid-state to a complex digital based device without part number change or notification from the manufacturer. These digital relays were previously installed as like-for-like replacements on three of the four Engineered Safeguard Features (ESF) channels for the Offsite Power Low Signal (OPLS). Because the relays were thought to be like-for-like; they were installed as a maintenance activity without a 10 CFR 50.59 review. The station has decided to correct this non-

LIC-15-01 19 Page 24 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change 1Activity Title 50.59 Evaluation Summary.

Number conformance by accepting the Condition and incorporating the necessary changes into the design and licensing basis. This 50.59 review will determine if a license amendment is required to accept this change.

Reason for Activity In August 2014, United Controls supplied four (4) time delay relays to Omaha Public Power-Fort Calhoun Station. The subject relay was qualified in accordance with IEEE 323-74/83, IEEE 344-1 975/1 987, and IEEE C37.98-1987, for use in mild environment safety related applications.

Per NRC Part 21 notifications, UCI was informed that the Allen Bradley relays base model 700RTC contain a Complex Programmable Logic (CPLD) which was unpublished. This design change could not be noticed since the external appearance of the relay and the relay part number remained the same.

Hence, UCI has qualified the subject relay as solid state relay whereas the presence of the CPLD device elects the item as a digital device which can be affected by EMI/RFI noises.

Due to the event described above, the station has declared a non-conformance on the installed relays.

Effect of Activity The affected relays are 27-T1/OPLS-A, 27-TI/OPLS-C, and 27-TI/IOPLS-D.

These relays provide the time delay function for the degraded voltage control relays. The OPLS signal is used to ensure offsite power voltage levels are sufficient to start and operate all reauired electrical loads in the event of a

LIC-15-O0119 Page 25 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change 1

Activity 'Title J 50.59 EvalUation Summary Number design basis accident (DBA). The three affected relays are located on the stations 4160 VAC electrical distribution buses. If an undervoltage condition is detected, then the control relays actuate; which in turn actuates the time delay relays. If voltage does not recover within the time delay period, then the time delay relays contacts are actuated and the electrical buses are transferred to the emergency diesel generators (EDG).

With the exception of being digital devices, the installed relays are functionally equivalent to the solid-state relays that were replaced. Therefore, the effect of the activity is limited to the software considerations as described in NEI 01-01, Guideline on Licensing Digital Upgrades EPRI TR-1 02348 Revision 1.

Summary of Conclusion for the Activity's 50.59 Review No other processes where identified by the Applicability Determination. The Screening determined that there is an adverse impact due to the possibility of software common cause failure. Since the physical replacement of the relay was previously screened and there are no human interface concerns with a relay; none of the other criteria were applicable. The Evaluation determined that prior NRC approval is not required because the function of the relay is extremely simple and has a strong pedigree to support reliable operation.

EC 66490 Add a Varistor Across HCV-1 041A-Description of Activity 20A & HCV-1042A-20A This EC will install a varistor across the terminals of HCV-1041A-20A (Main Steam Valve HCV-1041A Pilot Solenoid Valve) and HCV-1042A-20A (Main Steam Valve HCV-1042A Pilot Solenoid Valve).

LIC-15-01 19 Page 26 Changes, Tests, and Experiments Performed Pursuant to 10 CER 50.59 2015 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number Reason for Activity Varistors will be installed across the pilot (opening) solenoid coils for HCV-1041A and HCV-1042A to arrest induced voltage spikes. Based on troubleshooting associated with CR 2015-07217, spikes greater than the PC-11 5A and PC-I118A setpoint of 17.5 MA for the trip of shutdown cooling were discovered during the cycling of both HCV-1041A and HCV-1042A. Installing varistors for the purpose of arresting voltage spikes is a common industry practice.

Effect of Activity HCV-1041A and HCV-1042A are Main Steam Isolation Valves (MS IVs) used to isolate the steam generators from the main steam header. The valves are provided to isolate the steam generators during normal and accident conditions. While the addition of the varistor to the control loop will improve performance of the external loops affected by the EMI spike, it also reduces the reliability of the 1041A and 1042A control loops by increasing the probability of an existing failure mechanism (i.e. shorted varistor, shorted solenoid). In general, failure of HCV-1041A and HCV-1042A, regardless of the initiating failure mechanism, could result in the inability to control an excessive reactor coolant system cooldown rate and resultant reactivity insertion following a main steam break incident.

Summary of Conclusion for the Activity's 50.59 Review The Applicability Review determined that no other regulatory processes are applicable. The screening has determined that there may be an adverse

LIC-15-01 19 Page 27 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number impact to the MSIV design functions due to the addition of components to the control loops. The evaluation determined that prior NRC approval is not required because the addition of varistors does not fulfill any of the 10 CFR 50.59 criteria because the reduction in reliability is minimal and no new failure modes are created (i.e. no new accidents or malfunctions result).

EC 85749 increasing Portable Fan Size and Description of Activity (OI-VA-2 Changing the Flow Path for R47)

Switchgear Room Supplemental WHAT IS BEING EVALUATED BY THIS DOCUMENT Cooling During the July 2014 Inspection Manual Chapter (IMC) 0350 inspection, the NRC identified that a non-conservative design input was used for the electrical heat load in the switchgear room heat up analysis calculation, FC06102 Revision 2. This was documented in NCV 2014009-009 in September 2014. On September 9, 2014, CR 2014-11223 was written to address immediate operability. The immediate operability determination (IOD) determined that existing supplemental cooling methods in Ol-VA-2 1 (Rev. 44) were adequate during cooler weather, but inadequate during warmer weather/design conditions. These supplemental methods were already in OI-VA-2 Rev 44 specifically to address loss of all normal switchgear room cooling.

To provide additional assurance of operability at higher ambient temperatures, Operability Evaluation 14-015 was generated and Procedure OI-VA-2 Revision 45 was issued. OI-VA-2 Revision 45 provided additional detail on using existing portable fans to supply air from the turbine building through the switchgear room. The OpEval showed that these fans could maintain the room below 1200F after a loss of all normal cooling at supply air temperatures below 900F. Use of these fans was an interim solution until a larger portable fan could be obtained.

LIC-15-01 19 Page 28 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORO. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number This document evaluates changes to Procedure Ol-VA-2 Attachment 11, Calculation FC061 02, and a USAR change to Section 9.10. These documents are being issued under EC 65749. Specifically, the changes are:

OI-VA-2 Attachment 11 Rev 45 is being revised to o

Utilize a new large capacity portable fan to provide switchgear room supplemental cooling in the event that all normal cooling is lost. The new portable fan replaces the smaller existing portable fans and is stored in an existing storage cage in the switchgear room.

o Change the supplemental cooling flow path to take air from the south end of the turbine building and return it to the North end.

The existing procedure revision discharges air out through the RCA portion of the Auxiliary Building. This new path is preferable since it eliminates breaching the RCA barrier.

FC06102 Rev 3 is being revised to:

o Evaluate the larger capacity portable fan and revised air flow path to show that it will maintain switchgear room temperature below the USAR required temperature of 120°F using supply air sources (turbine building or outside air) that are at or below the USAR design temperatures. The USAR design temperature is 95°F for outside air (USAR Section 9.10.1) and 1 05°F for turbine building air (USAR Section 9.10, Table 9.10-1).

o Eliminate GOTHIC runs/cases from the calculation that are not utilized (e.g., supplying air from the diesel rooms).

