LIC-05-0098, License Amendment Request (Lar), Deletion of Shutdown Margin Surveillance Requirement and Miscellaneous Administrative Changes.

License Amendment Request (Lar), Deletion of Shutdown Margin Surveillance Requirement and Miscellaneous Administrative Changes.

ML052240078
Person / Time
Site:	Fort Calhoun
Issue date:	08/11/2005
From:	Ridenoure R Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LIC-05-0098
Download: ML052240078 (50)

Text

Omaha Public Power District 444 South 16th Street Mall Omaha NE 68102-2247 LIC-05-0098 August 11,2005 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 (LAR),

"Deletion of Shutdown Margin Surveillance Requirement and Miscellaneous Administrative Changes" Pursuant to Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), the Omaha Public Power District (OPPD) requests to amend Operating License No. DPR-40 for Fort Calhoun Station, Unit No. 1 (FCS). The proposed amendment includes various changes to the Technical Specifications (TS). Specifically, OPPD seeks to delete the surveillance requirement (SR) of TS 2.10.2(9)b(iii) to verify shutdown margin every 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift during low power physics testing. This change will make TS 2.10.2(9)b more consistent with SR 3.1.7 of NUREG-1432, Standard Technical Speczjications - Combustion Engineering Plants, Revision 3.

The Containment Structural Tests Report of TS 5 . 9 . 3 ~is proposed for deletion. Amendment 216 deleted TS 3.5(5), which required submittal of the TS 5 . 9 . 3 ~report. The deletion of the report and the remaining changes described in Attachment 1 are considered administrative in nature.

As such, these changes are unlikely to involve significant hazards considerations.

Attachment 1 provides a description of the proposed change and confirmation of applicability.

Attachment 2 provides examples of amendments considered unlikely to involve significant hazards considerations. Attachment 3 provides the existing TS pages marked-up with the proposed changes. Attachment 4 provides the proposed TS pages.

OPPD requests NRC approval by March 1, 2006 with a 90-day implementation period. No commitments to the NRC are made in this letter.

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

Employment with Equal Opportunity 4171

U. S. Nuclear Regulatory Commission LIC-05-0098 Page 2 I declare under penalty of perjury that the foregoing is true and correct (Executed on August 11, 2005).

If you should have any questions regarding this submittal, please contact T. C. Matthews at (402) 533-6938.

Sincerely, /-,

/

/Attachments: 1. Omaha Public Power District Evaluation

2. Examples of No Significant Hazards Amendments

3. Markup of Technical Specification Pages

4. Proposed Technical Specification Pages cc: Division Administrator - Public Health Assurance, State of Nebraska

LIC-05-0098 Page 1 Omaha Public Power District Evaluation For Amendment of Operating License

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY SAFETY ANALYSIS 6.0 ENVIRONMENTAL EVALUATION

7.0 REFERENCES

LIC-05-0098 Page 2

1.0 DESCRIPTION

The Omaha Public Power District (OPPD) requests an amendment to Operating License No.

DPR-40 for Fort Calhoun Station, Unit No. 1 (FCS). The proposed amendment includes various changes to the Technical Specifications (TS). OPPD proposes to delete the surveillance requirement (SR) of TS 2.10.2(9)b(iii) to verify shutdown margin (SDM) every 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift during low power physics testing. The SRs that will remain in TS 2.10.2(9)b following the deletion of (iii) adequately verify SDM during low power physics testing and are consistent with similar requirements in NUREG-1432, Standard Technical Specifications-Combustion Engineering Plants, Revision 3 (Reference 7.3).

OPPD also proposes to delete TS 5.9.3c, Containment Structural Tests Report. Amendment 216 (Reference 7.5) deleted TS 3.5(5), which contained requirements for submitting the report from the TSs and moved the contents of TS 3 3 5 ) to the FCS Updated Safety Analysis Report (USAR). Thus, although the report is listed in TS 5.9.3c, the FCS Technical Specifications contain no requirement to submit it. In accordance with paragraph (b)(2)(viii) of 10 CFR 50.55a, conditions indicative of containment structural deterioration or degradation are reported in the Inservice Inspection (ISI) Report prepared in accordance with ASME Section XI, Subsection IWA-6000 requirements. TS 5 . 9 . 3 ~can be deleted since information equivalent to that formerly submitted in accordance with TS 3 3 5 ) is contained in the IS1 report required by 10 CFR 50.55a and TS 5.9.3a. This is considered an administrative change based on the discussion above.

