LIC-04-0117, License Amendment Request, Revisions of Technical Specifications Table 1-1 and Section 4.0.

License Amendment Request, Revisions of Technical Specifications Table 1-1 and Section 4.0.

ML043290064
Person / Time
Site:	Fort Calhoun
Issue date:	11/23/2004
From:	Bannister D Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LIC-04-0117
Download: ML043290064 (32)

Text

Omaha Public Power District 444 South 16th Street Mall Omaha NE 68102-2247 November 23,2004 LIC-04-0 117 U.S. Nuclear Regulatory Commission ATTN.: Document Control Desk Washington, DC 20555-0001

Reference:

Docket No. 50-285

SUBJECT:

Fort Calhoun Station Unit No. 1 License Amendment Request, Revisions of Technical Specifications Table 1-1 and Section 4.0 Pursuant to 10 CFR 50.90, Omaha Public Power District (OPPD) hereby transmits an application for amendment to the Fort Calhoun Station Unit 1 (FCS) Operating License.

The proposed amendment (1) revises the descriptive wording of Technical Specifications Table 1-1, RPS Limiting Safety System Settings, for the Reactor Trip setpoint for Low Steam Generator Water Level to relocate unnecessary detail, and (2) converts Technical Specifications Section 4.0, Design Features, to the content of NUREG-1432, Revision 3. Standard Technical Specifications -for Combustion Enaineerinn Plants. These changes will be needed to support operation of FCS after major components (steam generators, pressurizer, and reactor vessel head) are replaced in 2006.

The proposed Technical Specification changes have been evaluated in accordance with 10 CFR 50.91(a)(l) using criteria in 10 CFR 50.92(c); it has been determined that these changes involve no significant hazards considerations. The bases for these determinations, information supporting the change, a no significant hazards consideration, and an environmental consideration are included in the attached submittal.

Attachment 1 provides the Description of Changes and Justification. Attachment 2 provides the No Significant Hazards Evaluation and the technical bases for this requested change to the Technical Specifications. Attachment 3 contains the marked-up pages and Attachment 4 contains the clean version reflecting the requested Technical Specification and Basis changes.

OPPD requests approval of the proposed amendment by November 30, 2005. Although the proposed changes are needed to support operation after major component replacements in 2006, these changes are considered to be administrative and consistent with the current plant configuration and licensing basis; they can therefore be implemented before 2006. OPPD requests a 60 day implementation period. No commitments are made to the NRC in this letter.

Employment with Equal Opportunity 4171

U. S. Nuclear Regulatory Commission LIC-04-0117 Page 2 I declare under penalty of perjury that the forgoing is true and correct. (Executed on November 23,2004.)

If you have any questions or require additional information, please contact Tom Matthews of my staff at 402-533-6938.

Sincerely, V

D. J. Bannister Manager - Fort Calhoun Station DJBRLJlrlj Attachments:

1. Description of Changes and Justification

2. Fort Calhoun Station's Evaluation

3. Markup of Technical Specification Pages

4. Proposed Technical Specifications (clean) cc: Division Administrator - Public Health Assurance, State of Nebraska

LIC-04-0 117 Page 1 ATTACHMENT 1 Description of Changes and Justification For Revisions of Technical Specifications Table 1-1 and Section 4.0 TABLE 1: LESS RESTRICTIVE REQUIREMENTS - REMOVAL OF DETAIL Change Affected Summary of Change New Change Characterization No. FCS Location Control Technical SPecification 1 1.o FCS TS 1.0 Table 1-1, Item USAR 10 CFR Relocation of Table 1-1 3, Trip Setpoints description 50.59 descriptive Item 3 is revised to relocate the information statement (Top of feedwater which does not ring; 4 10 below normal meet 10 CFR water level). This phrase is 50.36 criteria.

descriptive information which does not meet 10 CFR 50.36 criteria. This information for the replacement steam generators will be moved to the USAR to relocate unnecessary detail from the FCS Tech Specs under the change controls of 10 CFR 50.59. This information is not contained in the

omparable table for Standard Technical Specifications (STS)

NUREG- 1432.