USAR Section 9.10 page 7 is being revised to clarify what "supplementary cooling" entails. This has always consisted restarting existing fans or use of portable fans; however, that was never clearly defined in the USAR nor

LIC-15-01 19 Page 29 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change 1Activity Title 1

"50.59 Evaluation Summary Number in any previous design basis documents or NRC correspondence. The changes clarify that supplementary cooling includes a portable fan sized to maintain the room below 120°F after losing all normal cooling.

After implementation of these documents through EC 65749, OpEval 14-15 will be closed.

BACKGROUND The following is provided as background to help set the context of this evaluation.

Brief Description of the Switchgear Rooms and Room Coolincq The switchgear rooms contain the electrical system normal and safety related components (e.g., 4160V buses, 480V buses, inverters, breakers, etc.) that feed power to normal power production equipment and all safety related equipment required by the Technical Specifications. Technical Specification 2.7 requires that the electrical system be operable whenever the reactor is above 300°F. USAR Section 9.10 states "The electronic equipment used in the plant safety related component can operate at 120F continuously."

The switchgear room is normally cooled by air conditioning units VA-87/89 and VA-88/90 and ventilation fans VA-41/45A/45B. The ventilation and cooling equipment associated with the switchgear room is not safety related, not completely protected from High Energy Line Breaks (HELBs), nor designed to withstand external events (e.g., seismic, flood, wind). Certain design and licensing events, like Seismic or HELB, can completely (Seismic) or partially (HELB) disable all normal switchgear room cooling. As a result,

LIC-15-01 19 Page 30 Changes, Tests, and Experiments Performed Pursuant to 10 CER 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change

[Activity Title 50.59 Evaluation Summary Number L

the use of temporary supplemental cooling is discussed in USAR Section 9.10 and has been considered an acceptable backup method since the station was licensed. The electronic equipment used in the plant safety related components can operate at 120°F continuously per USAR Section 9.10. Therefore, the station would be outside of its design and licensing basis if supplemental cooling cannot maintain the switchgear room less than 1200F after a loss of all normal cooling.

Summary of Operability Evaluation 14-015 Operability Evaluation 14-015 shows that after a complete loss of all normal switchgear room ventilation and cooling, the portable fans and flow path used in the revised version of OI-VA-2 Attachment 11 (Rev. 45) are adequate to maintain the switchgear room below 120°F when supply air is at or below 90°F (later revised to 98°F). The primary basis for the OpEval is the analysis performed in FC06102 Rev.3. This calculation was revised to eliminate the non-conservative errors. It provides the time available for the switchgear room to heatup under worst case design conditions and provides the minimum supplemental air flow requirements needed to keep the room below 1 20°F after loss of all normal cooling. In addition to the design case, the calculation assessed lesser cases (e.g., lower supplemental cooling supply air temperatures) for use in potential operability evaluations during cooler weather. The OpEval was written because the portable fan used in OI-VA-2 1 is undersized and therefore, may not have maintain the room below 1 20°F with supply air from the turbine building at the USAR design temperature of 105°F.

I ________________________________________

L

LIC-15-01 19 Page 31 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number Sum mary of the Existincq Version of OI-VA-2 Attachment 11 ( Rev 45)

OI-VA-2 Attachment 11 Rev 45, previously evaluated under 50.59, establishes air flow through the switchgear room by using portable fans to blow air from the south end of the turbine building and in through blocked open switchgear room door 1011-4. Airflow exits the north end of the room through blocked open door 1011-7. After leaving the room, a series of blocked open doors then allow the air to exhaust into the north end of the turbine building. Calculation FC06102 Rev. 4 shows that after a loss of all switchgear room cooling, the time available to establish this supplemental method of cooling is 65 minutes. This is adequate time for the operators to perform the actions to implement cooling with the portable fans. The basis for the time available to perform the actions and an assessment of the time required was previously evaluated.

The procedure establishes the appropriate barrier permits. However, a plant shutdown is required by the procedure since the turbine building-to-switchgear room door is blocked opened. These doors are High Energy Line Break (HELB) barriers between the switchgear room and the turbine building and required as part of the design basis for HELB described in USAR Appendix M. Disabling barriers is controlled by barrier control procedures CC-AA-201, Plant Barrier Control Program and SO-G-58, Control of Fire Protection System Impairments.

Reason for Activity This activity is being performed to complete the corrective actions necessary to close Operability Evaluation 14-015. The actions include changes to Procedure OI-VA-2 Attachment 11, Calculation FC061 02, and a USAR

LIC-1 5-0119 Page 32 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change 1Activity Title 50.59 Evaluation Summary Number change to Section 9.10. These documents are being issued under EC 65749. Specifically, the changes are:

OI-VA-2 Attachment 11 Rev 45 is being revised to o

Utilize a new large capacity portable fan to provide switchgear room supplemental cooling in the event that all normal cooling is lost. The new portable fan replaces the smaller existing portable fans and is stored in an existing storage cage in the switchgear room.

o Change the supplemental cooling flow path to take air from the south end of the turbine building and return it to the North end.

The existing procedure revision discharges air out through the RCA portion of the Auxiliary Building. This new path is preferable since it eliminates breaching the RCA barrier.

FC061 02 Rev 3 is being revised to:

o Evaluate the larger capacity portable fan and revised air flow path to show that it will maintain switchgear room temperature below the USAR required temperature of 120°F using supply air sources (turbine building or outside air) that are at or below the USAR design temperatures. The USAR design temperature is 95°F for outside air (USAR Section 9.10.1) and I105°F for turbine building air (USAR Section 9.10, Table 9.10-1).

oEliminate GOTHIC runs/cases from the calculation that are not utilized (e.g., supplying air from the diesel rooms).

USAR Section 9.10 page 7 is being revised to clarify what "supplementary cooling" entails. This has always consisted restarting existing fans or use of portable fans; however, that was never clearly defined in the USAR nor in any previous design basis documents or NRC correspondence. The changes clarify that supplementary cooling includes a portable fan sized

LIC-15-01 19 Page 33 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title

• 50.59 Evaluation Summary Number to maintain the room below 120°F after losing all normal cooling.

Effect of Activity This activity consists of a change to the supplementary cooling method used in OI-VA-2 Attachment 11. The revised procedure will utilize a larger portable fan and a different air flow path. The change in flow path exhausts air back to the north end of the turbine building instead of outside through the RCA portion of the auxiliary building. The larger fan applies a larger load (20 hp) on the diesels. The analysis that supports this work is FC06102 Rev 4 and EC 65749.

USAR section 9.10 is revised to add clarity to the discussion on switchgear room supplementary cooling; however, the USAR change does not affect the safety analyses described in the USAR.

