LIC-18-0003, License Amendment Request (LAR) 18-01; Revised Fort Calhoun Station Permanently Defueled Technical Specifications to Align to Those Requirements for Permanent Removal of Spent Fuel from Spent Fuel Pool

License Amendment Request (LAR) 18-01; Revised Fort Calhoun Station Permanently Defueled Technical Specifications to Align to Those Requirements for Permanent Removal of Spent Fuel from Spent Fuel Pool

ML18275A323
Person / Time
Site:	Fort Calhoun
Issue date:	09/28/2018
From:	Fisher M Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LAR 18-01, LIC-18-0003
Download: ML18275A323 (80)

Text

Omaha Public Power District 10 CFR 50.90 10 CFR 50.36 LIC-18-0003 September 28, 2018 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Fort Calhoun Station, Unit No. 1 Renewed Facility Operating License No. DPR-40 NRC Docket No. 50-285

Subject:

License Amendment Request (LAR) 18-01; Revised Fort Calhoun Station Permanently Defueled Technical Specifications to Align to Those Requirements for Permanent Removal of Spent Fuel from Spent Fuel Pool.

References:

1. Letter from OPPD (T. Burke) to USNRC (Document Control Desk), "Certification of Permanent Removal of Fuel from the Reactor", dated November 13, 2016 (LIC-16-0074) (ML16319A254)

2. Letter from OPPD (M. Fisher) to USNRC (Document Control Desk), "Fort Calhoun Station, Unit No. 1, Post-Shutdown Decommissioning Activities Report",

dated March 30, 2017 (LIC-17-0033) (ML17089A759)

3. Letter from OPPD (M. Fisher) to USNRC (Document Control Desk),

"Fort Calhoun Station Irradiated Fuel Management Plan", dated March 31, 2017 (LIC-17-0031) (ML17093A594)

4. NUH-003, "Updated Final Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel" (Revision 16),

dated July 27, 2017 (ML17213A407)

In accordance with the provisions of 10 CFR Part 50.90, the Omaha Public Power District (OPPD), is submitting a request for an amendment to the 10 CFR Part 50 License and the Permanently Defueled Technical Specifications (POTS) for Fort Calhoun Station (FCS),

Unit No. 1. The proposed amendment would revise the 10 CFR Part 50 License and associated POTS to reflect the requirements for Independent Spent Fuel Storage Installation (ISFSI) only, consistent with the permanent removal of all spent fuel from the Spent Fuel Pool (SFP).

By letter on November 13, 2016 (Reference 1), FCS provided certification of the permanent removal of fuel from the Reactor Vessel to the NRC in accordance with 10 CFR Part 50.82(a)(1 )(i) and (ii). On March 30, 2017 (Reference 2), FCS submitted the Post-Shutdown Decommissioning Activities Report (PSDAR) and on March 31, 2017(Reference 3), FCS submitted an updated Irradiated Fuel Management Plan (IFMP). Therefore, the 10 CFR Part 50 license for FCS no longer permits operation of the Reactor or emplacement or retention of fuel in the Reactor Vessel in accordance with 10 CFR Part 50.82(a)(2). The PSDAR for FCS documented the OPPD expectation that all spent fuel be completely transferred to the ISFSI by the end of 2022. OPPD awarded contracts in the first quarter of 2018 which expedited transferring all spent fuel to the ISFSI by the middle of 2020. In support of this condition, the FCS license and associated POTS 444 South 16th Street Mall

• Omaha, NE 68102-2247

U. S. Nuclear Regulatory Commission LIC-18-0003 Page 2 are being proposed for revision to comport to a facility configuration with all spent nuclear fuel in dry storage within the ISFSI, in accordance with 10 CFR Part 50.36(c)(6).

The NRC approved AREVA TN Americas' Amendment 14, to the standardized NUHOMS Certificate of Compliance (CoC) No. 1004 for Spent Fuel Storage Casks, on April 25, 2017 (Reference 4). This revision deleted the License Condition requiring a return to the SFP for inspection. With the approval of the CoC, there is no longer a requirement to return spent fuel to the SFP.

As discussed in this submittal, the remammg design basis accidents (DBA) and transients associated with fuel analyzed in Chapter 14 of the FCS Defueled Safety Analysis Report (DSAR) are no longer applicable for the condition where all spent nuclear fuel is transferred to dry cask storage within an ISFSI.

The enclosure contains a description of the proposed changes, the supporting technical analyses, and the significant hazards consideration determination. Attachment 1 of the enclosure provides the existing 10 CFR Part 50 License and the POTS pages (including associated basis sections) marked-up to show the proposed changes. Attachment 2 of the enclosure provides retyped (clean) pages with the changes proposed by Attachment 1 and denoted by revision bars in the margin. The proposed changes have been reviewed and approved by the FCS Plant Operations Review Committee (PORC).

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

There are no regulatory commitments contained within this letter.

OPPD requests approval of the proposed license amendment by October 18, 2019. Once approved, the amendment will be implemented within ninety (90) days following FCS's submittal of written notification to the NRC that all spent nuclear fuel assemblies have been transferred out of the SFP and placed in dry storage within the ISFSI.

If you should have any questions regarding this submittal or require additional information, please contact Mr. Bradley H. Blome - Director License and Regulatory Assurance at (402) 533-6041.

I declare under penalty of perjury that the foregoing is true and correct. Executed on September 28, 2018.

~::J!IJ.d~

Mary J. Fisher, Vice President, Energy Production & Nuclear Decommissioning MJF/dmp

Enclosure:

OPPD's Evaluation of the Proposed Change

U.S. Nuclear Regulatory Commission L1 C-18-0003 Page 3 c: K. M. Kennedy, NRC Regional Administrator, Region IV M. C. Layton, NRC Director, Division of Spent Fuel Management J. D. Parrot, NRC Senior Project Manager R. S. Browder, NRC Senior Health Physicist, Region IV Director of Consumer Health Services, Department of Regulation and Licensure, Nebraska Health and Human Services, State of Nebraska

OPPD's Evaluation of the Proposed Change License Amendment Request (LAR) 18-01; Revised Fort Calhoun Station Permanently Defueled Technical Specifications (POTS) to Align to Those Requirements for Permanent Removal of Spent Fuel from Spent Fuel Pool (SFP).

1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusion

6.0 REFERENCES

Attachments:

1. Mark-up of 10 CFR Part 50 License and Permanently Defueled Technical Specifications pages

2. "Clean" 10 CFR Part 50 License and Permanently Defueled Technical Specifications DPR-40 Enclosure 1

1.0

SUMMARY

DESCRIPTION The Omaha Public Power District (OPPD) hereby requests an amendment to Fort Calhoun Station, Unit No. 1 (FCS) Renewed Facility Operating License No. DPR-40 to implement a change to the Permanently Defueled Technical Specifications (POTS) to revise the License Conditions, definitions, and POTS sections to align with those required for an Independent Spent Fuel Storage Installation (ISFSI) only POTS. The proposed amendment would revise the Operating License and associated POTS to reflect removal of all FCS spent nuclear fuel from the Spent Fuel Pool (SFP) and its transfer to dry storage cask (DSC) within the onsite ISFSI. The proposed changes include the relocation of administrative controls from the POTS to NO-FC-10, Quality Assurance Topical Report (QATR)

(Reference 6.1 ), which is a licensee- controlled document.

DPR-40 Enclosure 1

10 CFR Part 50 License and POTS Change Summary Table:

The following table provides a summary of which sections are being deleted in their entirety and which sections are being revised in the POTS. The details of, and justification for the proposed changes follow in subsequent sections (arranged by 10 CFR Part 50 License Condition or POTS Section).

! DELETED REVISED License Conditions l 1.8 Aging Affects 3.B Technical Specifications  !

• 3. G Mitigating Actions Permanently Defueled Technical Specifications  !

I Table of Contents I Deleted Sections Definitions I All Terms

! 2.0 Limiting Conditions for Operation

' 2.8 Refueling 3.0 Surveillance Requirements 3.0.1 Surveillance Requirements 3.0.2 Surveillance Requirements 3.0.3 Surveillance Requirements 3.0.4 Surveillance Requirements 3.0.5 Surveillance Requirements 3.2 Equipment and Sampling Tests 4.0 Design Features 4.1 Site 4.3 Fuel Storage I 5.0 Administrative I 5.1 Responsibility 5.2 Organization I 5.3 Facility Staff Qualification 5.4 Training i 5.8 Procedures i 5.9 Reporting Requirements 5.10 Record Retention 5.11 Radiological 5.16 Radiological Effluents and Environmental Monitoring Programs I 5.17 ODCM I 5.20 Technical Specification Basis DPR-40 Enclosure 1

2.0 DETAILED DESCRIPTION On November 13, 2016, FCS submitted the Certification of Permanent Removal of Fuel from the Reactor Vessel (Reference 6.2) to the NRC in accordance with 10 CFR Part 50.82(a)(1 )(ii). On March 30, 2017, FCS submitted the Post-Shutdown Decommissioning Activities Report (PSDAR)

(Reference 6.3) and on March 31, 2017, FCS submitted an updated Irradiated Fuel Management Plan (IFMP) (Reference 6.4). Therefore, the 10 CFR Part 50 license for FCS no longer will permit operation of the reactor or emplacement or retention of fuel in the reactor vessel in accordance with 10 CFR Part 50.82(a)(2). The PSOAR for FCS, documented that OPPO expects to have all spent fuel transferred to the ISFSI by the end of the year 2022. OPPD awarded contracts in the first quarter of 2018 which expedited transferring all spent fuel to the ISFSI by the middle of 2020. Transfer of fuel out of the SFP supports decommissioning of FCS, which involves the eventual dismantlement of the SFP.

In support of this condition, the FCS license and associated POTS are being proposed for revision, in accordance with 10 CFR 50.36(c)(6), to comport to a facility configuration with all spent nuclear fuel in dry storage within the onsite ISFSI at FCS using casks certified for use under a general 10 CFR 72 license.

The NRC approved AREVA TN Americas' Amendment 14, to the standardized NUHOMS Certificate of Compliance (CoC) No. 1004 for Spent Fuel Storage Casks, on April 25, 2017 (Reference 6.5). This revision deleted the requirement to return spent fuel to the SFP for inspection. With the approval of the CoC, there is no longer a requirement to return spent fuel to the SFP.

The existing POTS contain Limiting Conditions for Operation (LCO) that provide for appropriate functional capability of equipment required for safe storage and management of irradiated fuel with fuel stored in a SFP. As such , the existing POTS provide a level of control in excess of that needed for safe storage and management of irradiated fuel in an ISFSI. The majority of the existing POTS sections are only applicable when irradiated fuel assemblies are within the SFP. Once all spent fuel assemblies have been transferred to the ISFSI, all remaining LCO (and associated Surveillance Requirements (SR)) will no longer be applicable and are being proposed for deletion. The POTS being proposed reflect the removal of all spent fuel from the SFP. The proposed changes will result in POTS that will be applicable to FCS after the last spent fuel assembly has been removed from the SFP and placed within the ISFSI.

A new POTS design requirement is being added that prohibits storage of spent fuel in the SFP to comport to the ISFSI-Only Emergency Plan and the ISFSI-Only Security Plan which do not address the SFP or items stored within it. This will also align the requirement to the AREVA CoC 1004, Amendment 14, to no longer require fuel to be returned to the SFP.

The proposed changes to the POTS also involve relocating administrative requirements to either the FCS QATR or other licensee - controlled documents, and subsequently controlling them in accordance with 10 CFR 50.54(a), 10 CFR 50.71(e), and 10 CFR 50.59, respectively. This relocation is being proposed pursuant to the criteria contained in 10 CFR 50.36 and in accordance with recommendations contained in NRC Administrative Letter 95-06 (Reference 6.6).

DPR-40 Enclosure 1

Pending Licensing Actions under NRC Review There are no other pending license amendment requests associated with License Conditions or Technical Specifications (TS) currently docketed for FCS. Therefore, no disposition of other changes ,

as they relate to this license amendment request, is needed.

DPR-40 Enclosure 1

License Condition Changes:

The proposed changes to the Facility Operating License are as follows:

Eliminate License Condition 1.8. related to aging affects during periods of extended operation.

Revise License Condition 3.8. related to technical specifications.

Eliminate License Condition 3.G. related to mitigation strategy.

DPR-40 Enclosure 1

Technical Specifications Changes The cover page will replace the term "operating licensing" with "renewed facility license" in the title The deleted sections will be removed from the POTS in their entirety including all pages and numbering.

Formatting and page numbering will be changed to match only the remaining sections.

Definitions Being Deleted:

All of the following definitions will be deleted:

Fuel Handling Operations Certified Fuel Handler (CFH)

Non-Certified Operator (NCO)

Actions Operable - Operability Offsite Dose Calculation Manual (ODCM)

Unrestricted Area Permanently Defueled Technical Specification Section Changes:

The proposed changes also include a renumbering and removal of pages and sections where appropriate, to condense and reduce the number of pages in the POTS without affecting the technical content. The POTS table of contents is also revised accordingly. It also includes header changes to track deleted sections.

Section 2.8, Fuel Handling, including figures will be deleted in its entirety.

The basis for this section will be deleted to reflect the section's removal.

Section 3.0.1, Surveillance Requirements, will be deleted in its entirety.

Section 3.0.2, Surveillance Requirements, will be deleted in its entirety.

Section 3.0.3, Surveillance Requirements, will be deleted in its entirety.

Section 3.0.4, Surveillance Requirements, will be deleted in its entirety.

Section 3.0.5, Surveillance Requirements, will be deleted in its entirety.

The basis for these sections will be deleted to reflect the sections' removal.

Section 3.2, Equipment and Sampling Tests, including Table 3-4 and Table 3-5, will be deleted in its entirety.

The basis for this section will be deleted in its entirety.

DPR-40 Enclosure 1

Section 4.1, Site, will be revised to remove the document description providing the location of the sites exclusion area boundary.

Section 4.3, Fuel Storage, will be revised to only describe the limitation associated with the SFP being permanently defueled. The reference will be deleted from this section to reflect the deletion of the associated requirement.

Section 5.1.1, Responsibility, will be deleted from the POTS in its entirety. The content will be relocated to the QATR.

Section 5.1.2, Responsibility, will be deleted from the POTS in its entirety.

Section 5.2.1, Organization, will be deleted from the POTS in its entirety. The content will be revised to remove the CFH requirements and will be relocated to the QATR.