The remaining changes proposed in this license amendment request (LAR) consist of grammatical andlor administrative corrections that are considered unlikely to involve significant hazards considerations as explained below. References 7.2 and 7.4 are amendments to the FCS Technical Specifications that are similar in nature to this one and are precedents for this LAR.

2.0 PROPOSED CHANGE

Change 1 TS 2.10.2(9)b(iii) is proposed for deletion. This Specification currently states that SDM shall be verified every 8-hour shift during low power physics testing. However, Operations Department personnel assigned to shift duties work 12-hour, not 8-hour shifts (the definition of a shift was revised by Amendment 218 (Reference 7.4) to be 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> long). Reactivity control during low power physics testing is provided primarily by trippable control element assemblies (CEAs) whose position is determined every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in accordance with TS 2.10.2(9)b(i)2.

Should the SDM required by TS 2.10.2(9)b(i) be unavailable, TS 2.10.2(9)b(ii) requires boration to be initiated immediately and continued until the SDM required by the Core Operating Limits Report (COLR) is achieved. Thus, the TS 2.10.2(9)b(iii) requirement to verify SDM every 8-hour shift is inaccurate and unnecessary. The SRs that remain in TS 2.10.2(9)b following the deletion of (iii) adequately verify SDM during low power physics testing and are consistent with SR 3.1.7 of NUREG-1432, Revision 3 (Reference 7.3).

LIC-05-0098 Page 3 Change 2 The change proposed for TS 5.9.3c, Special Reports deletes the Containment Structural Tests Report. Amendment 216 deleted TS 3.5(5), which contained requirements for submitting the report and moved the contents of TS 3 3 5 ) to the FCS Updated Safety Analysis Report (USAR).

Thus, although the Containment Structural Tests Report is listed in TS 5.9.3c, there is no TS requirement to submit it. In accordance with paragraph (b)(2)(viii) of 10 CFR 50.55a, conditions indicative of containment structural deterioration or degradation are reported in the Inservice Inspection (ISI) Report prepared in accordance with ASME Section XI, Subsection IWA-6000 requirements. Thus, since equivalent information is contained in the IS1 report required by 10 CFR 50.55a and TS 5.9.3a, the Containment Structural Tests Report should be deleted from TS 5.9.3~. In accordance with FR guidance (48 FR 14864), this change is administrative. [See , Example (i)]

Change 3 Page 3 of the Table of Contents (TOC) is revised to show TSs 5.5, 5.6, 5.9.1 and 5.9.2 as "Not Used." Amendment 228 (Reference 7.2) revised TSs 5.5, 5.6, and 5.9.2 to show they are "Not Used" but did not revise the TOC accordingly. Amendment 231 (Reference 7.1) deleted the remaining routine reports (Monthly Operating Report and Annual Occupational Exposure Report) of TS 5.9.1 and thus the TOC and TS 5.9.1 are revised to show that TS 5.9.1 is "Not Used." In accordance with FR guidance (48 FR 14864), these changes are administrative. [See Attachment 2, Example (i)]

Change 4 The changes proposed for TS 2.1.6 and its Basis improve grammatical structure by inserting appropriate punctuation and revising sentences with missing, incomplete or inappropriate words.

These are administrative changes. In accordance with FR guidance (48 FR 14864), this change is administrative. [See Attachment 2, Example (i)]

Change 5 Table 3-13, Steam Generator Tube Inspections is revised to delete two occurrences of the phrase "Prompt notification to the NRC pursuant to specification 5.6." Amendment 228 (Reference 7.2) deleted TS 5.6, Reportable Event Action because its requirements were redundant to 10 CFR 50.72 and 50.73. Therefore, since the FCS TSs no longer contains TS 5.6, the statements contained in Table 3-13 should be deleted. In accordance with FR guidance (48 FR 14864), this change is administrative. [See Attachment 2, Example (i)]

Change 6 Page 10 of TS 5.0 designated "Not Used" is proposed for deletion and all subsequent TS 5.0 pages are renumbered accordingly. In accordance with FR guidance (48 FR 14864), this change is administrative. [See Attachment 2, Example (i)]