This activity is evaluated for the impact that it has on plant operations, the design basis, or safety analysis described in the UFSAR. The specific items that potentially impact operations, the design bases, or safety analysis described in the UFSAR are addressed in the 50.59 and are as follows:

1. Defeating/Opening HELB and Fire Barrier Doors

2. Adding Load to the Diesel Generators

3. The physical force/effect of blowing air into the switchgear room on switchgear room components

4. Adding Operator Actions that are Time Critical Actions (TCAs)

LIC-15-01 19 Page 34 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number Summary of Conclusion for the Activity's 50.59 Review This activity requires a 50.59 Screening. That screening indicated that a 50.59 Evaluation was required because the activity involves a change to an SSC that may adversely affect UFSAR described design functions.

Specifically, the switchgear room supplementary cooling makes changes to the Auxiliary Building door alignments that breach USAR described barriers and adds load to the diesels.

The conclusion of the 50.59 Evaluation was that NRC prior approval is not required and there are no required changes to the Technical Specifications.

This conclusion is based on a review of the proposed changes to Ol-VA-2 against the eight 50.59 Evaluation questions. Those key activities are:

1. Defeating/Opening HELB and Fire Barrier Doors.

2. Adding Load to the Diesel Generators

3. The physical force/effect of blowing air into the switchgear room on switchgear room components

4. Adding Operator Actions that are Time Critical Actions (TCAs)

EC 63785 Replace Pressurizer Heaters Description of Activity The purpose of EC 63785 is to replace all 36 safety-related pressurizer heater elements with new Westinghouse heater elements of an improved design.

The pressurizer is a vertically standing vessel designed to operate with the top half full of steam and the bottom half full of water. The pressurize heaters

___________________________________enter the pressurizer vessel from the bottom. The configuration of the bottom

LIC-15-01 19 Page 35 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 150.59 Evaluation Summary Number_

_ I_

head and support plate result in two configurations of pressurizer heaters.

The new 'inner' pressurizer heaters will be component number WATLOW STE 69LX-4590; the new 'outer' pressurizer heaters will be component number WATLOW STE 69LX-4591.

The replacement Westinghouse pressurizer heaters elements are longer, eliminating the internal heater elements in the zone adjacent to the support plate. This minimizes operating differential temperature stresses in the heater sheath. In addition, the heater sheaths will be annealed following installation to reduce residual stresses and shot peened to induce compressive stresses on its outer surface region/layer. These improvements will minimize the potential for Inter Granular Stress Corrosion Cracking (IGSCC) of the pressurizer heater sheath.

The other design change is that the pressurizer heater will be connected to the pressurizer heater nozzles with a filet weld rather than a full penetration weld. The weld stresses will remain within ASME Code allowable limits (FC07276). The replacement pressurizer heater electrical connection and power requirements are the same as the existing pressurizer heaters.

The pressurizer heaters are designed for submerged operation. In the existing design, during a system upset condition where the pressurizer water level decreases below 32 percent, the pressurizer heaters are deenergized.

The 'cut out' set point [of] the [pressurizer heaters] will be raised to 35% (Ref.

FC071 80 Ri) to accommodate the longer heaters. The new set point is still within the limits for the ECS transients and Safety Analysis (calculation FC0840 1).

The Ioncier heaters will reduce the RCS volume by a maximum of 0.237 ft3.

LIC-15-01 19 Attachment I Page 36 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title

{ 50.59 Evaluation Summary Number I

I__________________________________________________________________

Change Activity Title 50.59 Evaluation Summary, Number The total Reactor Coolant System (RCS) volume (6616 ft3) will be reduced by 0.004%. In addition, the weight of the heaters will be increased by 85 lbs.

This results in an overall pressurizer weight increase.

Reason for Activity Fort Calhoun Station (ECS) replaced its pressurizer in 2006. In May of 2010, pressurizer heater # 26 failed followed by pressurizer heater # 16. The heaters were replaced with identical heaters and the Unit restarted. The root cause analysis (2012-04327) identified that the heaters failed due to IGSCC in the outside diameter of the heater sheath tube in the vicinity of the support plate. Heater No. 26 had visible signs of IGSCC which lead to a failure of the ASME Class pressure boundary. Inspections of heater No. 16 identified no signs of IGSCC or pressure boundary cracking supporting the cause of failure being electrical. Similar pressurizer heater sheath failures have also been reported at Palo Verde, St Lucie, Sizewell, and other plants. Replacing the pressurizer heaters will reduce the vulnerability of IGSCC related failures.

Effect of Activity Replacing the heaters will improve the reliability of the pressurizer and the ASME Class I pressure boundary. Replacing the pressurizer heaters will require a new pressurizer low water level 'cut out' set point. The 'cut out' set point of the [pressurizer heaters] will be raised from 32% to 35% (FC07180 RI) to accommodate the longer heaters. The new set point is still within the limits for the ECS transients and Safety Analysis (calculation FC08401). The longer heaters will take up slightly more space in the pressurizer reducing the pressurizer volume by 0.237 ft3. The combined longier heaters weiqht is

LIC-i5-01 19 Page 37 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

.Change

• ActivityTitle

-50.59 Evaluation Summaryi N um ber.

approximately 85 lbs., increasing the overall pressurizer weight.

Summary of Conclusion for the Activity's 50.59 Review:

The conclusions of the 50.59 screening is that all aspects of the design change do not involve a change to an System, Structure or Component (SSC) that adversely affects an UFSAR described design function with the exception that the heaters are larger and take up volume displacing reactor coolant volume inventory. The reduction in ROS volume was determined to be an adverse change resulting in the need for a 50.59 evaluation.

The conclusion of the 50.59 evaluation is that the small reduction in RCS volume (0.237 ft3) has a negligible effect on the frequency, likelihood, or consequences of Design Basis Accidents (DBAs). The conclusion of the 50.59 Screen and Evaluation is that the change can be made without prior NRC approval.

EC 64310 Add a Varistor Across FCV-269Y-Description of Activity 20 This EC will install a varistor across the terminals of FCV-269Y-20, Blending Tee CH-13 Boric Acid Inlet Valve Solenoid Valve.

Reason for Activity Internal QE has found that this type of valve has the potential to induce a voltage spike when its solenoid is de-energized causing spurious indications on nearby equipment. A varistor is being installed to suppress this spike; as recommended by Op-Eval 14-013; which documents EMI impacts on A/JI-

LIC-15-0119 Page 38 Changes, Tests, and Experiments Performed Pursuant to 10 CER 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number 007Y (Al-N I), Power Margin and Setpoint Dual Meter (Channel A) as a result Change Activity Title 50.59 Evaluation Summary Number of the solenoid operating.

Effect of Activity FCV-269Y is the boric acid flow control valve used for RCS makeup in the Chemical Volume Control System (CVCS). While the addition of the varistor to the control loop will improve performance of the external loops affected by the EMI spike, it also reduces the reliability of the 269Y control loop by creating a new failure mechanism (i.e. failed varistor). Failure of FCV-269Y could result in the inability to control reactivity; which could result in a plant shut down. Because A/JI-007Y also interfaces with the Reactor Protective System (RPS), the change also indirectly impacts the design functions of the RPS.

Summary of Conclusion for the Activity's 50.59 Review The Applicability Review has determined that no other regulatory processes are applicable. The screening has determined that there is an adverse impact to the CVCS design functions due to additional components being added to the control loop. The evaluation had determined that prior NRC approval is not required because the addition of the varistor does not fulfill any of the 10 CER 50.59 criteria because the reduction in reliability is minimal and no new failure modes are created (i.e., no new accidents or malfunction results).