Section 5.2.2, Facility Staff, including Table 5.2-1, will be deleted in its entirety.

Section 5.3.1, Facility Staff Qualification, will be deleted from the POTS in its entirety. The content will be revised to remove the CFH requirements and change specific qualification exceptions, and relocated to the QATR.

Section 5.4.1, Training, will be deleted from the POTS in its entirety. The content will be relocated to the QATR.

Section 5.4.2, Training, will be deleted from the POTS in its entirety.

Section 5.8.1 (a), Procedures, will be deleted in its entirety.

Section 5.8.1 (d), Programs, will be deleted from the POTS in its entirety. The content will be relocated to the QATR.

Section 5.8.2, Procedures, will be deleted in its entirety.

Section 5.8.3, Procedures, will be deleted in its entirety.

Section 5.9, Unique Reporting Requirements, will be deleted from the POTS in its entirety. The content will be relocated to the QATR.

Section 5.1 0, Record Retention, will be deleted in its entirety. The content will be relocated to the QATR.

Section 5.11.1, Radiation Protection Program, will be revised to place the note closer to its associated content.

Section 5.11.2, Radiation Protection Program, will be revised to remove the Shift Manager (SM) responsibility of controlling the Locked High Radiation Area keys from this section.

OPR-40 Enclosure 1

Section 5.16.1, Radiological Effluent and Environmental Monitoring Program, will be deleted from the POTS in its entirety. The content, revised to remove reference to a deleted TS section and relocated to the QATR.

Section 5.16.2, Radiological Effluent and Environmental Monitoring Program, will be deleted from the POTS in its entirety. The content will be relocated to the QATR.

Section 5.17, Offsite Dose Calculation Manual (ODCM), will be deleted from the POTS in its entirety.

The content will be relocated to the QATR.

Section 5.20, Technical Specification Basis, will be deleted in its entirety.

DPR-40 Enclosure 1

3.0 TECHNICAL EVALUATION

On November 13, 2016, FCS submitted the certification of permanent removal of fuel from the reactor vessel. Since both the certification of permanent cessation of power operations and of permanent removal of fuel from the reactor vessel for FCS have been submitted in accordance with 10 CFR Part 50.82(a)(1 )(i) and (ii), the 10 CFR Part 50 license no longer will authorize reactor operation or emplacement or retention of fuel in the reactor vessel in accordance with 10 CFR Part 50.82(a)(2).

As a result of the certifications submitted by OPPD in accordance with 10 CFR Part 50.82(a){1 ), the consequent removal of authorization to operate the reactor or to emplace or retain fuel in the reactor vessel in accordance with 10 CFR Part 50.82(a)(2), and the complete removal of fuel from the SFP none of the accident or event scenarios postulated in the Defueled Safety Analysis Report (DSAR) are credible.

OPPD is in the process of decommissioning FCS. In support of this activity, the spent fuel is being transferred from the SFP to the ISFSI. The proposed changes to the FCS POTS reflect the removal of all the spent fuel from the SFP. With no spent fuel in the SFP, the design bases for spent fuel storage in the SFP and the design basis accident for fuel handling are no longer applicable.

The FCS DSAR, Chapter 14, Safety Analysis, currently addresses the design basis accidents (DBA) and transient scenarios applicable to FCS in the permanently defueled condition with irradiated fuel stored in the SFP. These postulated accidents are predicated on spent fuel being stored in the SFP.

However, upon transfer of all irradiated fuel to storage in the ISFSI, the accident scenarios postulated in the DSAR associated with the storage of fuel are no longer possible. The ISFSI is a passive system that does not rely on electrical power for heat transfer. With removal of the spent fuel from the SFP, there are no remaining spent fuel assemblies to be monitored and there are no credible fuel related accidents that require actions of a SM , CFH, or an NCO to prevent occurrence or mitigate the consequences.

OPPD plans to continue using a decommissioning method called SAFSTOR, in which most fluid systems are drained and the plant is left in a stable condition until final dismantlement. Administrative controls that are required to be in place when decontamination or dismantling activities of radioactive systems, structures, and components are being performed are designed to minimize the likelihood of an off-normal or accident event, and thereby the consequences of such an event. The proposed changes do not have an adverse impact on the remaining decommissioning activities or any of their postulated consequences .

The spent fuel will be stored in the ISFSI until it is shipped off site in accordance with the schedules described in the PSDAR, as updated.

During decommissioning (with all spent fuel in dry storage within an ISFSI), no plant systems are relied upon for spent fuel storage. In this condition there are no credible accidents whose prevention or mitigation would need to be addressed by plant POTS. The spent fuel storage canisters used in the ISFSI are subject to their own Certification of Compliance (CoC) and associated storage canister Technical Specifications. The NRC approved AREVA TN Americas ' Amendment 14, to the standardized NUHOMS Certificate of Compliance (CoC) No. 1004 for Spent Fuel Storage Casks, on April 25, 2017 (Reference 6.5). This revision deleted the requirement to return spent fuel to the SFP for DPR-40 Enclosure 1

inspection. With the approval of the CoC, there is no longer a requirement to return spent fuel to the SFP.

A discussion of the DSAR Chapter 14, DBA is provided in Section 4.1, "Applicable Regulatory Requirements/Criteria," of this submittal. There are no accident scenarios that apply to the condition with all spent fuel stored in dry casks within an ISFSI. Therefore, no analyzed accidents associated with the storage of fuel remain applicable to FCS in the condition with all spent fuel stored in dry casks within an ISFSI.

The definition of safety-related structures, systems, and components (SSCs) in 10 CFR 50.2, "Definitions," states that safety-related SSCs are those relied on to remain functional during and following design basis events to assure:

1. The integrity of the reactor coolant boundary;

2. The capability to shut down the reactor and maintain it in a safe shutdown condition; or,

3. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in 10 CFR 50.67(b)(2) or§ 100.11.

The first two criteria (integrity of the reactor coolant pressure boundary and safe shutdown of the reactor) are not applicable to a plant in a permanently defueled condition. The third criterion is related to preventing or mitigating the consequences of accidents that could result in potential offsite exposures exceeding limits. However, after all spent fuel assemblies have been transferred to the ISFSI, there are no longer any SSCs at FCS that are required to be relied upon for accident mitigation. Therefore, with no fuel stored in the SFP, none of the SSCs at FCS meet the definition of a safety-related SSC as stated in 10 CFR 50.2.

10 CFR 50.36, "Technical Specifications," promulgates the regulatory requirements related to the content of Technical Specifications. As detailed in subsequent sections of this proposed amendment, this regulation lists four criteria to define the scope of equipment and parameters that must be included in a plant's Technical Specifications. A discussion of the applicability of these four criteria in the permanently defueled condition with all fuel removed from the SFP is provided in Section 4.1, "Applicable Regulatory Requirements/Criteria ," of this submittal. In a permanently defueled condition with all spent fuel in dry storage within an ISFSI, the scope of equipment and parameters that need be included in the FCS POTS is limited to a description of the design features and high radiation area administrative controls.

DPR-40 Enclosure 1

LICENSE CONDITION CHANGE BASIS:

License Condition 1.8.; the proposed change removes the requirements to maintain the sites aging management program. The requirements within 10 CFR 54.4 associated with the plant systems, structures, and components (SSG) are no longer applicable to a station with all fuel stored in the ISFSI.

The implementation scope for this requirement included:

"(a) Plant systems, structures, and components within the scope of this part are-

1) Safety-related systems, structures, and components which are those relied upon to remain functional during and following design-basis events (as defined in 10 CFR 50.49 (b)(1)) to ensure the following functions-(i) The integrity of the reactor coolant pressure boundary; (ii) The capability to shut down the reactor and maintain it in a safe shutdown condition ; or (iii) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to those referred to in§ 50.34(a)(1), § 50.67(b)(2), or§ 100.11 of this chapter, as applicable.

(2) All non-safety-related systems, structures, and components whose failure could prevent satisfactory accomplishment of any of the functions identified in paragraphs (a)(1 )(i), (ii), or (iii) of this section.

(3) All systems, structures, and components relied on in safety analyses or plant evaluations to perform a function that demonstrates compliance with the Commission's regulations for fire protection (10 CFR 50.48), environmental qualification (10 CFR 50.49), pressurized thermal shock

( 10 CFR 50.61 ), anticipated transients without scram ( 10 CFR 50.62), and station blackout

( 10 CFR 50.63).

(b) The intended functions that these systems, structures, and components must be shown to fulfill in

§ 54.21 are those functions that are the bases for including them within the scope of license renewal as specified in paragraphs (a)(1)- (3) of this section ."

All fire protection requirements associated with 10 CFR 50.48(f) are controlled within the fire protection program and do not require the separate aging requirements associated with 10 CFR 54.4.

Prior to cessation of operations, FCS license renewal commitments for aging management were incorporated into Chapter 15, "Aging Management Programs," of the Updated Final Safety Analysis Report (USAR), and subsequent to permanent defueling of the DSAR, this is updated in accordance with 10 CFR 50.71(e). Therefore, changes to these license renewal commitments continue to be evaluated pursuant to the criteria in 10 CFR 50.59.

There is no other equipment meeting the requirements of this standard that are needed in the ISFSI only condition. Therefore, none of the requirements associated with the scope § 54.4 remain germane and the deletion is consistent with the requirements associated with the decommissioning plant can be removed from the FCS License Conditions.

DPR-40 Enclosure 1

License Condition 3.B.; the proposed change included in this amendment revises the wording to the license. The paragraph is changed to restore the wording in the License Condition that describes the incorporation of changes to the license. It also removes the specific wording for the implementation of the PDTS. This wording was placed in the License Condition to support the implementation of the PDTS and is proposed to be revised to correctly describe how changes are incorporated. This is an administrative change that restores the portion of the License Condition wording associated with incorporation of changes to that proceeding the implementation of the PDTS.

License Condition 3.G.; the proposed language change associated with operating the facility in this License Condition is removed to reflect that the FCS license no longer allows storage of spent fuel in the SFP. The NRC issued this License Condition on August 23, 2007, to incorporate the requirements for the Interim Compensatory Measures (ICM) Order EA-02-026, Section B.S.b mitigation strategies (dated February 25, 2002). Subsequently, 10 CFR 50.54(hh)(2) became effective on May 26, 2009.

This License Condition provides mitigation strategies and response procedure requirements for loss of large areas of the plant due to explosions or fire. However, as stated in 10 CFR 50.54(hh)(3), this section does not apply to a defueled reactor that has submitted the certification for permanent removal of fuel under 10 CFR 50.82(a).

On November 28, 2011, the NRC issued a letter that rescinded Item B.S.b of the ICM Order EA-02-26. Therefore, neither the ICM Order nor 10 CFR 50.54(hh) continue to apply to FCS.

It is noted that SECY-16-0142, Draft Final Rule- Mitigation of Beyond-Design-Basis Events, dated December 15, 2016 (Reference 6. 7), would, in part, relocate requirements existing in § 50.54(hh)(2) for mitigation of the effects of a loss of large area of the plant due to explosions or fire to a new section 10 CFR 50.155. As discussed in Enclosure 1 to SECY-16-0142, the NRC concludes that it is inappropriate to apply requirements to implement mitigation measures for large fires and explosions to a permanently shutdown and defueled reactor, where the fuel was moved to an ISFSI or removed from the site. The applicability of the proposed§ 50.155 states that holders of operating licenses for which the NRC has docketed the§ 50.82(a)(1) certifications need not meet the requirements of this section once all irradiated fuel has been permanently removed from the SFP. Since this will be the condition of the FCS facility prior to implementing this amendment request, FCS will meet the requirements as described in SECY 0142 and no longer require implementation of this requirement. Therefore, the deletion of this License Condition is consistent with the requirements associated with the fuel located only in the ISFSI.

DPR-40 Enclosure 1

TECHNICAL SPECIFICATIONS CHANGE BASIS:

The deleted sections will be removed from the POTS in their entirety including all pages and numbering. The formatting and page numbering of the remaining sections will be changed to match those of only those remaining sections.

Cover Sheet, provides a title sheet for Appendix A of the license, which is the Permanently Defueled Technical Specifications. The term "operating" will be removed and verbiage will be added to the title to align this cover sheet to reflect the site license and the remainder of the license conditions. This is an administrative change that more correctly labels the Appendix, there is no technical information associated with the cover sheet. Therefore, the revision to the page is acceptable.

Definitions, provides defined terms that are applicable throughout the POTS and POTS Basis. After transfer of spent fuel from the SFP to the ISFSI is complete, there will no longer be any applicable LCO or SR in the FCS POTS. As such, the definitions described below will no longer be needed. Therefore, deleting these definitions from the POTS during implementation after the spent fuel is transferred from the SFP to the ISFSI is acceptable.

Section 2.0, Limiting Conditions for Operations, contain LCO that provide for appropriate functional capability of plant equipment required for safe maintenance and storage of fuel assemblies in the SFP.

This section is being deleted which reflects the permanent removal of spent fuel from the SFP. The sections are no longer applicable and are proposed to be deleted in their entirety to meet requirements that reflect the permanently defueled condition of the SFP. After the transfer of spent fuel from the SFP to the ISFSI, there will no longer be any applicable LCO or SR. As such, the LCO described below will no longer be needed. Therefore, deleting these LCO from the POTS during implementation after the spent fuel is transferred from the SFP to the ISFSI is acceptable.

With the POTS sections and figure deleted the applicable basis sections will also be removed .

Section 2.8, Fuel Handling, all of its subsections and bases will be deleted as discussed below:

Section 2.8.3, Fuel Handling Operations - Spent Fuel Pool, establishes the acceptable spent fuel assembly storage within the SFP, the requirements during fuel movement within the SFP for SFP water level. All sections are being deleted to reflect a permanently defueled SFP status.

Section 2.8.3(1), Spent Fuel Assembly Storage, ensures the SFP keff remains< 0.95 with unborated water. It also ensures that placement of fuel within the SFP meets specific burn up requirements or physical restrictions to prevent exceeding storage reactivity requirements. Specific SR in Section 3.2 are only applicable with irradiated fuel assemblies in the SFP. Following the transfer of all fuel assemblies from the SFP to the ISFSI, FCS will no longer store spent fuel in the SFP and this section, and associated Figure 2-10 will no longer be needed. As such, this section and figure may be deleted in its entirety with no impact on continued safe storage and maintenance of irradiated fuel in an ISFSI at FCS. With the POTS section deleted in its entirety the applicable bases and references will also be removed.