2 4.2 FCS Section 4.2, USAR 10 CFR Relocation of Containment Design 50.59 descriptive Features, is not included in information the STS because these which does not Features are addressed in the meet 10 CFR USAR. 50.36 criteria.

3 4.3.1 & 4.3.3 FCS Sections 4.3.1, Reactor USAR 10 CFR Relocation of Coolant System, and 4.3.3, 50.59 descriptive Emergency Core Cooling, information are not included in the STS which does not because these features do not meet 10 CFR

LIC-04-0117 Page 2 neet 10 CFR 50.36(~)(4) 50.36 criteria.

riteria and are contained in he USAR.

4 4.4.1 TCS TS 4.4.1 states that, USAR 10 CFR Relocation of Thenew unirradiated fuel 50.59 descriptive

)undies will normally be information

tored in the dry new fuel which does not

torage rack with an effective meet 10 CFR nultiplication factor of less 50.36 criteria.

han 0.9. This information is lot included in the STS 3ecause these features are iddressed in the USAR.

4.4.1 X S TS 4.4.1 states that, USAR 10 CFR Relocation of The new fuel storage rack is 50.59 descriptive located 18-9 above the information nain floor of Room 25A which does not which provides for adequate meet 10 CFR irainage and precludes 50.36 criteria..

flooding of the new fuel storage rack. This information is not included in the STS because these features are addressed in the USAR.

4.4.1 FCS TS 4.4.1 states that, USAR 10 CFR Removal of New fuel may also be 50.59 descriptive stored in shipping containers information or in the spent fuel pool which is already racks which have a located in Tech.

maximum effective Spec Bases multiplication factor of 0.95 2.8.3(1) and with Fort Calhoun Type C contained in fuel and unborated water. USAR Reference This statement is not relevant 9.5-6, EA-FC-to the discussion of the new 96-001, Rev 0 fuel storage racks. and does not Therefore, it is not included meet 10 CFR in the STS. ~~

50.36 criteria.

7 4.4.1 FCS TS 4.4.1 states that, USAR 10 CFR Relocation of The new fuel storage racks 50.59 descriptive are designed as a Class I information structure. This statement is which does not not included in the STS meet 10 CFR because it is contained in the 50.36 criteria.

USAR.

LIC-04-0117 Page 3 4.4.2 FCS TS 4.4.2 states that, USAR 10 CFR Relocation of

• Irradiatedfuel bundles will 50.59 descriptive be stored prior to off-site information shipment in the stainless which does not steel lined spent fuel pool. meet 10 CFR The spent fuel pool is 50.36 criteria.

normally filled with borated water with a concentration of at least the refueling boron concentration. T h s statement is not included in the STS because it is contained in the USAR 4.4.2 FCS TS 4.4.2 states that, USAR 10 CFR Relocation of The spent fuel racks are 50.59 descriptive designed as a Class I information structure. This statement is which does not not included in the STS meet 10 CFR because it is contained in the 50.36 criteria.

USAR 4.4.2 FCS TS 4.4.2 states that, USAR 10 CFR Relocation of Normally the spent fuel 50.59 descriptive pool cooling system will information maintain the bulk water which does not temperature of the pool meet 10 CFR below 120°F. Under other 50.36 criteria.

conditions of fuel discharge, the fuel pool water temperature is maintained below 140°F. This statement is not included in the STS because it is contained in the USAR 4.4.2 FCS TS 4.4.2 states that, USAR 10 CFR Removal of The spent fuel racks are 50.59 descriptive designed and will be information maintained such that the which is already calculated effective located in Tech.

multiplication factor is no Spec Bases greater than 0.95 (including 2.8.3(1) and all known uncertainties) contained in assuming the pool is flooded USAR Reference with unborated water. The 9.5-6, EA-FC-racks are divided into 2 96-001, Rev 0 regions. Storage in Region 1 and does not and Region 2 of the spent meet 10 CFR

LIC-04-0117 Page 4 fuel racks shall be restricted 50.36 criteria.

to fuel assemblies having initial enrichment less than or equal to 4.5 weight percent of U-235. Region 1 and 2 cells are surrounded by Boral. Acceptance criteria for fuel storage in Regions 1 and 2 are delineated in Section 2.8 of these Technical Specifications.