J ______________________________________

.1

LIC-15-01 19 Page 39 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 20)15 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title 50.59 Evaluation Summary Number EC 58237 Containment Internal Structure Description of Activity:

RVH Stand Area EC 58237 scope is to replace the existing Reactor Vessel Head (RVH) Stand with a new stand that changes the load distribution to existing Containment Internal Structure (CIS). The existing stand is 4 concrete piers that sit on the 1045'-0" slab. The new stand will be a metal frame above the slab and transfer the loads to vertical walls and columns. This is a Class I Structure and all the load combinations as defined in USAR 5.11 are applicable.

The new frame will be coated with an approved coating for the Containment Building.

The additional steel reduces the volume and adds surface area to the heat sink analysis.

There is a modification to the curb around the reactor cavity and hand rails to allow base plates to be installed for the new RVH stand.

The method for seismically restraining the RVH on the stand is changing from controlling tipping of the RVH to allowing the RVH to slide.

A steel extension is added to the top of the shield ring to address ALARA where the location of the head is approximately 3 feet higher in the new RVH stand.

The existing seismic brace that restricted movement in the south west direction (which put the shutdown cooling at risk) is no longer required. Most of the steel structural shapes are removed.

The new reactor vessel head is anchored to the Containment Internal Structure with through bolt anchors.

The use of slip critical connections in AISC 13th edition is used to

LIC-15-01 19 Page 40 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title.

015 Evlato Summary...

Num ber evaluate composite beams.

Reason for Activity During reactor core refueling the Reactor Vessel Head is removed using the overhead crane and rested on a designated Reactor Vessel (RV) Head Stand that is mounted to the top of the Operating Floor at Elevation 1 045'-0" inside Containment. The RVH load on the stand is supported by the Containment Internal Structure at Fort Calhoun Station.

An over stress condition of the existing Containment Internal Structure, including this RV head stand area, was documented by OPPD in CRs 2014-04219 and 2012-04392. OPPD committed to resolving the reactor head stand load and margin issue in their 12/2/13 letter to the NRC titled "Integrated Report to Support Restart of Fort Calhoun Station and Post Restart Commitments for Sustained Improvement." In addition, the NRC issued a letter on May 14, 2014, Fort Calhoun - NRC Integrated Inspection Report

1. NRC-14-0053 that also identified this condition and requires completion prior to the next use of the head stand.

The head stand is being replaced to restore the Containment Internal Structure to design basis requirements of UFSAR Section 5.11.

Effect of Activity The replacement of the existing head stand with a new head stand frame restores the head stand and containment internal structure to design basis for a Class I structure per the criteria defined in USAR Section 5.11. The replacement head stand is qualified for OBE (Operating Basis Earthquake)

LIC-15-01 19 Page 41 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change Activity Title T50.59 Evaluation Summary Number and SSE (Safe Shutdown Earthquake) loading with the reactor vessel head resting on the head stand. The RVH has no safety functions when placed on the stand.

The new structure adds qualified coatings into containment and does not increase the debris generated inside containment following a LOCA.

The new head stand structure increases the volume of components inside containment by approximately 100 ft3. This decreases the containment free volume by approximately 0.01% which would increase accident containment pressure by less than 0.01 psid. Since the peak accident pressure is more than 2 psi less than the allowable design pressure of 60 psig the change has no impact and is acceptable.

The modification to the curbs and handrails meet design basis. The function of the concrete curb is replaced with metal kick plates.

The RVH stand meets seismic design basis and methodology as defined in USAR Appendix F.

Modifying the shield ring (steel extension) is a passive change to minimize shine from the bottom of the reactor vessel head to the workers that maintain the RVH when it is located in the stand. The new stand is approximately 3 feet higher than the existing stand. The shield ring does not have any safety related functions.

The existing seismic brace near the entrance to the RVH laydown area and the new RVH stand was installed to protect safety related equipment during Mode 5 (shut down cooling). The new RVH stand is designed so that

LIC-15-01 19 Page 42 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CFR 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change TActivity Title,

{..

50.59 Evalu~ation Summary

,Number the reactor vessel head will see minimal controlled sliding during a seismic event and will not tip. Therefore the seismic brace is no longer required.

The design of the RVH anchorage is through bolts. The original code of record is silent on this type of anchors. All post installed anchor designs do not address through bolt design. The design basis for the through bolt is a methodology that is different than the design of record.

Summary of Conclusion for the Activity's 50.59 Review The conclusions of the 50.59 screening is that all aspects of the design change screen out with two exceptions. These are that the design and analysis methods utilized for 1). anchoring the new Reactor Vessel Head (RVH) stand to the Class I containment structure and 2). for the connection between the pedestal the RVH sits on and the girder, are a change in USAR described evaluation methodology warranting a Safety Evaluation. The changes do not involve a change to an System, Structure or Component (SSC) that adversely affects an UFSAR described design function. These activities do not adversely alter how the station is controlled and does not involve a test or experiment. The methodology for seismic design, from tipping to sliding is within the current ECS design basis and does not add a new method as described in FSAR/USAR Appendix F. Sliding was approved in the FSAR (original construction) for the polar crane where friction was credited to limit movement during a seismic event.

The methodology for anchor design is not within the design basis and will be evaluated.

LIC-15-0119 Page 43 Changes, Tests, and Experiments Performed Pursuant to 10 CFR 50.59 2015 Evaluations Note - The 10 CER 50.59 evaluations summarized below are for the most part, unedited summaries as approved by the PORC. As a result, the language may be in future tense.

Change fActivity Title

[50.59 Evaluation Summary Number

__Th deincieifoth ne recoveslha stn mette requirements for a Class I Structure defined in the USAR Section 5.11.

There is no change to the Technical Specifications.

The conclusion of the 50.59 Screen is that the methodology evaluation is required. All other activities are covered by the 50.59 screening process.

The conclusion of the 50.59 Evaluation is that the methodology is not adverse to the original code of record and does not require NRC approval.

LIC-15-O1 19 Page 1 Quality Assurance Program Changes Fort Calhoun Station, Unit No. 1 Renewed Facility Operating License No. DPR-40 NRC Docket No. 50-285 And Fort Calhoun Station Independent Spent Fuel Storage Installation NRC Docket No.72-054

LIC-15-0119 Page 2 QA Program Description Change Number

[Date of Change]

Revision 1 /

This Quality Assurance Topical Report (QATR) revision includes:

9/30/2014 changing the title Corporate Health Physicist to Manager-Radiation Protection; deleting the Division Manager-Human Resources function for the Fitness for Duty Program as this function has transitioned under the Manager-Site Security; modifying the title of the Manager-Emergency Planning and Administration to Manager-Emergency Planning; substantial changes made to the Nuclear Safety Review Board (NSRB) description; revised form numbers for 10 CFR 50.59 and 10 CER 72.48 evaluations; and removal of reference to the Corrective Action Review Board (CARB).