Section 2.8.3(2), Spent Fuel Pool Water Level , specifies requirements to ensure that the minimum water level in the SFP meets the iodine decontamination factor assumptions used in the fuel handling accident (FHA) analysis of record. The specified water level shields and minimizes the general area dose when the storage racks are filled. The water also provides shielding during the movement of spent fuel. The specific SR in Section 3.2 is applicable only during movement of irradiated fuel DPR-40 Enclosure 1

assemblies in the SFP. Following the transfer of all fuel assemblies from the SFP to the ISFSI, FCS will no longer store spent fuel in the SFP and this section will no longer be needed. As such, this section may be deleted in its entirety with no impact on continued safe storage and maintenance of irradiated fuel in an ISFSI at FCS. With the POTS section deleted in its entirety the applicable bases and references will also be removed.

Section 2.8.3(3), Spent Fuel Pool Boron Concentration, specifies requirements to ensure that the minimum boron concentration meets the minimum concentration in the SFP. The boron minimized the possibility of an accident that could affect public health and safety from occurring when fuel assemblies are stored in the spent fuel pool. The specific SR in Section 3.2 is applicable only during storage of irradiated fuel assemblies in the SFP. Following the transfer of all fuel assemblies from the SFP to the ISFSI, FCS will no longer store spent fuel in the SFP and this section will no longer be needed. As such, this section may be deleted in its entirety with no impact on continued safe storage and maintenance of irradiated fuel in an ISFSI at FCS. With the POTS section deleted in its entirety the applicable bases and references will also be removed.

Section 3.0, Surveillance Requirements, establishes the standards and periods used to implement SR for plant systems. These sections are being deleted as discussed below. With the POTS section deleted in its entirety the applicable basis section will also be also removed .

Section 3.0.1, Surveillance Requirements, establishes the requirement for specific surveillance intervals and maximum extension. Following the transfer of all fuel assemblies from the SFP to the ISFSI, FCS will no longer store spent fuel in the SFP and there will be no requirements remaining that would apply to this section. Therefore, this section may be deleted in its entirety with no impact on continued safe storage and maintenance of irradiated fuel in an ISFSI at FCS. With the POTS section deleted in its entirety the applicable bases and references will also be removed.

Section 3.0.2, Surveillance Requirements, establishes surveillance intervals for individual specifications. Following the transfer of all fuel assemblies from the SFP to the ISFSI, FCS will no longer store spent fuel in the SFP and there will be no requirements remaining that would apply to this section. Therefore, this section may be deleted in its entirety with no impact on continued safe storage and maintenance of irradiated fuel in an ISFSI at FCS. With the POTS section deleted in its entirety the applicable bases and references will also be removed.

Section 3.0.3, Surveillance Requirements, establish the requirement to apply all codes and standards within the POTS to Sections 3.01 and 3.02. Following the transfer of all fuel assemblies from the SFP to the ISFSI, FCS will no longer store spent fuel in the SFP and there will be no requirements remaining that would apply to this section. Therefore, this section may be deleted in its entirety with no impact on continued safe storage and maintenance of irradiated fuel in an ISFSI at FCS. With the POTS section deleted in its entirety the applicable bases and references will also be removed.

Section 3.0.4, Surveillance Requirements, establishes the requirement that surveillances must be met during specified conditions in the specification for which the requirements of the LCO apply. Following the transfer of all fuel assemblies from the SFP to the ISFSI, FCS will no longer store spent fuel in the SFP and there will be no requirements remaining that would apply to this section. Therefore, this section may be deleted in its entirety with no impact on continued safe storage and maintenance of irradiated fuel in an ISFSI at FCS. With the POTS section deleted in its entirety the applicable bases and references will also be removed .

OPR-40 Enclosure 1

Section 3.0.5, Surveillance Requirements, establishes requirements for a delay of performing a surveillance. Following the transfer of all fuel assemblies from the SFP to the ISFSI, FCS will no longer store spent fuel in the SFP and there will be no requirements remaining that would apply this section.

Therefore, this section may be deleted in its entirety with no impact on continued safe storage and maintenance of irradiated fuel in an ISFSI at FCS. With the POTS section deleted in its entirety the applicable bases and references will also be removed .

Section 3.2, Equipment and Sampling Tests, establishes surveillance requirements for plant equipment and conditions related to safety. This section contains system SR that will no longer be required with the purposed removal of the associated system operability requirements and POTS sections.

The following surveillance test requirements from POTS Section 3.2, included in Table 3-4, are proposed to be deleted:

Table 3-4 item:

1. Table 3-4 item 1, SFP Boron The following surveillance test requirements from POTS Section 3.2, included in Table 3-5, are proposed to be deleted:

Table 3-5 items:

1. Table 3-5 item 1, SFP Racks

2. Table 3-5 item 2, SFP Level

3. Table 3-5 item 3, Spent Fuel Assembly Storage Following the transfer of all fuel assemblies from the SFP to the ISFSI, FCS will no longer store spent fuel in the SFP and there will be no requirements remaining that would apply to this section. Therefore, this section may be deleted in its entirety with no impact on continued safe storage and maintenance of irradiated fuel in an ISFSI at FCS. With the POTS section deleted in its entirety the applicable bases and references will also be removed.

Section 4.0, Design Features, provides information and design requirement associated with plant systems. Sections will be deleted or revised as described in each change basis.

Section 4.1, Site, provides a description of the site and boundaries. After the transfer of spent fuel from the SFP to the ISFSI, there will no longer be any fuel assemblies in the SFP. As such, certain design features will have no relevance to, and no longer apply to, the storage of fuel assemblies in an ISFSI.

The site exclusion area description, based on requirements contained in § 100.3 regarding reactor accident dose analyses, is being proposed for deletion. With the submittal in accordance with

§ 50.82(a)(2), FCS 10 CFR Part 50 license no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel, this design feature statement is no longer needed, and its description may be deleted. The removal of the description describing a reference in the FSAR, as updated, that defines the exclusion area boundary, does not alter any regulatory requirements related to licensee authority over the site location, and therefore does not have an impact on continued safe storage and maintenance of irradiated fuel in an ISFSI. Therefore, the revision of this section after the spent fuel is transferred from the SFP to the ISFSI is acceptable.

Section 4.3, Fuel Storage, establishes requirements regarding the design, use, and maintenance of spent fuel storage racks, prevention of SFP drainage, and spent fuel capacity limitations. After the DPR-40 Enclosure 1

transfer of spent fuel from the SFP to the ISFSI, there will no longer be any fuel assemblies in the SFP.

As such, certain design features will have no relevance to, and no longer apply to, the storage of fuel assemblies in an ISFSI. Adding a new design feature stating that spent fuel shall not be stored in the SFP documents the premise on which this proposed amendment is based (i.e., spent fuel no longer being stored in the SFP). Therefore, the revision of this section after the spent fuel is transferred from the SFP to the ISFSI is acceptable.

Section 4.3, References, provides the applicable reference used to justify content located in this section. The section including the information associated with reference has been deleted. Therefore, it is no longer required and will be deleted . This is an administrative change and does not remove technical information associated with this section. Therefore, the revision to the references is acceptable.

Section 5.0, Administrative Controls, establishes the requirements associated with personnel, administrative programs, reporting, and POTS basis control. This section is proposed to be revised to include only those administrative requirements needed with all of the spent fuel in the ISFSI. All of the sections in 5.0, with the exception of Section 5.11, Radiation Protection Program, are being deleted in their entirety, as shown in Attachment 1 and 2. Pertinent information being deleted from POTS is being relocated to the QATR. POTS, Section 5.11 is being retained with the exception of removal of the SM responsibility. Precedence shows the NRC considers relocating these requirements to the quality assurance program (QAP) acceptable because of the controls imposed by 10 CFR 50, Appendix B, the existence of an NRC approved QAP, and the QAP change control process in 10 CFR 50.54(a).

Maintaining relocated requirements in accordance with the change control process in 10 CFR 50.54 provides adequate control based on the ISFSI-only status of the facility. With the transfer of the spent fuel to the ISFSI , the administrative controls pertaining to the safe storage of spent fuel within the spent fuel pool are no longer needed or applicable. After the transfer of spent fuel from the SFP to the ISFSI, there will no longer be any fuel assemblies in the SFP. As such, the associated administrative controls will have no relevance to and no longer apply to the storage of fuel assemblies in an ISFSI. Therefore, deleting the associated POTS Sections after the last fuel transfer from the SFP to the ISFSI is acceptable.

Section 5.1, Responsibility, provides a description and requirements regarding certain key management responsibilities. The section is being proposed for deletion or relocation to the QATR. NRC Administrative Letter 95-06 (Reference 6.6) provides a discussion concerning the relocation of Technical Specifications administrative controls to a QAP. Precedence shows the NRC considers relocating these requirements to the QAP acceptable because of the controls imposed by 10 CFR 50, Appendix B, the existence of an NRC approved QAP, and the QAP change control process in 10 CFR 50.54(a). Maintaining relocated requirements in accordance with the change control process in 10 CFR 50.54 provides adequate control based on the ISFSI-only status of the facility. With the transfer of the spent fuel to the ISFSI, the administrative controls pertaining to the safe storage of spent fuel within the SFP are no longer needed or applicable.

Section 5.1.1, Responsibility, establishes the requirements for the plant manager. The remaining requirements of this section related to the responsibilities of the plant manager will be deleted from the POTS and relocated to the QATR. Relocating these responsibilities to the QATR is consistent with NRC Administrative Letter 95-06, Relocation of Technical Specification Administrative Controls Related to Quality Assurance. As part of this change administrative control for signature is also being added to this section in the QATR. Therefore, the deletion of this section in its entirety is acceptable.

OPR-40 Enclosure 1

Section 5.1.2, Responsibility, establishes the requirements for the SM. The SM responsibilities are being eliminated. With removal of all of the spent fuel from the SFP, a need for the SM for spent fuel management no longer exists. The position of SM described in this section is a holdover from the function of supervising multiple functions of a nuclear power plant. With the limited requirements for supervision of the passive fuel storage at the ISFSI or with respect to the decommissioning of the former power generation facility, the SM position and shift command function are no longer required and the proposed deletion of this section in its entirety is acceptable.

Section 5.2, Organization, provides a description and requirements regarding onsite and offsite organizations and facility staffing. Descriptions include lines of authority and staff responsibilities.

Requirements include the associated POTS Table 5.2-1, which specifies a minimum shift crew composition staffing requirement for CFH and NCO. Section 5.2 also specifies requirements for fuel handling operations and supervision. Providing onsite and offsite organization descriptions in the QATR is consistent with NRC Administrative Letter 95-06. Therefore, the proposed deletion and relocation are acceptable.

Section 5.2.1, Organization, establishes the requirements for plant lines of authority. Section 5.2.1(d) provides requirements for organizational freedom of the CFH trainers, and the health physics and quality assurance personnel. FCS proposes to eliminate the portion of Section 5.2.1(d) pertaining to CFH trainers. The remainder of Section 5.2.1 will be deleted from the POTS and relocated to the QATR to provide an equivalent description of the requirements for organizational freedom of the health physics and quality assurance personnel. Relocating these responsibilities to the QATR is consistent with NRC Administrative Letter 95-06, Relocation of Technical Specification Administrative Controls Related to Quality Assurance. Therefore, the proposed deletion and relocation are acceptable.

Section 5.2.2, Facility Staff, establishes the requirements for personnel required at the station. Facility organization which assures safe facility operations and safety of the nuclear fuel. The section maintained staff to ensure the safe storage and movement of fuel, including an individual qualified in radiation protection procedures and designation of fire responsibilities. The associated POTS Table 5.2-1, which specifies a minimum shift crew composition staffing requirement for CFH and NCO is included in this section. The QATR and DSAR address the necessary organizational requirements for FCS after all spent fuel has been transferred to ISFSI. Following the transfer of all spent fuel to the ISFSI, and the new provision in Section 4.3 prohibiting storage of fuel in the SFP, there will no longer be a need for CFH or other specified personnel requirements in this section; therefore this proposed deletion of this section in its entirety is acceptable.

Table 5.2-1, Minimum Shift Crew Composition, establishes the minimum staff required to be onsite.

The requirements of TS 5.2.1 are being deleted from POTS. This table specifies the requirements of the CFH and NCO for the safe storage and movement of fuel in the SFP. Following the transfer of all spent fuel to the ISFSI, and the new provision in Section 4.3 prohibiting storage of fuel in the SFP, there will no longer be a need for CFH or NCO requirements in this section; therefore the proposed deletion of this table including notes in its entirety is acceptable.

Section 5.3, Facility Staff Qualification, establishes the requirements for prerequisite knowledge for the facility staff. Providing staff qualification descriptions in the QATR is consistent with NRC Administrative Letter 95-06, Relocation of Technical Specification Administrative Controls Related to Quality Assurance. Therefore, the proposed deletion and relocation are acceptable.

DPR-40 Enclosure 1

Section 5.3.1, Facility Staff Qualification, establishes the requirements for prerequisite knowledge for the facility staff. The minimum requirements in ANSI N18.1-1971 that contain the minimum requirements associated with facility staff qualifications. Section 5.3.1 will be deleted from the POTS and relocated to the QATR to provide an equivalent description of the requirements for facility staff qualifications. Relocating these responsibilities to the QATR is consistent with NRC Administrative Letter 95-06, Relocation of Technical Specification Administrative Controls Related to Quality Assurance. Therefore, the proposed deletion and relocation are acceptable.

Section 5.4, Training, establishes the requirements for training the facility staff. Providing staff qualification descriptions in the QATR is consistent with NRC Administrative Letter 95-06. Therefore, the proposed deletion and relocation are acceptable.

Section 5.4.1, Training, establishes the requirements for training of facility staff. The proposed change to this section to delete the requirements of ANSI/ANS 3.1-1993 from the POTS. Section 5.4.1 will be deleted from the POTS and relocated to the QATR to provide an equivalent description of the requirements for station staff qualifications. Relocating these responsibilities to the QATR is consistent with NRC Administrative Letter 95-06, Relocation of Technical Specification Administrative Controls Related to Quality Assurance. Therefore, the proposed deletion and relocation are acceptable.