These statements are not included in the STS because they are contained in the USAR and in some cases are restated bv the STSs.

12 4.5 FCS Section 4.5, Seismic USAR 10 CFR Relocation of Design for Class I Systems, 50.59 descriptive is not included in the STS information because these features are which does not addressed in the USAR. meet 10 CFR 50.36 criteria.

LIC-04-0117 Page 5 TABLE 2: ADMINISTRATIVE CHANGES TO STANDARD TECHNICAL SPECIFICATIONS Affected FCS Technical Summary of Change*

Specification 4.3.1.l.c and FCS Technical Specifications4.3.1.1.c and 4.3.1.1 .d are revised from the STS to 4.3.1.1.d include Region 2 and Region l, respectively, to assure consistent nomenclature of the high and low density fuel storage racks to be consistent with the nomenclature presented on existing FCS TS Figure 2-10.

4.3.1.e FCS TS 4.3.1.e and 4.3.1.f are revised from the STS wording of unacceptable and range to unacceptable domain to be consistent with the wording domain 4.3.1 .f used in existing FCS TS Figure 2-10.

4.3.1.e FCS TS 4.3.1.e is revised from the STS wording as follows: 1) Figure[3.7.18.1]

is changed to Figure 2-10 for Region 2 Unrestricted or Peripheral Cells, and 2) or in peripheral cells if acceptable for peripheral storage has been added following fuel storage racks(s), to address the additional restrictions imposed on loading of peripheral cells by existing FCS TS Figure 2-10.

4.3.1.f FCS TS 4.3.1 .f is revised from the STS wording Figure[3.7.18.1] to Figure 2-10 for Region 2 Unrestricted or Peripheral Cells to address the additional restrictions imposed on loading of peripheral cells by existing FCS TS Figure 2-

10. Following will be stored in, Region 1 has been added to specify: 1) where the spent fuel falling within this burnup domain will be stored and 2) incorporate the reference prescribing the storage requirement.

4.3.1.2.b FCS TS 4.3.1.2.b is revised from the STS wording of keff d . 9 8 if fully flooded with unborated water, which includes an allowance for uncertainties as described in [Section 9.1 of the FSAR] to keff 10.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Reference (2). because the licensed FCS limit of the new fuel storage racks is 0.95.

4.3.1.2 FCS TS 4.3.1.2 is revised from the STS 4.3.1.2.c wording of keff 10.98 if moderated by aqueous foam, which includes an allowance for uncertainties as described in [Section 9.1 of the FSAR]. to the deletion of an aqueous foam related specification because FCS has no system or capability of injecting foam into the new fuel storage racks.

• - The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certain examples in Reference 10.2 (48 FR 14864) of amendments that are considered not likely to involve significant hazards consideration. The above proposed changes are consistent with example (vii) which relates to a change to make a license conform to changes in the regulations, where the license change results in very minor changes to facility operations clearly in keeping with the regulations.

LIC-04-0117 Page 1 ATTACHMENT 2 Fort Calhoun Stations Evaluation For Revisions of Technical Specifications Table 1-1 and Section 4.0 1.O DESCRIPTION

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory RequirementdCriteria

7.0 REFERENCES

LIC-04-0117 Page 2 1.O DESCRIPTION This letter is a request to amend Operating License DPR-40 for Fort Calhoun Station Unit No. 1.

The proposed changes would revise the operating License through administrative changes to the Fort Calhoun Station (FCS) Technical Specifications, specifically: (1) revises the descriptive wording of TS Table 1-1, W S Limiting Safety System Settings, for the Reactor Trip setpoint for Low Steam Generator Water Level to relocate unnecessary detail, and (2) converts Section 4.0, Design Features, to the Design Features content of NUREG-1432, Revision 3, Standard Technical Specifications for Combustion Engineering Plants.