Revision 2 I This QATR revision includes:

12/2/2014 Revised Plant Review Committee (PRC) to Plant Operations Committee (PORC) and associated responsibilities to reflect the Exelon Quality Assurance Topical Report (QATR) description; Eliminated reference to fire protection systems and radioactive material packaging for transportation from the definition of Safety-Related.

Revision 3 /

The QATR has been revised to include applicable Exelon QATR changes made in revisions 88 and 89, as 03/26/2015 follows:

Clarify reporting structure for Document Control and Records Management. Also, clarified requirements for biennial procedure reviews.

Clarified the practice for post installation testing.

Clarified requirements for dispositioning non-conforming items in alignment with NQA-I.

Clarify NSRB audit requirements.

In addition, changes were made to reflect the Station's adoption of NFPA 805 fire protection standards per issuance of Technical SpeCification Amendment 275, to eliminate reference to ANSI standards that are addressed by NQA-1-1 994, and to clarify Manager-Systems Engineering and Manager-Engineering Programs responsibilities.

Revision 4/

This QATR revision:

11/19/2015 incorporates changes due to a shift of responsibilities for the Fort Calhoun Nuclear Oversight Department from the Vice-President Energy Delivery and Chief Compliance Officer to the Vice-President Energy Production & Marketing.

includes commitments that were inadvertently revised or were removed in the initial issue of the QATR.

LIC-15-O1 19 Page 1 10 CFR 71.106 Quality Assurance Program Approval for Radioactive Material Package Changes

LIC-15-01 19 Page 2 QA Program Description Change Number

[Date of change]

N/A No changes have been made to the quality assurance program approval for radioactive material package changes since August 13, 2015 when NRC Form 311, Quality Assurance Program Approval for Radioactive Material Packages was approved

_____________(QA Program Approval No. 71-0256, Rev. No. 8). See NRC letter dated August 13, 2015 (NRC-15-075) (ML15231A598).

LIC-1 5-0119 Page 1 Regulatory Commitments Revised in Accordance with NEI 99-04

LIC-1 5-0119 Page 2 Regulatory Commitments Revised in Accordance with NEI 99-04 Commitment

• Description Number AR 10237 This commitment tracked OPPD's commitment to implement a test program for verifying the heat transfer capability of individual heat exchangers in the CCW/RW systems and specified that CCW heat exchangers AC-IA, AC-1B, AC-IC, and AC-i1D would be tested annually. The commitment to test the heat exchangers was revised to complete inspection and cleaning on an 18-month frequency (cleaning is a license renewal commitment).

Since 1991, heat exchanger cleaning and testing has demonstrated that an 18-month cleaning interval is sufficient to ensure that the heat exchangers are capable of handling the heat load from a DBA. Testing after such cleaning has been determined to be unnecessary.

AR 9366 GL 88-17 required the implementation of procedures and administrative controls that generally avoid operations that deliberately or knowingly lead to perturbations to the RCS and/or to systems that are necessary to maintain the RCS in a stable and controlled condition while the RCS is in a reduced inventory condition.

In its response (LIC-88-1 106) dated January 4, 1989, OPPD committed to revise procedures such that if RCS water level was above the top of the hot leg nozzle prior to removing the RVH, the water level would be decreased until the steam generator U-tubes dump. This was intended to ensure that no unexpected rise in RCS water level would take place should vacuum inadvertently be lost in the steam generator U-tubes, In April 2015, the commitment was revised to clarify that the requirement to lower RCS water level until the steam generator U-tubes dump is necessary when draining the RCS to a reduced inventory, which allows for RVH removal prior to dumping the SG tubes. The RVH flange is removed at a level of 1012.5 feet, which is above reduced inventory conditions at 1010 feet (i.e., 2.5-foot margin). In a lowered inventory condition, operating experience has shown that the SG tubes dump when RCS level is below the top inner diameter of the hot leg (i.e., 1007.708 feet). Had the original commitment been maintained, it would cause entry into RCS reduced inventory conditions (i.e., < 1010 feet) when not required to remove the RVH.

AR 13723 In Amendment No. 155, the NRC approved OPPD's request to increase the spent fuel pool storage capacity to 1083 fuel assemblies. To ensure that the radiation dose to the divers was maintained ALARA, OPPD committed to follow draft RG DG-8006 ("Control of Access to High and Very High Radiation Areas in Nuclear Power Plants") if divers were used in the process of increasing the SFP storage capacity.

This one-time commitment applicable during re-racking of the SFP in the 1990's was treated as a programmatic commitment with no expiration date. The institutionalization of ALARA diving practices is demonstrated by OPPD's adoption of Exelon procedure RP-AA-461, "Radiological Controls for Contaminated Water Diving Operations."

Furthermore, no increase in the storage capacity of the SFP is planned. Thus, the commitment is no longer necessary and has been deleted.

LIC-15-01 19 Page 3 Regulatory Commitments Revised in Accordance with NEI 99-04 Commitment Description Number AR 14443 In response to a 1993 Notice of Violation, OPPO committed to develop a tag-out preparation guideline for reference during tag-out preparation to ensure that various plant configurations are considered prior to generating the tag-out. As a result of integration into the Exelon fleet, OPPD adopted Exelon procedure OP-FC-109-101, "Clearance and Tagging" that reflects current industry standard practices to ensure that plant configurations are considered when generating clearances and tag-outs. The previous OPPD procedures that contained the commitment have been superseded and

______________this commitment has been deleted as it is considered institutionalized by adoption of the Exelon procedure.

LIC-1 5-0119 Page 1 Information Removed from the USAR

LIC-15-O0119 Page 2 Information Removed from the USAR EC Number Description' EC 65695 During the Cycle 28 Core Reload Design, Note 5 was removed from Table 14.1-2, "Typical Operating Parameter Values Used in the Analysis of the Fort Calhoun Station." Note 5 previously stated "All events evaluated up to 545°F in Reference 14.1-4." Note 5 was incorporated in Table 14.1-2 in anticipation of increasing plant efficiency by raising core inlet temperature to 545°F. Although analyses were performed up to 545°F for Cycles 26 and 27, not all events were evaluated up to 545°F for Cycle 28. The note was deleted as there is currently no intention to change the core inlet

______________temperature to 5450F for Cycle 28 or any future cycle.

LIC-1 5-0119 Page 1 Summary of Technical Specification Basis Changes (TSBC)

LIC-1 5-0119 Page 2 TSBC No.

Description TS Page(s)

[Date]

14-002-0 TSBC 14-002-0 describes a normal minimum submergence level for the Raw Water pumps during a design 11-06-14basis low river event. TSBC 14-002-0 provides the basis for minimum submergence requirements of the safety related raw water pumps as low as 976'8".

14-003-0 Amendment No. 277 issued on November 6, 2014 changed footnote designations in Table 3-5 from asterisks to 12-04-14numbered footnotes. TSBC 14-003-0 made corresponding changes to footnote designations in the Basis of TS 3.2.

LIC-1 5-0119 Page 1 Technical Specification Basis Change (TSBC) Pages

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.16 River Level (Continued)

Basis.(Continued)

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 normal RW pump minimum submergence level (MSL)(2). The RW pumps can perform their design function during a design low river event of 976'9", where up to 1" of head loss can occur across the traveling screens, resulting in a cell level of 976 feet 8 inches(1,2).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 cell 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 monitoring river level to assure no sudden loss of water supply occurs, the level of the intake cells is monitored.