Section 5.4.2, Training, establishes the requirements for the approved training and retraining program for the CFH. Specifies requirements for a CFH training program. Following the transfer of all spent fuel to the ISFSI, and the new provision in Section 4.3 prohibition from storing spent fuel in the spent fuel pool, there will no longer be a need for CFH, which obviates the need for the associated training program. Therefore, this proposed deletion is acceptable.

Section 5.8, Procedures, addresses requirements for procedures and various programs listed in later POTS sections. FCS is proposing to relocate the requirements of this section to the QATR, except for Section 5.8.1 (a), specifying procedures applicable to the safe storage of nuclear fuel recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February, 1978, and 5.8.3, associated with fuel assemblies to be placed in Region 2 of the spent fuel racks, which will be eliminated.

Section 5.8.1, Procedures, provides a description and requirements regarding administration of written procedures. The transfer of the administrative controls in this section is consistent with the guidance in NRC Administrative Letter 95-06, Relocation of Technical Specification Administrative Controls Related to Quality Assurance, and therefore, is acceptable. After these administrative controls are incorporated into the QATR, any future changes are controlled in accordance with§ 50.54(a). This will provide adequate control for the facility with all spent fuel located within the ISFSI. The relocation of administrative controls for procedures to the QATR is consistent with NRC Administrative Letter 95-06 and will have no impact on safe storage and maintenance of spent fuel in the ISFSI. Therefore, the proposed deletion and relocation of the requirements are acceptable.

Section 5.8.1 (a), Procedures, implements the guidance in Regulatory Guide 1.33, Revision 2, Appendix A, addresses safety related activities carried out during the operation phase of nuclear power plants, including wet storage of nuclear fuel in a spent fuel pool. Following the transfer of all spent fuel to the ISFSI, the spent fuel pool will no longer be used for spent fuel storage. Additionally, as discussed above, FCS is adding a limitation in TS 4.3, which prohibits storage of spent fuel in the spent fuel pool.

After spent fuel storage is no longer allowed in the spent fuel pool, the specifications included in this section, would no longer be needed, so the proposed deletion is acceptable.

DPR-40 Enclosure 1

Section 5.8.1 (d), Programs, implements written procedures and administrative policies associated with maintaining control and records of radiological effluent and monitoring for the ODCM, Radiation Protection Program, and Technical Specifications Bases Control. The transfer of the administrative controls for maintaining control and records of radiological effluent and monitoring for the ODCM, and Radiation Protection Program procedure and administrative policies from this section to the QATR is consistent with the guidance in NRC Administrative Letter 95-06, Relocation of Technical Specification Administrative Controls Related to Quality Assurance, and therefore, is acceptable. After the administrative controls are incorporated into the QATR, any future changes are controlled in accordance with§ 50.54(a). This will provide adequate control for the facility with all spent fuel located within the ISFSI. The relocation of these administrative requirements to the QATR is consistent with NRC Administrative Letter 95-06 and will have no impact on safe storage and maintenance of spent fuel in the ISFSI. The Technical Specification Bases Control program will be deleted in its entirety. All of the Bases in the existing PDTS are being eliminated with the proposed changes to the corresponding sections. Since all the Bases will be deleted, there will no longer be a need for a TS Bases control program. The removal of the requirements in this section will not reduce the effectiveness of the program control or procedures. Therefore, the proposed deletion and relocation of the requirements are acceptable.

Section 5.8.2, Procedures, control of temporary procedures, addresses change controls established for safety related activities carried out during the operation phase of nuclear power plants, including wet storage of nuclear fuel in a spent fuel pool. Following the transfer of all spent fuel to the ISFSI, the spent fuel pool will no longer be used for spent fuel storage. Additionally, as discussed above, FCS is adding a limitation in TS 4.3, which prohibits storage of spent fuel in the spent fuel pool. After spent fuel storage is no longer allowed in the spent fuel pool , the specifications included in this section would no longer be needed. Therefore, the proposed deletion is acceptable.

Section 5.8.3, Procedures, control of SFP Region 2. Section 5.8.3 implements the guidance which govern the selection of fuel assemblies to be placed in Region 2 of the spent fuel racks (Technical Specification 2.8). Following the transfer of all spent fuel to the ISFSI, the spent fuel pool will no longer be used for spent fuel storage. Additionally, as discussed above, FCS is adding a limitation in TS 4.3, which prohibits storage of spent fuel in the spent fuel pool. After spent fuel storage is no longer allowed in the spent fuel pool, the specifications included in this section, would no longer be needed.

Therefore, the proposed deletion is acceptable.

Section 5.9, Reporting Requirements, provides a description and requirements regarding reports that are to be submitted in accordance with 10 CFR 50.4. FCS is proposing to delete Section 5.9, "Reporting Requirements" , from the PDTS and relocate the requirements to the QATR which is a licensee- controlled document in their entirety. The two requirements are Section 5.9.4(a), "Annual Radiological Effluent Release Report" and Section 5.9.4(b), "Annual Radiological Environmental Operating Report". The proposed change is to delete this section in its entirety, and relocate the requirements verbatim to the QATR. After these administrative controls are incorporated into the QATR, any future changes are controlled in accordance with§ 50.54(a). This will provide adequate control for the facility with all spent fuel located within the ISFSI. The relocation of administrative controls for reporting requirements to the QATR is consistent with NRC Administrative Letter 95-06 and will have no impact on safe storage and maintenance of spent fuel in the ISFSI, and therefore is acceptable.

Section 5.1 0, Record Retention, establishes the requirements for maintaining records. The proposed change to this section deletes the requirements from the PDTS to maintain records as described in the DPR-40 Enclosure 1

QAP. These requirements are already located in the QATR so there is no reduction in the record retention requirements. The location of these requirements in the QATR is consistent with NRC Administrative Letter 95-06, Relocation of Technical Specification Administrative Controls Related to Quality Assurance. Therefore, the proposed deletion and relocation are acceptable.

Section 5.11.1, Radiation Protection Program, establishes the requirements for personnel radiation protection. The proposed change relocates Section 5.11 .1(c) to before the associated note to improve readability. This is an administrative change that does not change the technical content of the section.

Therefore, the proposed change to this section is acceptable.

Section 5.11.2, Radiation Protection Program, establishes the requirements for personnel radiation protection. The proposed change removes the provision for the SM on duty to maintain high radiation door keys. With this position no longer required, the control of the locked high radiation area door keys will reside only with the Manager of Radiation Protection (MRP) or designee. As justified in Section 5.1.2, with removal of all of the spent fuel from the SFP, a need for the SM for spent fuel management no longer exists. Therefore, the deletion of the specified position responsibility in this section is acceptable.

Section 5.16, Radiological Effluents and Environmental Monitoring Programs, specifies the requirements and controls for the site effluents. The requirement for a Radioactive Effluent Controls Program will be maintained in accordance with 10 CFR 50. 54( a). Since the intent of this section is to ensure that the Radioactive Effluent Controls Program continues to meet the requirements of 40 CFR 190, 10 CFR 20, 10 CFR 50.36(a), and 10 CFR 50, Appendix I, and since this requirement will be maintained in the QATR, the relocated requirement will continue to be subject to regulatory controls.

Section 5.16.1, Radioactive Effluent Controls Program , specifies requirements for the control of radioactive effluents and for maintaining doses to the public from effluents as low as reasonably achievable (ALARA). FCS proposes to delete this requirement from the POTS and relocate it to the QATR, except for the last sentence, "The provisions of SR 3.0.2 and SR 3.0.3 are applicable to Radioactive Effluent Controls Program Surveillance Frequencies." This sentence will be deleted from the POTS and not relocated. Since Section 3.0.2 and 3.0.3 are being deleted, as described above, this sentence is no longer germane. The specific requirements associated with noble gas in Sections 5.16.1 (b), (g), and (h) will not be relocated to the QATR since after all spent fuel is transferred to the ISFSI and contained within dry storage casks, there will no longer be a requirement to monitor for noble gases released from the facility. The requirement for a Radioactive Effluent Controls Program will be maintained in accordance with 10 CFR 50.54(a). Since the intent of this section is to ensure that the Radioactive Effluent Controls Program continues to meet the requirements of 40 CFR 190, 10 CFR 20, 10 CFR 50.36(a), and 10 CFR 50, Appendix I, and since this requirement will be maintained in the QATR, the relocated requirement will continue to be subject to regulatory controls. This change is consistent with similar relocations approved by NRC of former TS requirements into a separate requirements manual. Therefore, the proposed deletion and relocation of the requirement are acceptable.

Section 5.16.2, Radiological Environmental Monitoring Program, provides monitoring of radionuclides in the environs of the plant. FCS proposes to delete these requirements from the POTS in its entirety, and relocate them to the QATR. The requirement for this program will be maintained in accordance with 10 CFR 50.54(a). Since this requirement will be maintained in the , the relocated requirement will continue to be subject to regulatory controls. This change is consistent with similar relocations OPR-40 Enclosure 1

approved by NRC of former TS requirements into a requirements manual. Therefore, the proposed deletion and relocation of the requirement are acceptable.

Section 5.17, Offsite Dose Calculation Manual (ODCM) specifies how to document, review, and approve changes to the ODCM. FCS proposes to delete these requirements from the POTS and relocate them to the QATR. After the administrative controls are incorporated into the QATR, any future changes are controlled in accordance with§ 50.54(a). This will provide adequate control for the facility with all spent fuel located within the ISFSI. The relocation of these administrative requirements to the QATR is consistent with NRC Administrative Letter 95-06 and will have no impact on safe storage and maintenance of spent fuel in the ISFSI. The reference to Section 5.8.2 in Section 5.17(c),

associated with ODCM temporary change control will not be transferred to the QATR. The remaining requirements for changes to the ODCM, including temporary changes, will be controlled by the QATR.

This change is consistent with similar relocations approved by NRC of former TS requirements into the QATR. Since the intent of this section is to ensure that the ODCM continues to meet the requirements of 40 CFR 190, 10 CFR 20, 10 CFR 50.36(a), and 10 CFR 50, Appendix I, and since this requirement will be maintained in the QATR, the relocated requirements will continue to be subject to regulatory controls. Therefore, the proposed deletion and relocation of the requirements is acceptable.

Section 5.20, Technical Specification (TS) Bases Control Program, establishes the requirements to update and maintain plant basis. Currently, the TS Bases are all related to storage of spent fuel in the spent fuel pool, specifically the requirements in Section 4.3, which are being deleted as described above. Following transfer of all spent fuel to the ISFSI, the spent fuel pool will no longer be used for spent fuel storage. Since all the TS Bases will be deleted, there will no longer be a need for a TS Bases Control Program. Therefore, the proposed deletion of this section is acceptable.

DPR-40 Enclosure 1

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.1.1 10 CFR Part 50.82, Termination of License 10 CFR Part 50.82(a)(2) states "Upon docketing of the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel, or when a final legally effective order to permanently cease operations has come into effect, the 10 CFR Part 50 license no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel."

4.1.2 10 CFR 50.2, Definitions, Safety-Related Structures, Systems and Components 10 CFR 50.2 defines safety-related structures, systems, and components (SSCs) as those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

1. The integrity of the reactor coolant pressure boundary

2. The capability to shut down the reactor and maintain it in a safe shutdown condition; or

3. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in§ 50.34(a)(1) or§ 100.11 of this chapter, as applicable.

4.1.3 10 CFR Part 50.36, Technical Specifications In 10 CFR Part 50.36, the Commission established its regulatory requirements related to the content of TS. In doing so, the Commission placed emphasis on those matters related to the prevention of accidents and mitigation of accident consequences; the Commission noted that applicants were expected to incorporate into their TS "those items that are directly related to maintaining the integrity of the physical barriers designed to contain radioactivity." (Statement of Consideration, "Technical Specification for Facility Licenses; Safety Analysis Reports," 33 FR 18610 (December 17, 1968))

TS are required to include items in the following five categories:

(1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCO);

(3) surveillance requirements (SR);

(4) design features; and (5) administrative controls.

These criteria, which were subsequently codified in changes to Section 36 of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50.36)

(60 FR 36953), also pertain to the Technical Specification requirements for safe storage of spent fuel. However, the rule does not specify the particular DPR-40 Enclosure 1

requirements to be included in a plant's TS, as specified in 10 CFR 50.36 (iii) "A licensee is not required to propose to modify technical specifications that are included in any license issued before August 18, 1995, to satisfy the criteria in paragraph (c){2)(ii) of this section."

However a general discussion of these considerations is provided below:

Criterion 1 of 10 CFR Part 50.36(c)(2)(ii)(A) states that TS LCOs must be established for "installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary." Since no spent fuel will be present in the SFP at FCS following permanent transfer to the ISFSI, this criterion is not applicable.

Criterion 2 of 10 CFR Part 50.36(c)(2)(ii)(B) states that TS LCOs must be established for a "process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." The purpose of this criterion is to capture those process variables that have initial values assumed in the DBA and transient analyses, and which are monitored and controlled during power operation. While this criterion was developed for operating reactors, there are some DBA which continue to apply to a plant authorized only to handle, store, and possess nuclear fuel. There are no longer DBA in scope applicable to a station with spent fuel permanently moved out of the SFP into an ISFSI.

Criterion 3 of 10 CFR Part 50.36(c)(2)(ii)(C) states that TS LCOs must be established for SSCs that are part of the primary success path and which function or actuate to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The intent of this criterion is to capture into TS only those SSCs that are part of the primary success path of a safety sequence analysis. Also captured by this criterion are those support and actuation systems that are necessary for items in the primary success path to successfully function. The primary success path of a safety sequence analysis consists of the combination and sequences of equipment needed to operate (including consideration of the single failure criterion), so that the plant response to DBA and transients limits the consequences of these events to within the appropriate acceptance criteria . There are no DSAR described transients that will assume the failure of or challenge a fission product barrier. The DSAR postulated accident will no longer continue to apply to the station with the spent fuel permanently removed from the SFP.

Criterion 4 of 10 CFR Part 50.36(c){2)(ii)(D) states that TS LCOs must be established for SSCs that operating experience or probabilistic risk assessment has shown to be significant to public health and safety. The intent of this criterion is that risk insights and operating experience be factored into the establishment of TS LCOs. There are no postulated accident sequences remaining at FCS with the spent fuel permanently removed from the SFP.

DPR-40 Enclosure 1

10 CFR Part 50.36(c)(5) Administrative Controls, states that they "... are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner." The particular administrative controls to be included in the TS, therefore, are the provisions that the Commission deems essential for the safe operation of the facility that are not already covered by other regulations.