Change 7 The last sentence of TS 5.16.lg is revised to change the period following the words "For noble gases" to a comma. In accordance with FR guidance (48 FR 14864), this change is administrative. [See Attachment 2, Example (i)]

LIC-05-0098 Attachment I Page 4 Change 8 TS 5.21 is revised to correct an error showing Revision 3 of Regulatory Guide 1.35, Insewice Inspection Of Ungrouted Tendons in Prestressed Concrete Containments as issued in 1989. The correct date of Regulatory Guide 1.35, Revision 3 is 1990. In accordance with FR guidance (48 FR 14864), this change is administrative. [See Attachment 2, Example (i)]

In summary, this LAR deletes an unnecessary SR, deletes a report that is redundant to Federal Regulations and proposes other administrative changes consistent with FR guidance (48 FR 14864). The SRs that remain in TS 2.10.2(9)b following the deletion of TS 2.10.2(9)b(iii) are consistent with NUREG-1432 (Reference 7.3) requirements for verifying SDM during low power physics testing. Information concerning containment structural deterioration or degradation is contained in the IS1 report required by TS 5.9.3a and thus TS 5 . 9 . 3 ~(Containment Structural Tests Report) should be deleted. The remaining changes proposed by this LAR correct errors that are clearly administrative in nature as explained above.

3.0 BACKGROUND

Change 1 The purpose of TS 2.10.2(9) is to permit relaxation of existing limiting conditions for operations (LCOs) to allow the performance of certain physics tests to determine CEA worth and shutdown margin. The SDM exemption of TS 2.10.2(9)b provides that a minimum amount of CEA worth is immediately available for reactivity control when physics tests are performed for CEA worth measurement. This special test exemption is required to permit the periodic verification of the actual versus predicted core reactivity condition occurring because of fuel bumup or fuel cycling conditions.

Testing is required prior to initial criticality, after each refueling shutdown, during startup, low power operation, power ascension, and at power operation. The physics tests requirements for reload fuel cycles ensure that the operating characteristics of the core are consistent with design predictions and that the core can be operated as designed.

Physics test procedures are written and approved in accordance with established formats. The procedures include all information necessary to permit a detailed execution of testing required to ensure that the design intent is met. Physics tests are performed in accordance with these procedures, and test results are approved prior to continued power escalation and long term power operation. Examples of physics tests include determination of critical boron concentration, CEA group worths, reactivity coefficients, flux symmetry, and core power distribution. It is acceptable to suspend certain LCOs for physics tests because fuel damage criteria are not exceeded. Even if an accident occurs during physics tests with one or more LCOs suspended, fuel damage criteria are preserved because adequate limits on power distribution and shutdown capability are maintained.

Change 2 Amendment 216 deleted TS 3.5(5) from the FCS Technical Specifications and moved it to the FCS USAR. Prior to Amendment 216 (Reference 7.5), TS 3.5(5)g required the Containment

LIC-05-0098 Attachment I Page 5 Structural Tests Report of TS 5 . 9 . 3 ~to be submitted if the acceptance criteria of TS 3.5(5)f(i) were not met. The acceptance criteria of TS 3.5(5)f(i) concerned the detection of significant structural deterioration by visual inspections of the exterior surface of the containment building.

The visual inspections specified in TS 3,5(5)b(i) through (iii) were to detect large areas of spall, severe scaling, grease leakage, D-cracking in areas exceeding 25 square feet, and other significant structural deterioration.

OPPD inspects the containment structure in accordance with the criteria of ASME Section XI, Subsections IWE and IWL, 1992 Edition with the 1992 Addenda. As required by paragraph (b)(2)(viii) of 10 CFR 50.55a, OPPD reports conditions indicative of containment deterioration or degradation in the IS1 Summary Report required by ASME Section XI, Subsection IWA-6000, and TS 5.9.3a. Thus, the Containment Structural Tests Report of TS 5 . 9 . 3 ~can be deleted since it is redundant to the requirements of 10 CFR 50.55a, and TS 5.9.3a.

Changes 3-8 The remaining changes noted in Section 2.0 above correct administrative, grammatical, or typographical errors identified by FCS staff. The Commission has provided guidance concerning the application of standards for determining whether a significant hazards consideration exists by providing certain examples (Attachment 2) of amendments that are considered not likely to involve significant hazards considerations. The proposed changes noted above cite one or more of these examples.