2.0 PROPOSED CHANGE

The proposed changes will be needed to support operation of FCS after major components (steam generators, pressurizer, and reactor vessel head) are replaced in 2006.

(1) Deletion of extra setpoint description from Item 3 of Technical Specifications Table 1-1. W S Limiting Safety System Settings, The existing Trip Setpoint description for Item 3, Low Steam Generator Water Level Reactor Trip, contains a parenthetical descriptive phrase below the main setpoint description 31.2% of Scale. OPPD proposes to relocate the parenthetical phrase (Top of feedwater ring: 4 10 below normal water level) to the USAR.

The low water level trip setpoint for each replacement steam generator will remain at 3 1.2%

of scale; however, the corresponding physical location will not be consistent with the existing parenthetical description due to sizing and geometry differences.

( 2 ) Conversion of Technical Specifications Section 4.0, Design Features, to the Design Features content of NUREG- 1432. Revision 3, Standard Technical Svecifications for Combustion Engineering Plants The existing Technical Specification Section 4.0, Design Features, is being changed to be identical to the Design Features content of NUREG-1432, Revision 3, Standard Technical Svecifications for Combustion Engineering Plants, through addition of standardized descriptions of the new and spent fuel storage facilities, and removal of descriptions of the containment, reactor coolant system, reactor core and control, emergency core cooling, and seismic design for Class I systems.

The 2006 component replacements affect the RCS volume value currently in TS 4.3.1; however, rather than revising just that value, OPPD has elected to convert the whole section to Standard Technical Specifications as part of a long term Technical Specification conversion plan. The RCS volume value is not included in STS.

LIC-04-0117 Page 3 Fort Calhoun Station (FCS) proposes to change the Technical Specification (TS) conditions, specifications, and requirements, as described in Attachment 1, Description of Changes and Justification.

In summary, the proposed amendment relocates redundant steam generator setpoint description from the Technical Specifications and updates the existing Technical Specifications Section 4.0, Design Features, with text and content identical to the Standard Technical Specifications.

3.0 BACKGROUND

OPPD plans to replace major components (steam generators, pressurizer, and reactor vessel head) at FCS in 2006, using the provisions of 10 CFR 50.59 to the extent possible. OPPD also has a long term plan to convert the FCS Technical Specifications to Standard Technical Specifications, using a phased approach. Based on a review to determine what Technical Specification changes are needed to support operation of FCS after the major components are replaced, OPPD has determined that the changes requested in this letter are necessary. These changes are considered to be administrative and consistent with the current plant configuration; they can therefore be implemented before 2006. Engineering and safety analyses specific to the new components are ongoing; any hture License Amendment Requests resulting from these analyses will be contained in separate correspondence.

4.0 TECHNICAL ANALYSIS

The proposed amendment affects two separate and distinct sections of the Technical Specifications which have been evaluated for technical significance.

(1) Deletion of extra setpoint description from Item 3 of Technical Specifications Table 1- 1, RPS Limiting Safety System Settings The existing Trip Setpoint description for Item 3, Low Steam Generator Water Level Reactor Trip, contains a parenthetical descriptive phrase below the main setpoint description 3 1.2% of Scale. This parenthetical phrase is (Top of feedwater ring: 4 10 below normal water level). The specific dimension for the distance between the feedwater ring and normal water level is a historical carryover from the initial 1972 Technical Specifications which define the steam generator level water inventory that is sufficient to provide a twelve minute margin before the auxiliary feedwater is required as stated in the existing Technical Specification 1.3(5) bases section. The setpoint analyses for the replacement steam generator, currently in progress, confirm the continued use of the existing Technical Specification Low Steam Generator trip setpoint value of 31.2% of Scale. However, the physical geometry of the replacement steam generators, when installed, will be different than the original steam generators. Elimination of the parenthetical descriptive statement relocates unnecessary and duplicative detail from the Technical Specifications, has no effect on the safety functions or requirement of the reactor protection system, and has no impact upon the USAR accident analysis. Corresponding changes to the Basis and inclusion of the information in the USAR will be appropriately controlled under 50.59 criteria.