Accordingly, the NRC staff determined that administrative control requirements that are not specifically required under Section 50.36(c)(5), and which are not otherwise necessary to obviate the possibility of an abnormal situation or an event giving rise to an immediate threat to the public health and safety, may be relocated to more appropriate documents (e.g., Quality Assurance Program, Security Plan, or Emergency Plan), which are subject to regulatory controls.

Similarly, while the required content of TS administrative controls is specified in 10 CFR Part 50.36(c)(5), particular details may be relocated to licensee-controlled documents, where other regulations provide adequate regulatory control.

10 CFR Part 50.36(c)(6), Decommissioning, applies only to nuclear power reactor facilities that have submitted the certifications required by 10 CFR Part 50.82(a)(1). For such facilities, TS involving safety limits, limiting safety system settings, and limiting control system settings; limiting conditions for operation; surveillance requirements; design features; and administrative controls will be developed on a case-by-case basis.

This proposed amendment deletes or moves the portions of the previous FCS license that are no longer applicable or can be contained in alternate locations associated with a permanently defueled facility while modifying the remaining portions to correspond to the ISFSI only condition.

4.1.4 10 CFR Part 50.51, Continuation of License 10 CFR Part 50.51(b) states "Each license for a facility that has permanently ceased operations, continues in effect beyond the expiration date to authorize ownership and possession of the production or utilization facility, until the Commission notifies the licensee in writing that the license is terminated. During such period of continued effectiveness the licensee shall:

(1) Take actions necessary to decommission and decontaminate the facility and continue to maintain the facility, including, where applicable, the storage, control and maintenance of the spent fuel, in a safe condition, and (2) Conduct activities in accordance with all other restrictions applicable to the facility in accordance with the NRC regulations and the provisions of the specific 10 CFR Part 50 license for the facility."

4.1.5 10 CFR Part 50.54(hh), Loss Due to Aircraft Threats and Explosion or Fire 10 CFR Part 50.54(hh) establishes the requirements for developing, implementing and maintaining procedures and strategies for addressing potential aircraft threats and large area fires or explosions.

DPR-40 Enclosure 1

10 CFR Part 50.54(hh)(3) states; "This section does not apply to a nuclear power plant for which the certifications required under§ 50.82(a) or§ 52.11 O(a)(1) of this chapter have been submitted."

4.1.6 Quality Assurance Topical Report (QATR) which is a licensee- controlled document, is an appropriate candidates for relocations of administrative controls due to the controls imposed by such regulations as 10 CFR 50.59, Appendix B to 10 CFR Part 50, the existing NRC-approved QA plans and commitments to industry QA standards, and the established QA program change control process of 10 CFR 50.54(a).

4.1.7 Administrative Letter (AL) 95-06 NRC Administrative Letter (AL) 95-06, "Relocation of Technical Specification Administrative Controls Related to Quality Assurance," provides guidance to licensees requesting amendments that relocate administrative controls to NRC-approved QA program descriptions, where subsequent changes are controlled pursuant to 10 CFR 50.54(a). AL 95-06 provides specific guidance in the areas of: ( 1) independent safety engineering group, (2) reviews and audits, (3) procedure review process, and (4) records and record retention. Some relocations are specifically discussed in AL 95-06, while others are similar in nature. Relocations not specifically discussed in AL 95-06 were assessed with respect to the appropriateness of the relocation. Editorial changes are allowed without basis by 10 CFR 50.54(a)(3).

This proposed amendment deletes the portions of the previous FCS POTS that are no longer applicable to a permanently defueled facility with all irradiated fuel in dry storage within an ISFSI, while modifying the remaining portions to correspond to the SAFSTOR decommissioning condition, consistent with Standard Technical Specifications {STS).

4.1.8 Design Basis Accidents (DBA)

Chapter 14 of the FCS DSAR describes the safety analysis aspects of the plant that were evaluated to demonstrate that the plant could be operated safely and that radiological consequences from postulated accidents do not exceed the guidelines of 10 CFR Part 50.67.

The FCS DSAR, Chapter 14, Safety Analyses, currently addresses the DBA and transient scenarios applicable to FCS in the permanently defueled condition with irradiated fuel stored in the SFP. The majority of these postulated accidents are predicated on spent fuel being stored in the SFP. However, upon transfer of all irradiated fuel to storage in the ISFSI, the accident scenarios predicated on spent fuel storage in the SFP are no longer possible. The remaining accidents do not rely upon TS requirements for prevention or mitigation. With all of the spent fuel having been removed from the SFP, there are no remaining fuel related design basis accidents or transients in Chapter 14 of the DSAR.

DPR-40 Enclosure 1

4.2 Precedent 4.2.1 Several plants currently in the decommissioning process, including Kewaunee (Reference 6.8) (ML15261A236), Crystal River Unit 3 (Reference 6.9)

(ML16243A249), Vermont Yankee (Reference 6.1 0) (ML17206A200), and San Onofre Units 2 and 3 (Reference 6.11) (ML16355A014) have revised the POTS to that reflect the requirements for ISFSI only decommissioning .

4.3 No Significant Hazards Consideration The Omaha Public Power District (OPPD) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR Part 50.92, "Issuance of amendment," as discussed below:

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

Response: No.

The proposed amendment would modify the FCS renewed facility operating license and POTS by deleting the portions of the license and POTS that are no longer applicable to a facility with no spent nuclear fuel stored in the spent fuel pool, while modifying the remaining portions to correspond to all nuclear fuel stored within an ISFSI. This amendment becomes effective upon removal of all spent nuclear fuel from the FCS SFP and its transfer to dry cask storage within an ISFSI. The definition of safety-related structures, systems, and components (SSCs) in 10 CFR 50.2 states that safety-related SSCs are those relied on to remain functional during and following design basis events to assure:

1. The integrity of the reactor coolant boundary;

2. The capability to shutdown the reactor and maintain it in a safe shutdown condition; or

3. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in 10 CFR 50.43(a}(1) or§ 100.11 .

The first two criteria (integrity of the reactor coolant pressure boundary and safe shutdown of the reactor) are not applicable to a plant in a permanently defueled condition.

The third criterion is related to preventing or mitigating the consequences of accidents that could result in potential offsite exposures exceeding limits. However, after all nuclear spent fuel assemblies have been transferred to dry cask storage within an ISFSI, none of the SSCs at FCS are required to be relied on for accident mitigation. Therefore, none of the SSCs at FCS meet the definition of a safety-related SSC stated in 10 CFR 50.2. The proposed deletion of requirements in the FCS POTS does not affect systems credited in any accident analysis at FCS.

Chapter 14 of the FCS DSAR described the DBA related to the SFP. These postulated accidents are predicated on spent fuel being stored in the SFP. With the removal of the spent fuel from the SFP, there are no remaining spent fuel assemblies to be monitored DPR-40 Enclosure 1

and there are no credible accidents that require the actions of a Shift Manager, Certified Fuel Handler, or a Non-certified Operator to prevent occurrence or mitigate the consequences of an accident associated with nuclear fuel. The proposed changes do not have an adverse impact on the remaining decommissioning activities or any of their postulated consequences. The proposed changes related to the relocation of certain administrative requirements do not affect operating procedures or administrative controls that have the function of preventing or mitigating any accidents applicable to the safe management of irradiated fuel or decommissioning of the facility. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed changes eliminate the operational requirements and certain design requirements associated with the storage of the spent fuel in the SFP, and relocate certain administrative controls to the QATR which is a licensee- controlled document.

After the removal of the spent fuel from the SFP and transfer to the ISFSI, there are no spent fuel assemblies that remain in the SFP. Coupled with a prohibition against storage of fuel in the SFP, the potential for fuel related accidents is removed. The proposed changes do not introduce any new failure modes.

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

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

Response: No.

The removal of all spent nuclear fuel from the SFP into storage in casks within an ISFSI, coupled with a prohibition against future storage of fuel within the SFP, removes the potential for fuel related accidents.

The design basis and accident assumptions within the FCS DSAR and the POTS relating to safe management and safety of spent fuel in the SFP are no longer applicable. The proposed changes do not affect remaining plant operations, systems, or components supporting decommissioning activities.

The requirements for systems, structures, and components (SSG) that have been deleted from the FCS POTS are not credited in the existing accident analysis for any applicable postulated accident; and as such, do not contribute to the margin of safety associated with the accident analysis.

DPR-40 Enclosure 1

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

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

5.0 ENVIRONMENTAL CONSIDERATION

This amendment request meets the eligibility criteria for categorical exclusion from environmental review set forth in 10 CFR Part 51.22(c)(9) as follows. The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR Part 51.22 (c)(9). Therefore, pursuant to 10 CFR Part 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

DPR-40 Enclosure 1

6.0 REFERENCES

6.1. NO-FC-1 0, Quality Assurance Topical Report (QATR), Revision 11 6.2. Letter from OPPD (T. Burke) to USNRC (Document Control Desk), "Certification of Permanent Removal of Fuel from the Reactor" dated November 13, 2016 (LIC 0074) (ML16319A254) 6.3. Letter from OPPD (M. Fisher) to US NRC (Document Control Desk), "Post-Shutdown Decommissioning Activities Report (PSDAR)," dated March 30, 2017 (LIC-17-0033)

(ML17089A759) 6.4. Letter from OPPD (M. Fisher) to USNRC (Document Control Desk), "Irradiated Fuel Management Plan," dated March 31,2017 {LIC-17-0031) (ML17093A594) 6.5. NUH-003 Revision 16, "Updated Final Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel", dated July 27, 2017 (ML17213A407) 6.6. The NRC Administrative Letter (AL) 95-06, "Relocation of Technical Specification Administrative Controls Related to Quality Assurance", dated December 12, 1995 (ML031110271) 6.7. SECY-16-0142, Draft Final Rule- Mitigation of Beyond-Design-Basis Events, dated December 15, 2016 (ML16301A005) 6.8. Kewaunee Power Station "Technical Specifications to Reflect Permanent Removal of Spent Fuel from Spent Fuel Pool", dated September 14, 2015 (ML15261A236).

6.9. Crystal River Unit 3- License Amendment Request #323, Revision 0, Permanently Defueled Technical Specifications for the Independent Spent Storage Installation, dated August 31, 2016 (ML16243A249) 6.1 0. Vermont Yankee- "Revision of License and Permanently Defueled Technical Specifications to Reflect Permanent Removal of Spent Fuel from the Spent Fuel Pool",

dated July 20, 2017 (ML17206A200) 6.11 . San Onofre Nuclear Generating Station, "Units 1, 2 and 3, Amendment Applications 225, 272, and 257 ISFSI-only Technical Specifications", dated December 15, 2016, (ML16355A014)

DPR-40 Enclosure 1

ATTACHMENT 1 Fort Calhoun Station, Unit No. 1 Renewed Facility License No. DPR-40 Mark-up of 10 CFR Part 50 License and Technical Specifications Pages

[Word-processor mark-ups using "red line/strikeout" feature for "new text/deleted text" respectively]

OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT 1 RENEWED FACILITY LICENSE NO. DPR-40

1. The U.S. Nuclear Regulatory Commission {the Commission) having previously made the findings set forth in License No. DPR-40 issued August 9, 1973, has now found that:

A. The application to renew License No. DPR-40 filed by Omaha Public Power District (the licensee) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter 1, and all required notifications to other agencies or bodies have been duly made; B DELETED Actions have boon identified and have boon or will be taken with rosJ:Joct to ( 1) managing tho effects of aging during tho period of extended OJ:JOFation on tho functionality of structures and comJ:)ononts that have boon identified to require review under 10 CFR 54.21 (a)(1 ), and (2) time limited aging analyses that have boon identified to require review under 10 CFR 54.21(c), such that thoro is reasonable assurance that the activities authorized by this renewed license will continuo to be conducted in accordance with tho current licensing basis, as defined in 10 CFR 54 .3, for Fort Calhoun Station, Unit No. 1, and that any changes made to tho J:)lant's current licensing basis in order to somply with 10 CFR 54 .2Q(a) are in assord with tho Act and tho Commission's regulations; C. The facility will be maintained in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; D. There is reasonable assurance: (1) that the activities authorized by this renewed license can be conducted without endangering the health and safety of the public, and (2) that such activities will be conducted in compliance with the rules and regulations of the Commission; E. Omaha Public Power District is technically qualified and financially qualified to engage in the activities authorized by this renewed license in accordance with the rules and regulations of the Commission; F. Omaha Public Power District has satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements,"

of the Commission's regulations; G. The issuance of this renewed license will not be inimical to the common defense and security or to the health and safety of the public; DPR-40 Amendment No. 297###

H. After weighing the environmental, economic, technical, and other benefits of the facility against environmental costs and considering available alternatives, the Commission concludes that the issuance of Renewed License No. DPR-40 is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied; and I. The receipt, possession, and use of source, byproduct, and special nuclear material as authorized by the renewed license will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40, and 70, including but not necessarily limited to 10 CFR Sections 30.33, 40.32, 70.23 and 70.31.

2. On the basis of the forgoing findings regarding this facility, Facility Operating License No.

DPR-40, issued August 9, 1973, is superseded by Renewed Facility License No. DPR-40, which is hereby issued to the Omaha Public Power District, to read as follows:

A. This renewed license applies to the Fort Calhoun Station, Unit 1, a pressurized water nuclear reactor and associated equipment (the facility), which is owned by the Omaha Public Power District. The facility is located in Washington County, Nebraska, and is described in the Final Safety Analysis Report as supplemented, amended, and updated and the Environmental Report as supplemented and amended.

B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses Omaha Public Power District:

( 1) Pursuant to Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess and use the facility as required for fuel storage at the designated location in Washington County, Nebraska in accordance with the procedures and limitations set forth in this renewed license; (2) Pursuant to the Act and 10 CFR Parts 40 and 70, to possess at any time special nuclear material that was used as reactor fuel, in accordance with the limitations for storage, as described in the Final Safety Analysis Report, as supplemented, amended, and updated; (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, or sealed sources for radiation monitoring equipment calibration; and to possess any byproduct, source and special nuclear material as sealed neutron sources previously used for reactor startup and reactor instrumentation; and fission detectors DPR-40 Amendment No. 297

(4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or instrument calibration or when associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by operation of the facility.