4.0 TECHNICAL ANALYSIS

Change 1 TS 2.10.2(9)b(i) allows SDM during low power physics testing to be reduced to the worth of the highest estimated CEA from the operable withdrawn CEAs and clarifies that during measurement of CEA worth, an allowance for the most reactive stuck CEA (of the groups withdrawn) is assumed when calculating SDM.

Each CEA not fully inserted must be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 7 days of reducing SDM to less than the limits of TS 2.10.2(1). During low power physics testing, the position of each CEA required to be trippable is determined at least once every two hours. These requirements are similar to those of NUREG-1432, (Reference 7.3) and ensure that SDM is sufficient at all times to enable the reactor to be quickly shutdown if necessary.

An audible count rate signal and makeup controller alarm warn the control room to allow corrective actions to be taken to isolate the primary makeup water source by closing valves andlor stopping the primary makeup water pumps or the charging pumps. Because of the equipment and controls and the administrative procedures provided for the boron dilution operation, the probability of erroneous dilution is considered very small. Nevertheless, if an unintentional dilution of boron in the reactor coolant does occur, numerous alarms and indications are available to alert the operator to the condition. For the hot standby, hot shutdown, cold shutdown, and refueling modes, the maximum reactivity addition due to the

LIC-05-0098 Page 6 dilution is slow enough to allow the operator to determine the cause of the dilution and take corrective action before the required shutdown margin is completely lost (Reference 7.8).

The SDM surveillance of TS 2.10.2(9)b(iii) whether performed at an 8 or 12-hour interval, is unlikely to discover that SDM is insufficient before the control room is alerted by an increase in the count rate or makeup controller alarm. Deletion of the requirement to verify SDM each shift is acceptable, because the SRs that remain in TS 2.10.2(9)b ensure that the SDM provided by the CEAs is adequate and that the CEAs are capable of full insertion. CEA positions will continue to be verified at least once per 2-hour interval during low power physics testing. Should the SDM provided by the CEAs be unavailable, boration is initiated immediately and continued until the SDM required by the COLR is met.

Change 2 Paragraph (b)(2)(viii) of 10 CFR 50.55a, requires that conditions indicative of containment deterioration or degradation be reported in the IS1 Summary Report required by ASME Section XI, Subsection IWA-6000, and TS 5.9.3a. Thus, the Containment Structural Tests Report of TS 5 . 9 . 3 ~is redundant to the requirements of 10 CFR 50.55a and TS 5.9.3a.

Changes 3-8 The remaining changes are administrative in nature as explained above.

5.0 REGULATORY SAFETY ANALYSIS 5.1 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 by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

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

Response: No This license amendment request (LAR) makes no changes to the design or operation of the plant that could affect system, component, or accident functions.

The deletion of Technical Specification (TS) 2.10.2(9)b(iii) eliminates the need to verify shutdown margin (SDM) every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> during low power physics testing. Reactivity equivalent to at least the highest estimated CEA worth is available from the operable CEA groups withdrawn (assuming the most reactive CEA of the groups withdrawn is stuck in the fully withdrawn position). Each CEA not fully inserted is demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 7 days of reducing SDM. Finally, the position of the trippable control element assemblies (CEAs) during low power physics testing continues to be verified every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The

LIC-05-0098 Page 7 SDM provided by the CEAs ensures that the operators can respond promptly to unexpected increases in core reactivity. Thus, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Deletion of the Containment Structural Tests Report is not an initiator of any previously evaluated accidents. OPPD will continue to report conditions indicative of containment deterioration or degradation in the Inservice Inspection (ISI) Summary Report required by 10 CFR 50.55a, ASME Section XI, Subsection IWA-6000, and TS 5.9.3a.

The Commission has provided guidance concerning the application of standards for determining whether a significant hazards consideration exists by providing certain examples (48 FR 14864) of amendments that are considered not likely to involve significant hazards considerations. One or more of these examples are cited to justify deletion of the Containment Structural Tests Report and for each of the remaining administrative changes. Thus, these changes do not increase the probability or consequences of any 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 This proposed change affects only the TSs and does not involve a physical change to the plant. No modifications are made to existing components nor will any new or different type of equipment be installed. The deletion of the surveillance requirement (SR) to verify SDM every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> during low power physics testing does not create the possibility of a new or different kind of accident.