LIC-04-0117 Page 4 Because the value/dimension for the normal water level to feedwater ring distance is not currently required by 10 CFR 50.36 criteria nor is the dimension utilized in Standard Technical Specifications, the proposed relocation of the parenthetical statement is acceptable.

(2) Conversion of Technical Specifications Section 4.0, Design Features, to content of NUREG-1432. Revision 3, Standard Technical Specifications for Combustion Engineerinn Plants The existing Technical Specifications Section 4.0, Design Features, format and content is being changed to be identical to the Design Features content of NUREG-1432, Revision 3, Standard Technical Specifications for Combustion Engineering Plants through addition of standardized descriptions of the new and spent fuel storage facilities, and proposed removal of descriptions of the containment, reactor coolant system, reactor core and control, emergency core cooling, and seismic design for Class I systems. An evaluation of each of the less restrictive requirements for removal of detail and administrative changes, tabulated in Attachment I, concludes that these changes are administrative in nature because they follow the form of NRC examples for determining whether a significant hazards consideration exists. These changes do not impact the USAR accident analyses.

The design of the facility is required to be described in the Updated Safety Analysis Report (USAR) by 10 CFR 50.34. In addition, the quality assurance (QA) requirements of Appendix B to 10 CFR Part 50 require that the plant design be documented in controlled procedures and drawings, and maintained in accordance with an NRC-approved QA plan (referenced in the USAR). Controls are specified in 10 CFR 45.59 for changing the facility as described in the USAR. Removing details of system design from the custom technical specifications is acceptable because this information will be adequately controlled in the USAR.

5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration Omaha Public Power District has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

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

Response: No.

The proposed changes are not related to an initiator of any previously evaluated accident. The proposed changes revise descriptive information only, and will not prevent safety systems from performing their accident mitigation function as assumed in the safety analysis.

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

LIC-04-0117 Page 5

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

Response: No.

The proposed changes only relocate descriptive information in the Technical Specifications to the USAR. Modifications will not be made to existing equipment nor will any new or different types of equipment be installed. The proposed changes to the Technical Specifications will not alter assumptions made in safety analysis and licensing bases.

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

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

Response: No.

The proposed administrative changes only relocate descriptive information in the FCS Technical Specifications to the USAR, and have no effect on safety margins.

Therefore, this technical specification change does not involve a significant reduction in the margin of safety.

Based on the above, Omaha Public Power District 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.

5.1 Applicable Regulatory Requirements/Criteria The technical analysis of the proposed amendment addressed the two separate and distinct sections of the Technical Specifications. Due to their administrative nature, the proposed changes move information from the Technical Specifications and present the same or equivalent information in the USAR. This change in information location does not change or affect commitments to Fort Calhoun Station design criteria presented in the FCS USAR Appendix G, the USAR accident analyses, approved methodologies, Regulatory Guides, or NUREGs.

For the changes listed in Attachment I, Table 1: Less Restrictive Requirement -

Removal of Detail, specific technical information which does not meet the criteria of 10 CFR 50.36 for inclusion in the Technical Specifications is being moved to the USAR under the guidance of 50.59. Changes listed in Table 2: Administrative Changes meet the Commissions guidance example concerning the application of the standards for determining whether a significant hazards consideration exists. The Description of

LIC-04-0117 Page 6 Changes and Justification," for the administrative changes for the proposed changes cites the appropriate NRC provided example.

The proposed changes do not alter, degrade, or prevent actions described or assumed in any accident analysis. The proposed changes will not change any assumptions previously made in evaluating radiological consequences or affect any fission product barriers, nor do they increase any challenges to safety systems. Therefore, the proposed change does not increase or have any impact on the consequences of events described and evaluated in Chapter 14 of the Fort Calhoun USAR.