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

A. DELETED B. Technical Specifications The Permanently Defueled Technical Specifications contained in Appendix A, as revised through Amendment No. ###, are hereby incorporated in the license replaced with the Perrnanently Defueled Technical Specifications (PDTS). Omaha Public Power District shall maintain the facility in accordance with the Permanently Defueled Technical Specifications.

C. Security and Safeguards Contingency Plans The Omaha Public Power District shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Fort Calhoun Station Security Plan, Training and Qualification Plan, Safeguards Contingency Plan," submitted by letter dated May 19, 2006.

DPR-40 Amendment No. 298###

D. DELETED E. DELETED F. DELETED G. Mitigation Strategy License Condition Develop and maintain strategies for addressing largo fires and explosions and that include tho follov;ing key areas:

(a) Fire fighting response strategy with the follmving elements:

1. Pro defined coordinated fire response strategy and guidance

2. Assessment of mutual aid fire fighting assets

a. Designated staging areas for equipment and materials

4. Command and control

5. Training of response personnel (b) Operations to mitigate fuel damage considering tho following:

1. Protection and use of personnel assets

2. Communications

a. Minimizing fire spread

4. Procedures for implementing integrated fire response strategy

5. Identification of readily available pro staged equipment

6. Training on integrated fire response strategy

7. Spent fuel pool mitigation measures (o) Actions to minimize release to include consideration of:

1. Water spray scrubbing

2. Dose to onsito responders

4. This license is effective as of the date of issuance and authorizes ownership and possession of Fort Calhoun Station until the Commission notifies the licensee in writing that the license is terminated.

FOR THE NUCLEAR REGULATORY COMMISSION Original Signed by:

J.E. Dyer J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments: 1. Appendix A- Permanently Defueled Technical Specifications

2. Appendix B - Deleted Date of Issuance: November 4, 2003 DPR-40 Amendment No. 297###

APPENDIX A TO OPERATI~JG RENEWED FACILIITY LICENSEWG NO. DPR-40 PERMENENTLY DEFUELED TECHNICAL SPECIFICATIONS FOR THE FORT CALHOUN STATION UNIT NO.1 OMAHA PUBLIC POWER DISCTRICT DOCKET NO. 50-285 DPR-40

PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS TABLE OF CONTENTS DEFINITIONS 2.0 liMITING CONDITIONS FOR OPERATION 2.8 Fuel Handling 3.0 SURVEillANCE REQUIREMENTS 3.2 Equipment and Sampling Tests 4.0 DESIGN FEATURES 4.1 Site 4.3 Fuel Storage 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.2 Organization 5.3 Facility Staff Qualifications 5.4 Training 5.8 Procedures 5.Q Reporting Requirements 5.Q.4 Unique Reporting Requirements 5.10 Record Retention 5.11 Radiation Protection Program 5.16 Radiological Effluents and Environmental Monitoring Programs 5.16.1 Radioacti*;o Effluent Controls Program 5.16.2 Radiological Environmental Monitoring Program 5.17 Offsito Dose Calculation Manual (ODCM) 5.20 Technical Specification (TS) Bases Control Program TOC - Page 1 Amendment No. 2Q7###

DPR-40

PERMANENTlY DEFUElED TECHNICAl SPECIFICATIONS TABlES TABlE OF CONTENTS TABlE SECTION Minimum Frequencies for Sampling Tests ....... .... ......................................... .... ......................... Section 3.2 35 Minimum Frequencies for Equipment Tests .... ... ..... ...... ............................................................ Section 3.2 5.2 1 Minimum Shift Crew Composition .......... ............. ............... .......... ...... .. ... ...... .... ......................... Section 5.0 TOG Page 2 Amendment No. 297 DPR-40

PERMANENTlY DEfUElED TECWNICAl SPECifiCATIONS fiGURES TABlE Of CONTENTS fiGURE DESCRIPTION SECTION 2 10 Spent fuel Pool Region 2 Storage Criteria .... ... ............................. ..... ......... ... ... ..... ............... .... Seotion 2.8 TOG Page 3 Amendment No. 297 DPR-40

PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS DEFINITIONS The following terms are defined fer uniform interpretation of these Specifications.

FUEL STORAGE AND HANDLING CONDITIONS fuel Handling Operations Any operation involving the shuffling, removal, or-replacement of irradiated fuel. The suspension of any fUEL HANDLING OPERATIONS shall not preclude completion of movement of a component to a safe, conservative position.

Certified Fuel Handler (CFH)

A CERTIFIED FUEL HANDLER is an individual who complies with provisions of the CERTIFIED FUEL HANDLER training program required by Technical Specification 5.4.2.

Non Certified Operator (NCO)

A NON CERTIFIED OPERATOR is a non licensed operator who complies with the applicable training requirements of Technical Specification 5.4.1, but is not a CERTIFIED FUEL HANDLER.

MISCEllANEOUS DEFINITIONS Actions ACTIONS shall be that part of a specification that prescribes required actions to be taken under designated conditions within specified completion times.

Operable Operability A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power sources, cooling and seal water, lubrication, and other auxiliary equipment that are required fer the system, subsystem, train, component or device to perform its specified safety functions(s) are also capable of performing their related support function(s).

Definitions Page 1 Amendment No. 207 DPR-40

DEFINITIONS MISCELLANEOUS DEFINITIONS (continued)

Offsite Dose Calculation Manual (ODCM)

The document(s) that contain the methodology and parameters used in the calculations of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent radiation monitoring \Alarn/High (trip) Alarm setpoints, and in the conduct of the Environmental Radiological Monitoring Program. Tho ODCM shall also contain:

1) The Radiological Effluent Controls and the Radiological Environmental Monitoring Program required by Specification 5.16.

2) Descriptions of the information that should be included in the Annual Radiological Environmental Operating Reports and Annual Radioactive Effluent Release Reports required by Specifications 5.9.4.a and 5.9.4 .b.

Unrestricted Area Any area at or beyond the site boundary access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

Definitions Page 2 Amendment No. 297 DPR-40

Sections 1.0 through 3.17 a.+ and Definitions have been deleted in their entirety.

2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Fuel Handling 2.8.1 DELETED 2.8.2 DELETED 2.8.3 Fuel Handling Operations Spent Fuel Pool 2.8.3(1) Spent Fuel Assemblv Storage Applicability Applies to storage of spent fuel assemblies 'i'.'honevor any irradiated fuel assembly is stored in Region 2 (including peripheral cells) of tho spent fuel pool.

Objective To minimize the possibility of an accident occurring during FUEL HANDLING OPERATIONS that could affect public health and safety.

Specification The combination of initial enrichment and burn up of each spent fuel assembly stored in Region 2 (including peripheral cells) of the spent fuel pool shall be v1ithin the acceptable burnup domain of Figure 2 10.

Required Actions (1) 'Nith the requirements of the LCO not met, initiate action to move the noncomplying fuel assembly immediately.

2.8 Page 1 Amendment No. 297 DPR-40

FIGURE 2-10 45000 -r------.-------r------.-------r------,-------,----~

40000 - ~----~-------r------;---~--r-----_,-------r-------v 350 ACCEP TAB LE BURN UP DOM AIN Q_

=l z

a:

=l 20000 - ~-----;----~-r------

m

_j w

=l LL UN ACCEPTABLE BURNU P DOMAIN (Requ i res Reg 10n 1 Stor age) 0-~~--~~-r.-~~-r~~-r.-,-+-r~,-+-r,-,1~1'1-, 1 '1-,1~1 1 .5 3.0 4.0 4 .5 5.0 ITIAL ENRICHMENT, wt% - 235 LIM TING BURNUP CRI TERIA FO R AC -PTABLE STORAGE I REGION 2 NO TES : 1. C el assembly (:::;: 4. 5% average U-235 ennohmentl ec on 1colly led WIth o f ul l l ength CEA mo':l be located on;jwhere n Regton 2 .

enpherol cel1 s ore those odjocent to the Spent Fuel Pool oH or the cesk l eydown ae-ea.

2.8- Page 2 Amendment No. 297 DPR-40

2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Fuel Handling 2.8.3 Fuel Handling Operations Spent Fuel Pool (continued) 2.8.3(2) Spent Fuel Pool 'Nater Level Applicability Applies to the v'later level of the spent fuel pool during FUEL HANDLING OPERATIONS in the spent fuel pool.

Objective To minimize the consequences of a fuel handling accident during FUEL HANDLING OPERATIONS in the spent fuel pool that could affect public health and safety.

Specification The spent fuel pool 'Nater level shall be > 23 ft. above the top of irradiated fuel assemblies seated in the storage racks.

Required Actions (1) 'Nith the spent fuel pool \Vater level not within limits, suspend FUEL HANDLING OPERJ\TIONS in the spent fuel pool immediately.

2.8.3(3) Spent Fuel Pool Boron Concentration Applicability Applies to the boron concentration of the spent fuel pool when fuel assemblies are stored in the spent fuel pool.

Objective To minimize the possibility of an accident that could affect public health and safety from occurring 'Nhen fuel assemblies are stored in the spent fuel pool.

Specification The spent fuel pool boron concentration shall be > 500 ppm.

Required Actions (1) VVith the spent fuel pool boron concentration< 500 ppm, suspend FUEL HANDLING OPERATIONS in the spent fuel pool immediately, and (2) Restore spent fuel pool boron concentration to > 500 ppm immediately.

2.8 Page 3 Amendment No. 297 DPR-40

2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Fuel Handling 2.8.3 Fuel Handling Operations Spent Fuel Pool (continued) 2.8.3(4) DELETED 2.8.3(5) DELETED 2.8.3(6) DELETED 2.8 Page 4 Amendment No. 297 DPR-40

2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Fuel Handling Basis 2.8.3(1) Spent Fuel Assembly Storage The spent fuel pool is designed for noncriticality by use of neutron absorbing material.

The restrictions on the placement of fuel assemblies vtithin the spent fuel pool, according to Figure 2 10, and the accompanying LCO, ensures that the kef!" of the spent fuel pool al'.vays remains < 0.95 assuming the pool to be flooded with unborated \Vater.

A spent fuel assembly may be transferred to the spent fuel pool Region 2 provided an independent verification of assembly burn ups has been completed and the assembly burn up meets the acceptance criteria identified in Figure 2 10. VI.' hen the configuration of fuel assemblies stored in Region 2 (including the peripheral cells) is not in accordance with Figure 2 10, immediate action must be taken to make the necessary fuel assembly movement(s) to bring the configuration into compliance with Figure 2 10. Acceptable fuel assembly burn up is not a prerequisite for Region 1 storage because Region 1 will maintain any type of fuel assembly that the plant is licensed for in a safe, coolable, subcritical geometry.

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

2.8.3(2) Spent Fuel Pool VIJater Level The minimum water level in the spent fuel pool meets the assumption of iodine decontamination factors following a fuel handling accident. VI/hen the water level is lower than the required level, the movement of irradiated fuel assemblies in the spent fuel pool is immediately suspended. This effectively precludes a fuel handling accident from occurring in the spent fuel pool. Suspension of FUEL HANDLING OPERATIONS shall not preclude completion of movement of a component to a safe, conservative position. \.Yhen "immediately" is used as a completion time, the required action should be pursued without delay and in a controlled manner.

2.8.3(3) Spent Fuel Pool Boron Concentration The basis for the 500 ppm boron concentration requirement with Boral poisoned storage racks is to maintain the k'**' below 0.95 in the event a misloaded unirradiated fuel assembly is located next to a spent fuel assembly. A misloaded unirradiated fuel assembly at maximum enrichment condition, in the absence of soluble poison, may result in exceeding the design effective multiplication factor. A misloaded irradiated fuel assembly is bounded by this requirement. Soluble boron in the spent fuel pool water, for which credit is permitted under these conditions, *~t~ould assure that the effective multiplication factor is maintained substantially less than the design condition.

This LCO applies whenever fuel assemblies are stored in the spent fuel pool. The boron concentration is periodically sampled in accordance with Specification 3.2 .

Sampling is performed periodically when fuel is stored in the spent fuel pool.

2.8 Page 5 Amendment No. 207 DPR-40

2.0 liMITING CONDITIONS FOR OPERATION 2.8 Fuel Handling Bases (continued) 2.8.3(3) Spent Fuel Pool Boron Concentration (continued)

VVhen "immediately" is used as a completion time, the required action should be pursued without delay and in a controlled manner. Suspension of FUEL HANDLING OPERl\TIONS shall not preclude completion of movement of a component to a safe, conservative position.

References (1) FSAR, as updated, Section 9.5 (2) FSAR, as updated, Section 14.18 Sections 2.9 through 2.23 have been deleted in their entirety.

2.8 Page 6 Amendment No. 297 DPR-40

3.0 SURVEillANCE REQUIREMENTS 3.0.1 Each surveillance requirement shall be performed 'Nithin the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval.

3.0.2 The surveillance intervals are stated in the individual Specifications.

3.0.3 The provisions of Specifications 3.0.1 and 3.0.2 are applicable to all codes and standards referenced within the Technical Specifications. The requirements of the Technical Specifications shall have precedence over the requirements of the codes and standards referenced 'Nithin the Technical Specifications.

3.0.4 Surveillance Requirements shall be met during the specified conditions in the individual Limiting Conditions for Operation, unless othenuise stated in the Surveillance Requirement.

Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the OPERABILITY requirements for the corresponding Limiting Condition for Operation. Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specifications 3.0.1 and 3.0.2, shall constitute noncompliance with the OPERABILITY requirements for the corresponding Limiting Condition for Operation except as provided in Specification 3.0.5. Tho time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not boon performed. Surveillance Requirements do not have to be performed on inoperable equipment.

3.0.5 If it is discovered that a Surveillance was not performed *.vithin its specified surveillance interval, then compliance with the requirement to declare the OPERABILITY requirements for the Limiting Condition for Operation not met may be delayed, from tho time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified surveillance interval, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed.

If the Surveillance is not performed 'Nithin the delay period, the OPERABILITY requirements for tho Limiting Condition for Operation must immediately be declared not met, and tho applicable ACTION(S) must be entered.

\'Vhen the Surveillance is performed within the delay period and the Surveillance is not met, the OPERABILITY requirements for the Limiting Condition for Operation must immediately be declared not met, and the applicable ACTION(S) must be entered.

3.0 Page 1 Amendment No. 297 DPR-40

3.0 SURVEILLANCE REQUIREMENTS

&sis Specifications 3.0.1 through 3.0.5 establish the general requirements applicable to Surveillance Requirements.