The SRs that remain ensure that the SDM provided by the CEAs is adequate and that the CEAs are capable of full insertion. CEA positions will continue to be verified at least once per 2-hour interval during low power physics testing. The SDM provided by the CEAs ensures that the operators can respond promptly to unexpected increases in core reactivity.

The deletion of a report that is redundant to federal regulations is an administrative change that does not create the possibility of a new or different kind of accident. OPPD will continue to report conditions indicative of containment deterioration or degradation in the IS1 Summary Report.

The remaining changes proposed by this LAR are administrative in nature. These changes do not impose different requirements and do not alter assumptions made in the safety analysis and licensing basis. Therefore, they do not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

LIC-05-0098 Page 8 Response: No The proposed changes do not affect any safety analysis assumptions. During low power physics testing, the position of the trippable CEAs will continue to be verified at 2-hour intervals. The deleted 8-hour SDM surveillance requirement is performed less frequently, is redundant and unnecessary. The SDM provided by the CEAs ensures that the operators can respond promptly to unexpected increases in core reactivity. The Containment Structural Tests Report can be deleted since OPPD will continue to report conditions indicative of containment deterioration or degradation in accordance with 10 CFR 50.55a in the IS1 Summary Report required by TS 5.9.3a.

The TS changes proposed by this LAR include deletion of an unnecessary surveillance requirement, deletion of a redundant reporting requirement, and administrative revisions.

The changes do not impose different requirements and do not alter assumptions made in the safety analysis and licensing basis. Therefore, the changes proposed by this LAR do not involve a significant reduction in a margin of safety. Based on the above, OPPD concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory RequirementsICriteria Change 1 Section XI of 10 CFR 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Processing Plants," requires that a test program be established to ensure that structures, systems, and components will perform satisfactorily in service. All functions necessary to ensure that specified design conditions are not exceeded during normal operation and anticipated operational occurrences must be tested. Testing is required prior to initial criticality, after each refueling shutdown, during startup, low power operation, power ascension, and at power operation. The physics tests requirements for reload fuel cycles ensure that the operating characteristics of the core are consistent with the design predictions, and that the core can be operated as designed (Reference 7.6).

Appendix I, "Startup Manual" of the USAR defines the requirements for initial testing of the facility, including physics tests. Requirements for reload fuel cycle physics tests are defined in ANSYANS- 19.6.1-1985 (Reference 7.6). Although these physics tests are generally accomplished within the limits of all LCOs, conditions may occur when one or more LCOs must be suspended to make completion of physics tests possible or practical. This is acceptable as long as the fuel design criteria are not violated. As long as LHR remains within its limit, fuel design criteria are preserved.

The initial condition criteria for accidents sensitive to core power distribution are preserved by the LHR and DNB parameter limits. The criteria for the loss-of-coolant accident (LOCA) are specified in 10 CFR 50.46, Acceptance Criteria for Emergency core Cooling Systems for Light Water Nuclear Power Reactors. The criteria for the loss of forced reactor coolant flow

LIC-05-0098 Page 9 accident are specified in Chapter 14.6 of the USAR. Operation within the LHR limit preserves the LOCA criteria; operation within the DNB parameter limits preserves the loss of flow criteria.

The SDM exemption of TS 2.10.2(9) provides that a minimum amount of CEA worth is immediately available for reactivity control when tests are performed for CEA worth measurement. This special test exemption is required to permit the periodic verification of the actual versus predicted core reactivity condition occurring because of fuel bunzup or fuel cycling conditions. It is acceptable to delete the requirement to verify SDM each 8-hour shift during low power physics testing because CEA positions will continue to be verified at 2-hour intervals. This ensures that the operators can respond promptly to unexpected increases in core reactivity.