Intake cell levels are also adversely affected by the flows associated with the non-safety related circulating water (CW) pumps since the large flow rates associated with the CW pumps create significant head losses even with relatively clean intake cell conditions.

However, the CW pumps have a much higher MSL requirement (983 feet 0 inches) and would become unstable and trip or be manually shutdown well before intake cell levels decrease to the RW pump MSL. The head loss associated with CW pump flow would then be recovered and intake cell levels would rise.

References (1)

USAR, Section 2.7.1.2

• (2)

USAR, Section 9.8 2.16 - Page 2 Amendment No. 274

,TSBC 07-002-0 TSBC-1 4-001-0 TSBC-1 4-002-0

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.2 Equipment and Sampling Tests (continued)

The spent fuel storage-decontamination areas air treatment system is designed to filter the building atmosphere to the auxiliary building vent during refueling operations. The charcoal adsorbers are installed to reduce the potential release of radioiodine to the environment. In-place testing is performed to confirm the integrity of the filter system. The charcoal adsorbers are periodically sampled to insure capability for the removal of radioactive iodine.

The Safety Injection (SI) pump room air treatment system consists of charcoal adsorbers which are installed in normally bypassed ducts. This system is designed to reduce the potential release of radioiodine in SI pump rooms during the recirculation period following a DBA. The in-place and laboratory testing of charcoal adsorbers will assure system integrity and performance.

Pressure drops across the combined HEPA filters and charcoal adsorbers, of less than 9 inches of water for the control room filters (VA-64A & VA-64B) and of less than 6 inches of water for each of the other air treatment systems will indicate that the filters and adsorbers are not clogged by amounts of foreign matter that would interfere with performance to established levels.

If significant painting, fire or chemical release occurs such that the HEPA filters or charcoal adsorbers could become contaminated from the fumes, chemicals or foreign materials, testing will be performed to confirm system performance.

Demonstration of the automatic and/or manual initiation capability will assure the system's availability.

Verifying Reactor Coolant System (RCS) leakage to be within the LCO limits ensures the integrity of the Reactor Coolant Pressure Boundary (RCPB) is maintained. Pressure boundary leakage would at first appear as unidentified leakage and can only be positively identified by inspection. Unidentified leakage is determined by performance of an RCS water inventory balance. Identified leakage is then determined by isolation and/or inspection. Since Primary to Secondary Leakage of 150 gallons per day cannot be measured accurately by an RCS water inventory balance, footnote 3 for line item 8a on Table 3-5 states that the Reactor Coolant System Leakage surveillance is not applicable to Primary to Secondary Leakage. Primary to secondary leakage is measured by performance of effluent monitoring within the secondary steam and feedwater systems.

3.2 - Page 2 Amendment No. 15,67,128,138,169,216., 257 TSBC-0-7-003-0!, 14-003-0

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.0 SURVEILLANCE REQUIREMENTS 3.2 Equipment and Samplinci Tests (continued)

Table 3-5, Item 8b verifies that primary to secondary LEAKAGE is less or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this surveillance requirement is not met, compliance with LCO 3.17, "Steam Generator Tube Integrity," should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference 5.

The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.

The Surveillance is modified by a footnote which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

The Surveillance Frequency of daily is a reasonable interval to trend primary to secondary LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref. 5).

Table 3-5, Item 25 verifies adequate measurements are taken to ensure that facility protective actions will be taken (and power operation will be terminated) in the event of high and/or low river level conditions. The high river level limit of less than 1004 feet mean sea level is based on the maximum elevation at which facility flood control measures provide protection to safety related equipment (i.e., due to restricted access/egress to the intake structure veranda once the flood barriers are installed prior to river level reaching 1004 feet msl). A continuous watch will be established at 1002 feet mean sea level to provide adequate response time for rising river levels in accordance with the abnormal operating procedure. The river level surveillance requirement specified also ensures sufficient net positive suction head is available for operating the RW pumps. The minimum river level of 976 feet 9 inches provides adequate suction to the RW pumps for cooling plant components. The surveillance frequency of "Daily" is a reasonable interval and models guidance provided in NUREG-0212, Revision 2, "Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors," Section 4.7.6. This surveillance requirement verifies that the Missouri River water level is maintained at a level greater than or equal to 976 feet 9 inches mean sea level. A continuous watch is established to monitor the river level when the river level reaches 980 feet mean sea level to assure no sudden loss of water supply occurs.

Table 3-5, Item 26 verifies the proper position of stops on high pressure safety injection system valves.

The valves have stops to position them properly so that flow is restricted to a ruptured cold leg, ensuring that the other cold legs receive at least the required minimum flow. The refueling frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for unplanned transients if the Surveillances were performed with the reactor at power.

References

1)

USAR, Section 9.10

2)

ASTM 04057, ASTM D975, ASTM D4176, ASTM 02622, ASTM D287, ASTM 6217, ASTM D2709

3)

ASTM 0975, Table 1

4)

Regulatory Guide 1.137

5)

EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines."

3.2 - Page 5 Amendment No. 229,.24g,2.5-7, 274, 280 TSBC-09-003-0,1 !!.001 -0,14-003-0