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

6.0 ENVIRONMENTAL CONSIDERATION

Based on the above considerations, the proposed amendment does not involve and will not result in a condition which significantly alters the impact of Fort Calhoun Station on the environment. Thus, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR Part 5 1.22(~)(9),and, pursuant to 10 CFR Part 5 1.22(b), no environmental assessment need be prepared.

7.0 REFERENCES

7.1 NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants" 7.2 Standards for Determining Whether License Amendments Involve No Significant Hazards Considerations, 48 FR 14864

LIC-04-0117 Page 1 Attachment 3 Markup of Technical Specification Pages

TECHNICAL S PECIFICAT10N I

TABLE OF CONTENTS (Continued) 2.13 DELETED 2.14 Engineered Safety Features System Initiation Instrumentation Settings 2.15 Instrumentationand Control Systems 2.16 River Level 2.17 Miscellaneous Radioactive Material Sources 2.18 DELETED 2.19 DELETED 2.20 Steam Generator Coolant Radioactivity 2.21 Post-Accident Monitoring Instrumentation 2.22 Toxic Gas Monitors 3.0 SURVEILLANCE REQUIREMENTS 3.1 Instrumentation and Control 3.2 Equipment and Sampling Tests 3.3 Reactor Coolant System and Other Components Subject to ASME XI Boiler and Pressure Vessel Code Inspection and Testing Surveillance 3.4 Reactor Coolant System Integrity Testing 3.5 Containment Test 3.6 Safety Injection and Containment Cooling Systems Tests 3.7 Emergency Power System Periodic Tests 3.8 Main Steam Isolation Valves 3.9 Auxiliary Feedwater System 3.10 Reactor Core Parameters 3.11 DELETED 3.12 Radioactive Waste Disposal System 3.13 Radioactive Material Sources Surveillance 3.14 DELETED 3.15 DELETED 3.16 Residual Heat Removal System Integrity Testing 3.17 Steam Generator Tubes 4.0 DESIGN FEATURES 4.1 Site 4.2 4.1.2 - - f 4.3 TOC - Page 2 Amendment No. 4 '!- ,

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TECHNICAL S PECIFICAT1ON TABLE OF CONTENTS (Continued) 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.2 Organization 5.3 Facility Staff Qualifications 5.4 Training 5.5 Review and Audit 5.5.1 Plant Review Committee (PRC) 5.5.2 Safety Audit and Review Committee (SARC) 5.6 Reportable Event Action 5.7 Safety Limit Violation 5.8 Procedures 5.9 Reporting Requirements 5.9.1 Routine Reports 5.9.2 Reportable Events 5.9.3 Special Reports 5.9.4 Unique Reporting Requirements 5.9.5 Core Operating Limits Report 5.9.6 RCS Pressure-TemperatureLimits Report (PTLR) 5.10 Record Retention 5.11 Radiation Protection Program 5.12 DELETED 5.13 Secondary Water Chemistry 5.14 Systems Integrity 5.15 Post-Accident Radiological Sampling and Monitoring 5.16 Radiological Effluents and EnvironmentalMonitoring Programs 5.16.1 Radioactive Effluent Controls Program 5.16.2 Radiological Environmental Monitoring Program 5.17 Offsite Dose Calculation Manual (ODCM) 5.18 Process Control Program (PCP) 5.19 Containment Leakage Rate Testing Program 5.20 Technical Specification (TS) Bases Control Program 5.21 Containment Tendon Testing Program 6.0 INTERIM SPECIAL TECHNICAL SPECIFICATIONS 6.1 DELETED 6.2 DELETED 6.3 DELETED 6.4 DELETED

TECHNICAL SPECIFICATIONS TABLE 1-1 RPS LIMITING SAFETY SYSTEM SETTINGS No. Reactor Trip Trip Setpoints 1 High Power Level (A) -~109.0%of Rated Power 4-Pump Operation 2 Low Reactor Coolant Flow (B)(F) 4-Pump Operation ->95% of 4 Pump Flow 3 Low Steam Generator Water Level 31.2% of Scale F