Change 2 With the approval of Amendment 216, the FCS technical specifications no longer contain a requirement to submit the Containment Structural Tests Report. OPPD inspects the containment structure using the criteria of ASME Section XI, Subsections IWE and IWL, 1992 Edition with the 1992 Addenda. As required by paragraph (b)(2)(viii) of 10 CFR 50.55a7 OPPD reports conditions indicative of containment deterioration or degradation in the IS1 Summary Report required by ASME Section XI, Subsection IWA-6000, and TS 5.9.3a. Since submittal of the Containment Structural Tests Report (1) is no longer required, and (2) is redundant to the requirements of 10 CFR 50.55a7 and TS 5.9.3a7 it should be deleted. This is considered an administrative change based on the discussion above.

Changes 3-8 Section 2.0 above provides a description of each proposed change and a justification as to why the change is considered administrative. The Commission has provided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing certain examples (Attachment 2) of amendments that are considered not likely to involve significant hazards consideration. The proposed changes cite one or more of these examples.

5.2.2 Design Basis (USAR)

FCS was licensed for construction prior to May 21, 1971, and at that time committed to the draft General Design Criteria (GDC) (Reference 7.7). The draft general design criteria are contained in Appendix G of the FCS USAR.

Change 1 The deletion of TS 2.10.2(9)b(iii) complies with FCS Design Criterion 6, Reactor Core Design, which is similar to CriterionlO, Reactor design of Appendix A to 10 CFR 50. The reactor core is designed to function throughout its design lifetime without exceeding acceptable fuel damage limits that have been stipulated and justified. The core design together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal

LIC-05-0098 Page 10 operation with appropriate margins for uncertainties and for transient situations that can be anticipated. Transient situations include the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all off-site power. This criterion is met. The fuel is designed for all expected conditions of normal operation and for anticipated transients.

The deletion of TS 2.10.2(9)b(iii) complies with FCS Design Criterion 13, Fission Process Monitors And Controls, which is similar to Criterion 27, Combined reactivity control systems capability of Appendix A to 10 CFR 50. Means shall be provided for monitoring and maintaining control over the fission process throughout core life and for all conditions that can reasonably be anticipated to cause variations in reactivity of the core, such as indication of position of control rods and concentration of soluble reactivity control poisons. This criterion is met.

The plant is provided with means to monitor and maintain control over the fission process throughout core life.

The deletion of TS 2.10.2(9)b(iii) complies with FCS Design Criterion 14, Core Protection Systems, which is similar to Criterion 20, Protection systems function of Appendix A to 10 CFR 50. Core protection systems together with associated equipment, shall be designed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits. This criterion is met. The reactor is protected by the reactor protection system from reaching a condition at which fuel damage might occur. The protection system is designed to monitor the reactor operating conditions and initiate a fast shutdown if any of measured variables exceed the operating limits. The signals, which will provide automatic reactor trip are identified in Table 7.2.-1 of the FCS USAR.

The deletion of TS 2.10.2(9)b(iii) complies with FCS Design Criterion 27 -

Redundancy Of Reactivity Control, which is similar to Criterion 22, Protection system independence of Appendix A to 10 CFR 50 At least two independent reactivity control systems, preferably of different principles, shall be provided.

This criterion is met. The reactivity control system employs two separate methods of adjusting reactivity; (1) mechanically driven control element assemblies and (2) adjustment of the concentration of boric acid in the reactor coolant. The CEA system controls short-term reactivity changes such as power changes and power distribution shaping, and is used for rapid shutdown for reactor protection. The boric acid shim control compensates for long term reactivity changes such as those associated with he1 burnup, variation in the xenon and samarium concentrations, and plant cooldown and heatup.

The deletion of TS 2.10.2(9)b(iii) complies with FCS Design Criterion 28, Reactivity Hot Shutdown Capability, which is similar to Criterion 26, Reactivity control system redundancy and capability of Appendix A to 10 CFR 50. At least two of the reactivity control systems provided shall independently be capable of

LIC-05-0098 Attachment I Page 11 making and holding the core subcritical from any hot standby or hot operating condition, including those resulting from power changes sufficiently fast to prevent exceeding acceptable fuel damage limits. This criterion is met. Two independent systems are provided for controlling reactivity by the addition or removal of poison from the core: (1) the mechanically driven control element assemblies and (2) the variation, by feed and bleed, of the concentration of dissolved boric acid in the reactor coolant. Either system acting independently is capable of making the core subcritical for a hot operating condition and holding it subcritical in the hot standby condition; in this context, hot standby implies a reactor coolant temperature not less than 515°F. With the shutdown margin available in the control element assemblies, a temperature reduction of at least 100°F from the hot standby condition can be sustained by the inserted control element assemblies (with the most reactive rod stuck) before boron injection is necessary to prevent any return to criticality.