LIC-1 5-0119 Page 1 List of Files on CD-ROM

1. File Name Size Sensitivity Location Folder Level 001 Index.pdf 35 KB Publicly Available CD-ROM 1-USAR 002 USAR 01-01.pdf 177 KB Publicly Available CD-ROM 1-USAR 003 USAR 01-02.pdf 229 KB Publiciy Availabie CD-ROM 1-USAR 004 USAR 01-03. df 192 KB Publicly Available CD-ROM 1-USAR 005 USAR 01-04.pdf 263 KB Publicly Available CD-ROM 1-USAR 006 USAR 01-05.pdf 211 KB Publicly Available CD-ROM 1-USAR 007 USAR 01-06.pdf 127 KB Publicly Available CD-ROM 1-USAR 008 USAR 01-07.pdf 109 KB Publicly Available CD-ROM 1-USAR 009 USAR 01-08.pdf 106 KB Publicly Available CD-ROM 1-USAR 010 USAR 01-09.pdf 158 KB Publicly Available CD-ROM 1-USAR 011 USAR 01-10.pdf 93 KB Publicly Available CD-ROM 1-USAR 012 USARO01-11.pdf 178 KB Publicly Available CD-ROM 1-USAR 013 USAR 01-12.pdf 172 KB Publicly Available CD-ROM 1-USAR 014 USAR 02-01.pdf 115 KB Publicly Available CD-ROM 1-USAR 015 USAR 02-02.pdf 132 KB Publicly Available CD-ROM 1-USAR 016 USAR 02-03.pdf 97 KB Publicly Available CD-ROM 1-USAR 017 USAR 02-04.pdf 124 KB Publicly Available CD-ROM 1-USAR 018 USAR 02-05.pdf 3,538 KB Publicly Available CD-ROM 1 -USAR 019 USAR 02-06.pdf 185 KB Publicly Available CD-ROM 1-USAR 020 USAR 02-07.pdf 194 KB Publicly Available CD-ROM 1-USAR 021 USAR 02-08.pdf 209 KB Publicly Available CD-ROM 1-USAR 022 USAR 02-09.pdf 201 KB Publicly Available CD-ROM 1-USAR 023 USAR 02-10.pdf 148 KB Publicly Available CD-ROM 1-USAR 024 USAR 02-11.pdf 129 KB Publicly Available CD-ROM 1-USAR 025 USAR 03-01.pdf 185 KB Publicly Available CD-ROM 1-USAR 026 USAR 03-02.pdf 222 KB Publicly Available CD-ROM 1-USAR 027 USAR 03-03.pdf 98 KB Publicly Available CD-ROM 1-USAR 028 USAR 03-04.pdf 403 KB Publicly Available CD-ROM 1-USAR 029 USAR 03-05.pdf 228 KB Publicly Available CD-ROM 1-USAR 030 USAR 03-06.pdf 212 KB Publicly Available CD-ROM 1-USAR 031 USAR 03-07.pdf 214 KB Publicly Available CD-ROM 1-USAR 032 USAR 03-08.pdf 238 KB Publicly Available CD-ROM 1-USAR 033 USAR 03-09.pdf 215 KB Publicly Available CD-ROM 1-USAR 034 USAR 03-10.pdf 103 KB Publicly Available CD-ROM 1-USAR 035 USAR 04-01.pdf 174 KB Publicly Available CD-ROM 1-USAR 036 USAR 04-02.pdf 248 KB Publicly Available CD-ROM 1-USAR 037 USAR 04-03.pdf 330 KB Publicly Available CD-ROM 1-USAR 038 USAR 04-04.pdf 192 KB Publicly Available CD-ROM 1-USAR 039 USAR 04-05.pdf 509 KB Publicly Available CD-ROM 1-USAR 040 USAR 04-06.pdf 130 KB Publicly Available CD-ROM 1-USAR 041 USAR 04-07.pdf 183 KB Publicly Available CD-ROM 1 -USAR 042 USAR 05-01.pdf 182 KB Publicly Available CD-ROM 1 -USAR 043 USAR 05-02.pdf 212 KB Publicly Available CD-ROM 1-USAR 044 USAR 05-03.pdf 214 KB Publicly Available CD-ROM 1-USAR 045 USAR 05-04.pdf 190 KB Publicly Available CD-ROM 1-USAR 046 USAR 05-05.pdf 233 KB Publicly Available CD-ROM 1-USAR 047 USAR 05-06.pdf 245 KB Publicly Available CD-ROM 1-USAR 048 USAR 05-07.pdf 222 KB Publicly Available CD-ROM 1-USAR 049 USAR 05-08.pdf 212 KB Publicly Available CD-ROM 1-USAR 050 USAR 05-09.pdf 271 KB Publicly Available CD-ROM 1-USAR

1. File Name Size Sensitivity Location Folder Level 051 USAR 05-10.pdf 227 KB Publicly Available CD-ROM 1-USAR 052 USAR 05-11.pdf 183 KB Publicly Available CD-ROM 1-USAR 053 USAR 05-12.pdf 121 KB Publicly Available CD-ROM 1-USAR 054 USAR 05-13.pdf 115 KB Publicly Available CD-ROM 1-USAR 055 USAR 06-01.pdf 200 KB Publicly Available CD-ROM 1-USAR 056 USAR 06-02.pdf 310 KB Publicly Available CD-ROM 1-USAR 057 USAR 06-03.pdf 202 KB Publicly Available CD-ROM 1-USAR 058 USAR 06-04.pdf 283 KB Publicly Available CD-ROM 1-USAR 059 USAR 06-05.pdf 176 KB Publicly Available CD-ROM 1-USAR 060 USAR 06-06.pdf 178 KB Publicly Available CD-ROM 1-USAR 061 USAR 07-01.pdf 178 KB Publicly Available CD-ROM 1-USAR 062 USAR 07-02.pdf 1,152 KB Publicly Available CD-ROM 1-USAR 063 USAR 07-03.pdf 307 KB Publicly Available CD-ROM 1-USAR 064 USAR 07-04.pdf 212 KB Publicly Available CD-ROM 1-USAR 065 USAR 07-05.pdf 298 KB Publicly Available CD-ROM 1-USAR 066 USAR 07-06.pdf 216 KB Publicly Available CD-ROM 1-USAR 067 USAR 07-07.pdf 175 KB Publicly Available CD-ROM 1-USAR 068 USAR 08-01.pdf 190 KB Publicly Available CD-ROM 1-USAR 069 USAR 08-02.pdf 207 KB Publicly Available CD-ROM 1-USAR 070 USAR 08-03.pdf 216 KB Publicly Available CD-ROM 1-USAR 071 USAR 08-04.pdf 213 KB Publicly Available CD-ROM 1-USAR 072 USAR 08-05.pdf 211 KB Publicly Available CD-ROM 1-USAR 073 USAR 08-06.pdf 100 KB Publicly Available CD-ROM 1-USAR 074 USAR 08-07.pdf 168 KB Publicly Available CD-ROM 1-USAR 075 USAR 09-01.pdf 184 KB Publicly Available CD-ROM 1-USAR 076 USAR 09-02.pdf 321 KB Publicly Available CD-ROM 1-USAR 077 USAR 09-03.pdf 216 KB Publicly Available CD-ROM 1-USAR 078 USAR 09-04.pdf 227 KB Publicly Available CD-ROM 1-USAR 079 USAR 09-05.pdf 204 KB Publicly Available CD-ROM 1-USAR 080 USAR 09-06.pdf 196 KB Publicly Available CD-ROM 1-USAR 081 USAR 09-07.pdf 162 KB Publicly Available CD-ROM 1-USAR 082 USAR 09-08.pdf 168 KB Publicly Available CD-ROM 1-USAR 083 USAR 09-09.pdf 181 KB Publicly Available CD-ROM 1-USAR 084 USAR 09-10.pdf 318 KB Publicly Available CD-ROM 1-USAR 085 USAR 09-11.pdf 474 KB Publicly Available CD-ROM 1-USAR 086 USAR 09-12.pdf 160 KB Publicly Available CD-ROM 1-USAR 087 USAR 09-13.pdf 202 KB Publicly Available CD-ROM 1-USAR 088 USAR 09-14.pdf 311 KB Publicly Available CD-ROM 1-USAR 089 USAR 10-01.pdf 177 KB Publicly Available CD-ROM 1-USAR 090 USAR 10-02.pdf 202 KB Publicly Available CD-ROM 1-USAR 091 USAR 10-03.pdf 179 KB Publicly Available CD-ROM 1-USAR 092 USAR 10-04.pdf 95 KB Publicly Available CD-ROM 1-USAR 093 USAR 10-05.pdf 91 KB Publicly Available CD-ROM 1-USAR 094 USAR 10-06.pdf 92 KB Publicly Available CD-ROM 1-USAR 095 USAR 11-01.pdf 2,729 KB Publicly Available CD-ROM 1-USAR 096 USAR 11-02.pdf 344 KB Publicly Available CD-ROM 1-USAR 097 USAR 11-03.pdf 234 KB Publicly Available CD-ROM 1-USAR 098 USAR 11-04.pdf 179 KB Publicly Available CD-ROM 1-USAR 099 USAR 11-05.pdf 174 KB Publicly Available CD-ROM 1-USAR 100 USAR 12-01.pdf 131 KB Publicly Available CD-ROM 1-USAR