Low Steam Generator Pressure (C) -~ 5 0 0psia High Pressurizer Pressure -~ 2 4 0 0psia Thermal Margin/Low Pressure (B)(F) 1750 psia to 2400 psia (depending on the reactor coolant temperature as shown in the Thermal Margin/Low Pressure 4 Pump Operation Figure provided in the COLR) 7 High Containment Pressure (D) -

<5 psig 8 Axial Power Distribution (E) (as shown in the Axial Power Distribution for 4 Pump Operation Figure provided in the COLR) 9 Steam Generator Differential Pressure -4 3 5 psid 1.0 - Page 11 Amendment No. =7,77,92,??? ,?79, w

TECHNICAL SPECIFICATI0NS 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, as defined in 10 CFR Part 100, Section 100.3(a), consists of approximately 1242 acres. The exclusion area boundary extent includes approximately 660 acres in Washington County, Nebraska, owned by the Omaha Public Power District (OPPD), and 582 acres in Harrison County, Iowa, on the east bank of the river directly opposite the facility, on which the District retains perpetual easement rights. The minimum exclusion area boundary point is located approximately at the 187.0 degree radial from the outer wall of the containment building and at a distance of 910 meters.

TECHNICAL SPECIFICATI0NS 4.0 DESIGN FEATURES (Continued) 4.0 - Page 2

TECHNICAL SPECIFICATI0NS 4.0 DESIGN FEATURES (Continued)

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LIC-04-0117 Page 1 Attachment 4 Clean-Typed Technical Specification Pages

TECHNI CAL SPECIFICAT1ON TABLE OF CONTENTS (Continued) 2.13 DELETED 2.14 Engineered Safety Features System Initiation InstrumentationSettings 2.15 Instrumentation and Control Systems 2.16 River Level 2.17 Miscellaneous Radioactive Material Sources 2.18 DELETED 2.19 DELETED 2.20 Steam Generator Coolant Radioactivity 2.21 Post-Accident Monitoring Instrumentation 2.22 Toxic Gas Monitors 3.0 SURVEILLANCE REQUIREMENTS 3.1 Instrumentation and Control 3.2 Equipment and Sampling Tests 3.3 Reactor Coolant System and Other Components Subject to ASME XI Boiler and Pressure Vessel Code Inspection and Testing Surveillance 3.4 Reactor Coolant System Integrity Testing 3.5 Containment Test 3.6 Safety Injection and Containment Cooling Systems Tests 3.7 Emergency Power System Periodic Tests 3.8 Main Steam Isolation Valves 3.9 Auxiliary Feedwater System 3.10 Reactor Core Parameters 3.11 DELETED 3.12 Radioactive Waste Disposal System 3.13 Radioactive Material Sources Surveillance 3.14 DELETED 3.15 DELETED 3.16 Residual Heat Removal System Integrity Testing 3.17 Steam Generator Tubes 4.0 DESIGN FEATURES 4.1 Site 4.2 Reactor Core 4.3 Fuel Storage TOC - Page 2 Amendment No. 1!,2722,-

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TECHNICAL SPECIFICAT10N TABLE OF CONTENTS (Continued) 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.2 Organization 5.3 Facility Staff Qualifications 5.4 Training 5.5 Review and Audit 5.5.1 Plant Review Committee (PRC) 5.5.2 Safety Audit and Review Committee (SARC) 5.6 Reportable Event Action 5.7 Safety Limit Violation 5.8 Procedures 5.9 Reporting Requirements 5.9.1 Routine Reports 5.9.2 Reportable Events 5.9.3 Special Reports 5.9.4 Unique Reporting Requirements 5.9.5 Core Operating Limits Report 5.9.6 RCS Pressure-TemperatureLimits Report (PTLR) 5.10 Record Retention 5.1 1 Radiation Protection Program 5.12 DELETED 5.13 Secondary Water Chemistry 5.14 Systems Integrity 5.15 Post-Accident Radiological Sampling and Monitoring 5.16 Radiological Effluents and Environmental Monitoring Programs 5.16.1 Radioactive Effluent Controls Program 5.1 6.2 Radiological EnvironmentalMonitoring Program 5.17 Offsite Dose Calculation Manual (ODCM) 5.18 Process Control Program (PCP) 5.19 Containment Leakage Rate Testing Program 5.20 Technical Specification (TS) Bases Control Program 5.21 Containment Tendon Testing Program 6.0 INTERIM SPECIAL TECHNICAL SPECIFICATIONS 6.1 DELETED 6.2 DELETED 6.3 DELETED 6.4 DELETED TOC - Page 3 Amendment No.