Specification 3.0.1 establishes the limit for which the specified time interval for Surveillance Requirements may be extended. It permits an aiiO'IJable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of station conditions that may not be suitable for conducting the surveillance; e.g., transient conditions or other ongoing surveillance or maintenance activities. It is not intended that this provision be used repeatedly as a convenience to extend surveillance intervals beyond that specified. Tho limitation of Specification 3.0.1 is based on engineering judgement and tho recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements.

This provision is sufficient to ensure that tho reliability ensured through surveillance activities is not significantly degraded beyond that obtained from tho specified surveillance interval.

The provisions of Specification 3.0.2 define tho surveillance intervals for use in the Technical Specifications. This clarification is provided to ensure consistency in surveillance intervals throughout the Technical Specifications. A few surveillance requirements have uncommon intervals. In such a case tho surveillance interval shall be performed as defined by the individual specifications.

Specification 3.0.3 extends the testing interval required by codes and standards referenced by tho Technical Specifications. This clarification is provided to remove any ambiguities relative to the frequencies for performing the required testing activities. Under the terms of this specification, tho more restrictive requirements of the Technical Specifications take precedence over the codes and standards referenced therein. Tho requirements of regulations take precedence over the TS. Therefore, test intervals governed by regulation cannot be extended by the TS.

Specification 3.0.4 establishes tho requirement that Surveillances must be met during other specified conditions in tho Specification for 'Nhich tho requirements of the Limiting Condition for Operation apply, unless otherv.<iso specified in the individual Surveillances.

This Specification is to ensure that Surveillances are performed to verify the OPERABILITY of systems and components, and that variables are within specified limits.

Failure to meet a Surveillance within tho specified surveillance interval, in accordance with Specifications 3.0.1 and 3.0.2, constitutes a failure to meet the OPERABILITY requirements for the corresponding Limiting Condition for Operation.

Systems and components are assumed to be OPERJ\BLE when the associated Surveillances have been mot. Nothing in this Specification, however, is to be construed as implying that systems or components are OPER/\BLE when either:

a. The systems or components are known to be inoperable, although still meeting tho Surveillances or 3.0 Page 2 Amendment No. 207 DPR-40

3.0 SURVEILLANCE REQUIREMENTS

~(continued)

b. The requirements of the Surveillance(s) are kno'Nn to be not met between required Surveillance performances.

Surveillances do not have to be performed when the unit is in a specified condition for Nhich the requirements of the associated Limiting Condition for Operation are not 1

applicable unless otherwise specified. The Surveillances associated with a special test exception (STE) are only applicable when the STE is used as an allowable exception to the requirements of a Specification.

Unplanned events may satisfy the requirements (including applicable acceptance criteria) for a given Surveillance. In this case, the unplanned event may be credited as fulfilling the performance of the Surveillance. This allowance includes those Surveillances 'Nhose performance is normally precluded in a specified condition.

Surveillances, including Surveillances invoked by ACTIONS, do not have to be performed on inoperable equipment because the /\CTIONS define the remedial measures that apply. Surveillances have to be met and performed in accordance with Specifications 3.0.1 and 3.0.2, prior to returning equipment to OPERABLE status.

Specification 3.0.5 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not been completed 1Nithin the specified surveillance interval. A delay period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified surveillance interval, whichever is greater, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with Specification 3.0.1, and not at the time that the specified surveillance interval vms not met.

This delay period provides adequate time to complete Surveillances that have been missed. This delay period permits the completion of a Surveillance before complying 'Nith ACTIONS or other remedial measures that might preclude completion of the Surveillance.

The basis for this delay period includes consideration of unit conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance 1Nith the requirements.

VVhen a Surveillance \Vith a surveillance interval based not on time intervals, but upon specified situations, or requirements of regulations (e.g., as modified by approved exemptions, etc.) is discovered to not have been performed 'Nhen specified, Specification 3.0.5 allows for the full delay period of up to the specified surveillance interval to perform the Surveillance.

3.0 Page 3 Amendment No. 297 DPR-40

3.0 SURVEILLANCE REQUIREMENTS

~(continued)

Hmuever, since there is not a time interval specified, the missed Surveillance should be performed at the first reasonable opportunity. Specification 3.0.5 provides a time limit for, and allmvances for the performance of, Surveillances that become applicable as a consequence of changes imposed by ACTIONS.

Failure to comply 'Nith specified surveillance intervals for Surveillance Requirements is expected to be an infrequent occurrence. Use of the delay period established by Specification 3.0.5 is a flexibility which is not intended to be used as a convenience to extend surveillance intervals. While up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the limit of the specified surveillance interval is provided to perform the missed Surveillance, it is expected that the missed Surveillance will be performed at the first reasonable opportunity. The determination of the first reasonable opportunity should include consideration of the impact on plant risk (from delaying the Surveillance as well as any plant configuration changes required to perform the Surveillance) and impact on any analysis assumptions, in addition to station conditions, planning, availability of personnel, and the time required to perform the Surveillance. This risk impact should be managed through the program in place to implement 10 CFR 50.65(a)(4) and its implementation guidance, NRC Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Po'Ner Plants." This Regulatory Guide addresses consideration of temporary and aggregate risk impacts, determination of risk management action thresholds, and risk management actions. The missed Surveillance should be treated as an emergent condition as discussed in the Regulatory Guide. The risk evaluation may use quantitative, qualitative, or blended methods. The degree of depth and rigor of the evaluation should be commensurate Y.'ith the importance of the component. Missed Surveillances for important components should be analyzed. If the results of the risk evaluation determine the risk increase is significant, this evaluation should be used to determine the safest course of action. All missed Surveillances will be placed in the corrective action program.

If a Surveillance is not completed within the allowed delay period, then the equipment is considered inoperable or the variable is considered outside the specified limits and the allowable outage time limits of the/\CTIONS for the applicable Limiting Condition for Operation begin immediately upon expiration of the delay period. If a Surveillance is failed within the delay period, then the equipment is inoperable, or the variable is outside the specified limits and the allowable outage time limits of the ACTIONS for the applicable Limiting Condition for Operation begin immediately upon the failure of the Surveillance.

Completion of the Surveillance 'Nithin the delay period allowed by this Specification, or within the allo'Nable outage time limits of the ACTIONS, restores compliance with Specification 3.0.4.

3.0 Page 4 Amendment No. 297 DPR-40

3.0 SURVEILLANCE REQUIREMENTS 3.1 DELETED 3.2 Equipment and Sampling Tests Applicability Applies to plant equipment and conditions related to safe storage and handling of nuclear fueh.

Objective To specify the minimum frequency and type of surveillance to be applied to critical plant equipment and conditions.

Specifications Equipment and sampling tests shall be conducted as specified in Tables 3 4 and 3 5.

The equipment testing and system sampling frequencies specified in Tables 3 4 and 3 5 are considered adequate, based upon experience, to maintain the status of the equipment and systems so as to assure safe storage and handling of nuclear fuel. Thus, those systems vvhere changes might occur relatively rapidly are sampled frequently and those static systems is not subject to changes are sampled less frequently.

3.2 Page 1 Amendment No. 297 DPR-40

TABLE 3 4 MINIMUM FREQUENCIES FOR SAMPLING TESTS Type of Measurement Sample and Analysis and /\nalvsis Frequency

1. Spent Fuel Pool Boron Concentration See Footnote (1) below (1) Weekly 'Nhen fuel assemblies are stored in the spent fuel pool.

3.2 Page 2 Amendment No. 297 DPR-40

TABlE 3 5 MINIMUM FREQUENCIES FOR EQUIPMENT TESTS Test Frequenc\'

1. Spent Fuel Pool Test neutron poison samples for 1, 2, 4, 7, and 10 years after Racks dimensional change, weight, neutron installation, and every 5 years aat~te~n*u~affiti~on~e~ha~n~g*eHa*n*dHs*p~e~cffiifi~c-----------4t~he~r~oaaf~to~r.

gravity chango.

2. Spent Fuel Pool Level Verify spent fuel pool water level is :.c: 23 ft. Prior to commencing, and weekly above the top of irradiated fuel assemblies during FUEL HANDLING seated in tho storage racks. OPERATIONS in the spent fuel

3. Spent Fuel Assembly Verify by administrative means that initial Prior to storing the fuel assembly Storage enrichment and burnup of tho fuel assembly in Region 2 (including peripheral is in accordance 'Nith Figure 2 10. cells).

3.2 Page 3 Amendment No. 297 DPR-40

3.0 SURVEillANCE REQUIREMENTS Sections 3.3 through 3.17 ha,Je been deleted in their entirety.

3.3 Page 1 Amendment No. 297 DPR-40

4.0 DESIGN FEATURES 4.1 Site The site for Fort Calhoun Station Unit No. 1 is in Washington County, Nebraska, on the west bank of the Missouri River and approximately nineteen miles north, northwest of the city of Omaha, Nebraska. The exclusion area description, as defined in 10 CFR Part 100, Section 100.3(a), is located in the Final Safety Analysis Report, as updated.

4.2 DELETED 4.3 Fuel Storage Spent Fuel shall not be stored in the spent fuel pool.

4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained wi-tffi

a. Fuel assemblies having a maximum U 235 enrichment of 4.5 weight percent,

b. k..# < 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.5 of the FS/\R, as updated,

c. /\ nominal 8.6 inch center to center distance between fuel assemblies placed in Region 2, the high density fuel storage racks,

d. /\ nominal 9.8 inches (East 'Nest) by 10.3 inches (North South) center to center distances between fuel assemblies placed in Region 1, the low density fuel storage racks,

e. Partially spent fuel assemblies with a discharge burnup in the "acceptable domain" of Figure 2 10 for "Region 2 Unrestricted" may be allo*Ned unrestricted storage in any of the Region 2 fuel storage racks in compliance \"'ith Reference (1 ).

f. Partially spent fuel assemblies VJith a discharge burnup between the "acceptable domain" and "Peripheral Cells" of Figure 2 10 may be allowed unrestricted storage in tho peripheral cells of the Region 2 fuel storage racks in compliance with Reference (1 ).

g. Partially spent fuel assemblies with a discharge burnup in the "unacceptable domain" of Figure 2 10 will be stored in Region 1 in compliance with Reference (1 ).

4.0- Page 1 Amendment No. 297###

DPR-40

4.0 DESIGN FEATURES 4.3 Fuel Storage 4.3.1 Criticality (continued) 4.3.1.2 DELETED 4.3.1.3 DELETED 4 .3.2 Drainage The spent fuel storage pool is designed and shall bo maintai~ed t? prevent inadvertent draining of the pool below 23 ft. above the top of 1rrad1ated fuel assemblies seated in the storage racks.

4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained *.vith a storage capacity limited to no more than 1083 fuel assemblies.

References:

(1) Letter from R. Wharton (NRC) to T. Patterson (OPPD), Amendment 174 to Facility Operating License No. DPR 40, (TAG NO. M94789) Dated July 30, 1996, NRC 96 0126.

4.0 Page 2 Amendment No. 297 DPR-40

Sections 5.1 through 5.10 have been deleted in their entirety.

5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.1.1 The plant manager shall be responsible for the overall facility, and the maintenance ef fuel, shall delegate in v1riting the succession to this responsibility during his absence.

5.1.2 The Shift Manager shall be responsible for the shift command function.

5.2 Organimtion 5.2.1 Onsite and offsite organizations shall be established for the facility and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear fuel.

a. Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels to and including all organization positions. Those relationships shall be documented and updated, as appropriate, in the form of organizational charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation.

These requirements shall be documented in the FSAR, as updated.

b. The plant manager shall be responsible for the overall facility and shall have control over those onsite activities necessary fur safe storage and maintenance of the nuclear fuel.

c. The corporate officer with responsibility for overall management of nuclear fuel shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the station to ensure safe management of nuclear fuel.

d. The individuals who train the CERTIFIED FUEL HANDLERs and NON CERTIFIED OPERATORS, and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their ability to perform their assigned functions.

5.2.2 Facility Staff The facility staff organization shall be as described in the FSAR, as updated and shall function as follows:

a. The minimum number and type of personnel required onsite fur each shift shall be as shovm in Table 5.2 1.

5.0- Page 1 Amendment No. 297 ###

DPR-40

5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.2 Facility Staff (continued)

b. An individual qualified in Radiation Protection Procedures shall be onsite during fuel handling operations or movement over storage racks containing fuel. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

c. The Shift Manager shall be a CERTIFIED FUEL HANDLER.

d. Fire protection program responsibilities are assigned to those positions and/or groups designated by asterisks in the FSAR, as updated.

e. DELETED 5.0 Page 2 Amendment No. 297 DPR-40

T.~BLE 5.2 1 MINIMUM SHIFT CREW COMPOSITION ~

Staffing Category Minimum Staffing CERTIFIED FUEL HANDLER 1(4 NON CERTIFIED OPERJ\TOR 1~

(i) This includes the individual with CERTIFIED FUEL HANDLER qualification supervising FUEL HANDLING OPERATIONS.

(ii) Shift crew composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 5.2 1 provided no fuel handling operations or movement over storage racks containing fuel is in progress. This provision does not permit any shift crev.' position to be unmanned upon shift change due to an oncoming shift crew member being late or absent.

(iii) At least one of these individuals must be in the control room at all times when fuel is in the Spent Fuel Pool.

(iv) The NON CERTIFIED OPERi\TOR position may be filled by a CERTIFIED FUEL HANDLER.

5.0 Page 3 Amendment No. 297 DPR-40

5.0 ADMINISTRATIVE CONTROLS 5.3 Facility Staff Qualification 5.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1 1971 for comparable positions with the requirement exception for those individuals in lieu of holding a Senior Reactor Operator license will be qualified as a CERTIFIED FUEL HANDLER, and with tho exception of tho Manager Radiation Protection (MRP),

who shall moot tho requirements sot forth in Regulatory Guide 1.8, Revision 3, dated May 2000, entitled "Qualification and Training of Personnel for Nuclear Power Plants."

5.4 Training 5.4 .1 A retraining and replacement training program for the plant staff shall be maintained under the direction of tho Plant Manager or designee and shall meet or exceed tho requirements of Section 6 of ANSI/ANS 3.1 1993, as modified by Regulatory Guido 1.8, Revision 3, dated May 2000.

5.4 .2 An NRC approved training and retraining program for the CERTIFIED FUEL HANDLER shall be maintained.