1. File Name Size Sensitivity Location Folder Level 101 USAR 12-02.pdf 199 KB Publicly Available CD-ROM 1-USAR 102 USAR 12-03.pdf 180 KB Publicly Available CD-ROM 1-USAR 103 USAR 12-04.pdf 168 KB Publicly Available CD-ROM 1-USAR 104 USAR 12-05.pdf 179 KB Publicly Available CD-ROM 1-USAR 105 USAR 12-06.pdf 167 KB Publicly Available CD-ROM 1-USAR 106 USAR 12-07.pdf 166 KB Publicly Available CD-ROM 1-USAR 107 USAR 12-06.pdf 112 KB Publicly Available CD-ROM 1-USAR 108 USAR 13-01.pdf 177 KB Publicly Available CD-ROM 1-USAR 109 USAR 13-02.pdf 174 KB Publicly Available CD-ROM 1-USAR 110 USAR 13-03.pdf 169 KB Publicly Available CD-ROM 1-USAR 111 USAR 13-04.pdf 172 KB Publicly Available CD-ROM 1-USAR 112 USAR 13-05.pdf 169 KB Publicly Available CD-ROM 1-USAR 113 USAR 14-01.pdf 496 KB Publicly Available CD-ROM 1-USAR 114 USAR 14-02.pdf 232 KB Publicly Available CD-ROM 1-USAR 115 USAR 14-03.pdf 222 KB Publicly Available CD-ROM 1-USAR 116 USAR 14-04.pdf 210 KB Publicly Available CD-ROM 1-USAR 117 USAR 14-05.pdf 178 KB Publicly Available CD-ROM 1-USAR 118 USAR 14-06.pdf 254 KB Publicly Available CD-ROM 1-USAR 119 USAR 14*-07.pdf 181 KB Publicly Available CD-ROM 1-USAR 120 USAR 14-08.pdf 243 KB Publicly Available CD-ROM 1-USAR 121 USAR 14-09.pdf 251 KB Publicly Available CD-ROM 1-USAR 122 USAR 14-10.pdf 258 KB Publicly Available CD-ROM 1-USAR 123 USAR 14-11.pdf 208 KB Publicly Available CD-ROM 1-USAR 124 USAR 14-12.pdf 315 KB Publicly Available CD-ROM 1-USAR 125 USAR 14-13.pdf 246 KB Publicly Available CD-ROM 1-USAR 126 USAR 14-14.pdf 141 KB Publicly Available CD-ROM 1-USAR 127 USAR 14-15.pdf 382 KB Publicly Available CD-ROM 1-USAR 128 USAR 14-16.pdf 302 KB Publicly Available CD-ROM 1-USAR 129 USAR 14-17.pdf 181 KB Publicly Available CD-ROM 1-USAR 130 USAR 14-18.pdf 228 KB Publicly Available CD-ROM 1-USAR 131 USAR 14-19.pdf 211 KB Publicly Available CD-ROM 1-USAR 132 USAR 14-20.pdf 203 KB Publicly Available CD-ROM 1-USAR 133 USAR 14-21.pdf 169 KB Publicly Available CD-ROM 1-USAR 134 USAR 14-22.pdf 202 KB Publicly Available CD-ROM 1-USAR 135 USAR 14-23.pdf 168 KB Publicly Available CD-ROM 1-USAR 136 USAR 14-24.pdf 265 KB Publicly Available CD-ROM 1-USAR 137 USAR 15-01.pdf 93 KB Publicly Available CD-ROM 1-USAR 138 USAR 15-02.pdf 218 KB Publicly Available CD-ROM 1-USAR 139 USAR 15-03.pdf 227 KB Publicly Available CD-ROM 1-USAR 140 USAR 15-04.pdf 282 KB Publicly Available CD-ROM 1-USAR 141 USAR Appendix A.pdf 113 KB Publicly Available CD-ROM 1-USAR 142 USAR Appendix B.pdf 101 KB Publicly Available CD-ROM 1-USAR 143 USAR Appendix C.pdf 14,630 KB Publicly Available CD-ROM 1-USAR 144 USAR Appendix D.pdf 12,090 KB Publicly Available CD-ROM 1-USAR 145 USAR Appendix E.pdf 87 KB Publicly Available CD-ROM 1-USAR 146 USAR Appendix F.pdf 396 KB Publicly Available CD-ROM 1-USAR 147 USAR Appendix G.pdf 446 KB Publicly Available CD-ROM 1-USAR 148 USAR Appendix H.pdf 769 KB Publicly Available CD-ROM 1-USAR 149 USAR Appendix I.pdf 371 KB Publicly Available CD-ROM 1-USAR 150 USAR Appendix J.pdf 91 KB Publicly Available CD-ROM 1-USAR

1. File Name Size Sensitivity Location Folder Level 151 USAR Appendix K.pdf 92 KB Publicly Available CD-ROM 1-USAR 152 USAR Appendix L.pdf 98 KB Publicly Available CD-ROM 1-USAR 153 USAR Appendix M.pdf 331 KB Publicly Available CD-ROM 1-USAR 154 USARA pendix N.pdf 260 KB Publicly Available CD-ROM 1-USAR USAR Figure Section-1-USAR 155 02.pdf 11,495KB Publicly Available CD-ROM 16USAR Figure Section-1,9 BPbil vial DRM 1-USAR 5703.pdf 64K ulcyAalbeC-O USAR Figure Section-1-USAR 157 04.pdf 664 KB Publicly Available CD-ROM USAR Figure Section-1-USAR 158 05.pdf 7590 KB Publicly Available CD-ROM USAR Figure Section-1-USAR 160 06.pdf 430 KB Publicly Available CD-ROM USAR Figure Section-1-USAR 161

.d 12,955 KB Publicly Available CD-ROM 6307.pdf 477K ulcyAalbeC-O 162 USAR Figure Section-1,9 BPbil vial DRM 1-USAR 11.pdf46K ullyvilleC-M USAR Figure Section-1-USAR 163 12.pdf 4,74 KB Publicly Available CD-ROM USAR Figure Section-1-USAR 164 14.pdf 4,64KB Publicly Available CD-ROM USAR Figures Spection-1-USAR 165 A2.pdf 482 KB Publicly Available CD-ROM 16USAR Figures Section-1-USAR 168 4.pdf 1,641 KB Publicly Available CD-ROM USAR Figures Appendix-4KBPbilAvlaeCDRM 1-USAR USAR~pd FgrsApni-2310 KB Publicly Available CD-ROM 1UA 10USAR Figures Appendix-10 BPbil vial DRM 1-USAR USARdfiueApedx 360 KB Publicly Available CD-ROM 1UA NO-FC-10, Quality 2-QATR 171 Assurance Topical Report 482 KB Publicly Available CD-ROM (QATR), Revision 4