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TECHNICAL SPECIFICATI0NS TABLE 1-1 RPS LIMITING SAFETY SYSTEM SETTINGS No. Reactor Trip Trip Setpoints 1 High Power Level (A) -

409.0% of Rated Power 4-Pump Operation 2 Low Reactor Coolant Flow (B)(F) 4-Pump Operation >95% of 4 Pump Flow Low Steam Generator Water Level 31.2% of Scale Low Steam Generator Pressure (C) -

>500 psia High Pressurizer Pressure -

~ 2 4 0 0psia Thermal Margin/Low Pressure (B)(F) 1750 psia to 2400 psia (depending on the reactor coolant temperature as shown in the Thermal Margin/Low Pressure 4 Pump Operation Figure provided in the COLR) 7 High Containment Pressure (D) -

<5 psig 8 Axial Power Distribution (E) (as shown in the Axial Power Distribution for 4 Pump Operation Figure provided in the COLR) 9 Steam Generator Differential Pressure (1 35 psid 1.0-Page11 Amendment No. &3&47,77,0?,?4 1,? 79, 24-0

TECHN1CAL SPECIFICATI0NS 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, as defined in 10 CFR Part 100, Section 100.3(a), consists of approximately 1242 acres. The exclusion area boundary extent includes approximately 660 acres in Washington County, Nebraska, owned by the Omaha Public Power District (OPPD), and 582 acres in Harrison County, Iowa, on the east bank of the river directly opposite the facility, on which the District retains perpetual easement rights. The minimum exclusion area boundary point is located approximately at the 187.0 degree radial from the outer wall of the containment building and at a distance of 910 meters.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 133 fuel assemblies. Each assembly shall consist of a matrix of Zircalloy or ZIRLO fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Control Element Assemblies The reactor core shall contain 49 control element assemblies (CEAs). The control material shall be silver indium cadmium, boron carbide, or hafnium metal as approved by the NRC.

4.3 Fuel Storaqe 4.3.1 Criticality 4.3.1 .I The spent fuel storage racks are designed and shall be maintained with:

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

b. bfi ~ 0 . 9 5if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.5 of the USAR, 4.0 - Page 1 Amendment No. 36

TECHNICAL SPEClFICAT1ONS 4.0 DESIGN FEATURES (Continued)

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

d. A nominal 9.8 inches (East-West) by 10.3 inches (North South) center to center distances between fuel assemblies placed in Region I, the low density fuel storage racks,

e. New or partially spent fuel assemblies with a discharge burnup in the acceptable domain of Figure 2-10 for Region 2 Unrestricted or Peripheral Cells may be allowed unrestricted storage in either fuel storage racks or peripheral cells if acceptable for peripheral cell storage, and

f. New or partially spent fuel assemblies with a discharge burnup in the unacceptable domain of Figure 2-10 for Region 2 Unrestricted or Peripheral Cells will be stored in Region 1 in compliance with the Reference (I).

4.3.1.2 The new fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent,

b. keff c 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Reference (2).

C. A nominal 16 inch center to center distance between fuel assemblies placed in the storage racks.

References:

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

(2) Ft. Calhoun USAR, Reference 9.5-1 4.0 - Page 2 Amendment No.

TECHNICAL SPECIFICAT1ONS 4.0 DESIGN FEATURES (Continued) 4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 23 ft.

4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than I083 fuel assemblies.

4.0 - Page 3 Amendment No.