5.5 Not Used 5.6 Not Used 5.7 Not Used 5.8 Procedures 5.8.1 \IIJritten procedures and administrative policies shall be established, implemented and maintained covering the following activities:

a. The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix. A, 1978;

b. DELETED

c. Not Used

d. All applicable programs specified in Specifications 5.11, 5.16, and 5.20.

5.0 Page 4 Amendment No. 297 DPR-40

5.0 ADMINISTRP..TIVE CONTROLS 5.8 Procedures (continued) 5.8.2 Temporary changes to procedures of 5.8.1 above may be made provided:

a. The intent of the original procedure is not altered.

b. The change is approved by two members of the facility supervisory staff, at least one of whom is qualified as a CERTIFIED FUEL HANDLER.

c. The change is documented, reviewed by a qualified reviewer and approved by either the plant manager or the department head designated by Administrative Controls as the responsible department head for that procedure within 14 days of implementation.

5.8.3 VIJritten procedures shall be implemented which govern the selection of fuel assemblies to be placed in Region 2 of the spent fuel racks (Technical Specification 2.8). These procedures shall require an independent verification of initial enrichment requirements and fuel burnup calculations for a fuel bundle to assure the "acceptance" criteria for placement in Region 2 are met. This independent verification shall be performed by individuals or groups other than those who performed the initial acceptance criteria assessment, but who may be from the same organization.

5.9 Reporting Requirements In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the appropriate NRC Regional Office unless otherwise noted.

5.9.1 Not Used 5.9.2 Not Used 5.9.3 DELETED 5.9.4 Unique Reporting Requirements

a. Annual Radioactive Effluent Release Report The Annual Radioactive Effluent Release Report covering the station during the previous calendar year shall be submitted before May 1 of each year. Tho report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be 1) consistent v:ith the objectives outlined in the ODCM and PCP, and 2) in conformance with 10 CFR 50.36a. and Section IV.B.1 of Appendix I to 10 CFR 50.

5.0 Page 5 Amendment No. 297 DPR-40

5.0 ADMINISTRATIVE CONTROlS 5.9 Reporting Requirements 5.9.4 Unique Reporting Requirements (continued)

b. Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the station during the previous calendar year shall be submitted before May 1 of each year.

The report shall inelude summaries, interpretations, and analysis of trends of the results of the Radiologieal Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) tho ODCM and (2) Seetion IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR 50.

c. Not Used 5.9.5 DELETED 5.9.6 DELETED 5.10 Reeord Retention 5.1 0.1 Reeords shall be retained as deseribed in tho Quality Assurance Program.

5.11 Radiation Protection Program Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

5.11 .1 In lieu of the "control device" required by paragraph 20.1601 (a) of 10 CFR Part 20, and as an alternative method allowed under§ 20.1601(c), each high radiation area (as defined in § 20.1601) in which the intensity of radiation is 1000 mrem/hr or less shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by required issuance of a Radiation Work Permit.* Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.

• Radiation Protection personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they comply with approved radiation protection procedures for entry into high radiation areas.

5.0- Page 2 6 Amendment No. 29+ ###

DPR-40

5.0 ADMINISTRATIVE CONTROLS 5.11 Radiation Protection Program (continued)

c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Manager-Radiation Protection (MRP) in the Radiation Work Permit.

5.11.2 The requirements of 5.11.1, above, shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem/hr** but less than 500 rads/hr***

(Locked High Radiation Area). In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Manager on duty and/or the MRP (or designee) with the following exception:

a. In lieu of the above, for accessible localized Locked High Radiation Areas located in large areas such as containment, where no lockable enclosure exists in the immediate vicinity to control access to the Locked High Radiation Area and no such enclosure can be readily constructed, then the Locked High Radiation Area shall be:

i. roped off such that an individual at the rope boundary is exposed to 1000 mrem/hr or less, ii conspicuously posted, and iii a flashing light shall be activated as a warning device.

• • At 30 centimeters (12 inches) from the radiation source or from any surface penetrated by the radiation.

• • • At 1 meter from the radiation source or from any surface penetrated by the radiation.

Sections 5.12 through 5.24 have been deleted in their entirety.

5.12 DELETED 5.13 DELETED 5.14 DELETED 5.15 DELETED 5.0- Page 2 7 Amendment No. 297###

DPR-40

5.0 ADMINISTRATIVE CONTROlS 5.16 Radiological Effluents and Environmental Monitoring Programs The following programs shall be established, implemented, and maintained.

5.16.1 Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for control of radioactive effluents and for maintaining the doses to individuals in UNRESTRICTED AREAs from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the functionality of radioactive liquid and gaseous radiation monitoring instrumentation including functionality tests and setpoint determination in accordance with the methodology in the ODCM.

b. Limitations on the concentration of radioactive material, other than dissolved or entrained noble gases, released in liquid effluents to unrestricted areas conforming to ten times 10 CFR 20.1001 20.2401, Appendix B, Table 2, Column

2. For dissolved or entrained noble gases, the concentration shall be limited to 2.0 E 04 ttCi/ml total activity.

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM.

d. Limitations on the annual and quarterly doses or dose commitment to individuals in unrestricted areas from radioactive materials in liquid effluents released to unrestricted areas conforming to Appendix I to 10 CFR Part 50.

e. Determination of cumulative doses from radioactive effluents for the current calendar quarter and current calendar year in accordance v1ith the ODCM on a quarterly basis.

f. Limitations on the functionality and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity in plant effluents.

g. Limitations on the concentration resulting from radioactive material, other than noble gases, released in gaseous effluents to unrestricted areas conforming to ten times 10 CFR 20.1001 20.2401 , Appendix B, Table 2, Column 1. For noble gases. the concentration shall be limited to five times 10 CFR 20.1001 20.2401, Appendix B, Table 2, Column 1.

5.0 Page 8 Amendment No. 297

5.0 ADMINISTRATIVE CONTROLS 5.16 Radiological Effluents and Environmental Monitoring Programs 5.16.1 Radioactive Effluent Controls Program (continued)

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents to unrestricted areas conforming to Appendix I to 10 CFR Part 50.

i. Limitations on the annual and quarterly doses to an individual beyond the site boundary from tritium, and all radionuclides in particulate form with half lives greater than 8 days in gaseous effluents released to unrestricted areas conforming to Appendix I to 10 CFR Part 50.

j. Limitations on the annual dose or dose commitment to an individual beyond the site boundary due to releases or radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

5.16.2 Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of tho effluent monitoring program and modeling of environmental exposure path'.Nays. The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

a. Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance v;ith the methodology and parameters in the ODCM.

b. A Land Use Census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitoring program are made if required by the results of this census.

c. Participation in an Interlaboratory Comparison Program to ensure that independent chocks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

5.0 Page 9 Amendment No. 297

5.0 ADMINISTRATIVE CONTROLS 5.17 Offsite Dose Calculation Manual (QDCM)

Changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained as required by the Quality Assurance Program. This documentation shall contain:

1. Sufficient information to support the change together \"'ith the appropriate analyses or e\taluations justifying the shange(s) and

2. A determination that the change \'Viii maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

b. Shall become effective after review and acceptance by the Plant Operations Review Committee and the approval of the plant manager.

s. Temporary changes to the ODCM may be made in accordance with Technical Specification 5.8.2.

d. Shall be submitted to the Nuclear Regulatory Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in 'A'hish any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed and shall indicate the date (e.g., month/year) the change was implemented.

5.18 DELETED 5.19 DELETED 5.0 Page 10 Amendment No. 297

5.0 ADMINISTRATIVE CONTROLS 5.20 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases v.'ithout prior NRC approval provided the changes do not require either of the foiiO't"ling:

1. A change in the TS incorporated in the license or

2. A change to the FSAR, as updated or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR, as updated.

d. Proposed changes that meet the criteria of 5.20.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented 'Nithout prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71 (e).

5.21 DELETED 5.22 DELETED 5.23 DELETED 5.24 DELETED 6.0 DELETED 5.0 Page 11 Amendment No. 297

ATTACHMENT 2 Fort Calhoun Station, Unit No. 1 Renewed Facility License No. DPR-40 "Clean" 10 CFR Part 50 License and Permanently Defueled Technical Specifications Pages

OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT 1 RENEWED FACILITY LICENSE NO. DPR-40

1. The U.S. Nuclear Regulatory Commission (the Commission) having previously made the findings set forth in License No. DPR-40 issued August 9, 1973, has now found that:

A. The application to renew License No. DPR-40 filed by Omaha Public Power District (the licensee) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter 1, and all required notifications to other agencies or bodies have been duly made; B DELETED C. The facility will be maintained in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; D. There is reasonable assurance: (1) that the activities authorized by this renewed license can be conducted without endangering the health and safety of the public, and (2) that such activities will be conducted in compliance with the rules and regulations of the Commission; E. Omaha Public Power District is technically qualified and financially qualified to engage in the activities authorized by this renewed license in accordance with the rules and regulations of the Commission; F. Omaha Public Power District has satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements,"

of the Commission's regulations; G. The issuance of this renewed license will not be inimical to the common defense and security or to the health and safety of the public; H. After weighing the environmental, economic, technical, and other benefits of the facility against environmental costs and considering available alternatives, the Commission concludes that the issuance of Renewed License No. DPR-40 is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied; and I. The receipt, possession, and use of source, byproduct, and special nuclear material as authorized by the renewed license will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40, and 70, including but not necessarily limited to 10 CFR Sections 30.33, 40.32, 70.23 and 70.31 .

DPR-40 Amendment No. ###

2. On the basis of the forgoing findings regarding this facility, Facility Operating License No.

DPR-40, issued August 9, 1973, is superseded by Renewed Facility License No. DPR-40, which is hereby issued to the Omaha Public Power District, to read as follows:

A. This renewed license applies to the Fort Calhoun Station, Unit 1, a pressurized water nuclear reactor and associated equipment (the facility), which is owned by the Omaha Public Power District. The facility is located in Washington County, Nebraska, and is described in the Final Safety Analysis Report as supplemented, amended, and updated and the Environmental Report as supplemented and amended.

B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses Omaha Public Power District:

( 1) Pursuant to Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess and use the facility as required for fuel storage at the designated location in Washington County, Nebraska in accordance with the procedures and limitations set forth in this renewed license; (2) Pursuant to the Act and 10 CFR Parts 40 and 70, to possess at any time special nuclear material that was used as reactor fuel, in accordance with the limitations for storage, as described in the Final Safety Analysis Report, as supplemented, amended, and updated; (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, or sealed sources for radiation monitoring equipment calibration; and to possess any byproduct, source and special nuclear material as sealed neutron sources previously used for reactor startup and reactor instrumentation; and fission detectors; (4) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or instrument calibration or when associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by operation of the facility.

DPR-40 Amendment No. 297

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

A. DELETED B. Technical Specifications The Permanently Defueled Technical Specifications contained in Appendix A, as revised through Amendment No. ###, are hereby incorporated in the license. Omaha Public Power District shall maintain the facility in accordance with the Permanently Defueled Technical Specifications.

C. Security and Safeguards Contingency Plans The Omaha Public Power District shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Fort Calhoun Station Security Plan, Training and Qualification Plan, Safeguards Contingency Plan," submitted by letter dated May 19, 2006.

4. This license is effective as of the date of issuance and authorizes ownership and possession of Fort Calhoun Station until the Commission notifies the licensee in writing that the license is terminated.

FOR THE NUCLEAR REGULATORY COMMISSION Original Signed by:

J.E. Dyer J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments: 1. Appendix A - Permanently Defueled Technical Specifications

2. Appendix B - Deleted Date of Issuance: November 4, 2003 DPR-40 Amendment No. ###

APPENDIX A TO RENEWED FACILITY LICENSE NO. DPR-40 PERMENENTLY DEFUELED TECHNICAL SPECIFICATIONS FOR THE FORT CALHOUN STATION UNIT NO. 1 OMAHA PUBLIC POWER DISCTRICT DOCKET NO. 50-285 DPR-40

PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS TABLE OF CONTENTS 4.0 DESIGN FEATURES 4.1 Site 4.3 Fuel Storage 5.0 ADMINISTRATIVE CONTROLS 5.11 Radiation Protection Program TOC- Page 1 Amendment No. ###

DPR-40

PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS Sections 1.0 through 3.17 and Definitions have been deleted in their entirety.

4.0 DESIGN FEATURES 4.1 Site The site for Fort Calhoun Station Unit No. 1 is in Washington County, Nebraska, on the west bank of the Missouri River and approximately nineteen miles north, northwest of the city of Omaha, Nebraska.

4.2 DELETED 4.3 Fuel Storage Spent Fuel shall not be stored in the spent fuel pool.

4.0 - Page 1 Amendment No. ###

DPR-40

Sections 5.1 through 5.10 have been deleted in their entirety.

5.0 ADMINISTRATIVE CONTROLS 5.11 Radiation Protection Program Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

5.11.1 In lieu of the "control device" required by paragraph 20.1601 (a) of 10 CFR Part 20, and as an alternative method allowed under§ 20.1601(c), each high radiation area (as defined in§ 20.1601) in which the intensity of radiation is 1000 mrem/hr or less shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by required issuance of a Radiation Work Permit.* Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.

c. An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Manager-Radiation Protection (MRP) in the Radiation Work Permit.

• Radiation Protection personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they comply with approved radiation protection procedures for entry into high radiation areas .

5.0- Page 1 Amendment No. ###

DPR-40

5.0 ADMINISTRATIVE CONTROLS 5.11 Radiation Protection Program (continued) 5.11.2 The requirements of 5.11.1, above, shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem/hr** but less than 500 rads/hr*** (Locked High Radiation Area). In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the MRP (or designee) with the following exception:

a. In lieu of the above, for accessible localized Locked High Radiation Areas located in large areas such as containment, where no lockable enclosure exists in the immediate vicinity to control access to the Locked High Radiation Area and no such enclosure can be readily constructed, then the Locked High Radiation Area shall be:

i. roped off such that an individual at the rope boundary is exposed to 1000 mrem/hr or less, ii conspicuously posted, and iii a flashing light shall be activated as a warning device.

• • At 30 centimeters (12 inches) from the radiation source or from any surface penetrated by the radiation .

• • • At 1 meter from the radiation source or from any surface penetrated by the radiation.

Sections 5.12 through 5.24 have been deleted in their entirety.

6.0 DELETED 5.0- Page 2 Amendment No. ###

DPR-40