It is also a requirement that either system be able to insert negative reactivity at a sufficiently fast rate to prevent exceeding acceptable fuel damage limits as the result of a power change. This criterion is met. The CEAs are inherently capable of inserting negative reactivity at a rate greater than that achieved by the charging pumps and the concentrated boric acid solution.

The deletion of TS 2.10.2(9)b(iii) complies with FCS Design Criterion 29, Reactivity Shutdown Capability, which is similar to Criterion 26, Reactivity control system redundancy and capability of Appendix A to 10 CFR 50. At least one of the reactivity control systems provided shall be capable of making the core subcritical under any condition (including anticipated operational transients) sufficiently fast to prevent exceeding acceptable fuel damage limits. Shutdown margins greater than the maximum worth of the most effective control rod when hlly withdrawn shall be provided. This criterion is met. The mechanical control system is capable of making the core subcritical under any condition, including anticipated operational transients, sufficiently fast to prevent fuel damage in excess of acceptable limits.

This control system is designed to provide a minimum shutdown margin of 2.4%

Ap assuming all CEAs, except the one of highest worth, are inserted in the core.

The deletion of TS 2.10.2(9)b(iii) complies with FCS Design Criterion 30, Reactivity Hold Down Capability, which is similar to Criterion 27, Combined reactivity control systems capability of Appendix A to 10 CFR 50. At least one of the reactivity control systems provided shall be capable of making and holding the core subcritical under any condition with appropriate margins for contingencies.

This criterion is met. The chemical control system changes reactivity slowly, but its range of reactivity worth is very large. It can handle the total excess reactivity plus a large shutdown margin (greater than 4% with all control rods withdrawn). It can also shut the reactor down with appropriate margins for contingencies from any normal operating condition without the use of control rods.

LIC-05-0098 Page 12 The deletion of TS 2.10.2(9)b(iii) complies with FCS Design Criterion 31, Reactivity Control Systems Malfunction, which is similar to Criterion 25, Protection system requirements for reactivity control functions of Appendix A to 10 CFR 50.

The reactivity control systems shall be capable of sustaining any single malfunction, such as unplanned continuous withdrawal (not ejection) of a control rod, without causing a reactivity transient, which could result in exceeding acceptable fuel damage limits. This criterion is met. Limits have been placed on the maximum rate at which reactivity can be increased by unplanned continuous withdrawal of CEAs. The number of CEAs in the core, the assignment of CEAs into operating groups and the design rate of withdrawal were established to assure fuel integrity in the event of uncontrolled CEA withdrawal. While an inadvertent withdrawal of CEAs is considered unlikely, the reactor protective system is designed to terminate any such transient with an adequate margin to DIVB. The analysis that supports this is described in Section 14.2 of the USAR. This analysis shows that sufficient protection is provided by the high power level trip, the high pressurizer pressure trip, the thermal margin trip and the steam generator water level trip to prevent the minimum DNB ratio from falling below 1.3 in the event of continuous withdrawal of CEAs.

The deletion of TS 2.10.2(9)b(iii) complies with FCS Design Criterion 32, Maximum Reactivity Worth Of Control Rods, which is similar to Criterion 28, Reactivity limits of Appendix A to 10 CFR 50. Limits which include considerable margin, shall be placed on the maximum reactivity worth of control rods of elements and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactivity cannot (a) rupture the reactor coolant pressure boundary or (b) disrupt the core, its support structures, or other vessel internals sufficiently to impair the effectiveness of emergency core cooling. This criterion is met. Limits have been placed on the maximum reactivity worth of individual CEAs so that the rate at which reactivity can be increased by CEA ejection, for example, will not cause an unacceptable rupture of the coolant pressure boundary or disrupt the core or reactor internals sufficiently to impair the effectiveness of emergency core cooling. The number and extent of insertion of CEAs in the core was selected to assure that the maximum reactivity worth of a single CEA is within a preselected safe limit. To confirm this, an analysis was made for the assumption that a CEA is ejected instantaneously from the core. The analysis, which is described in Section 14.13 of the USAR, shows that the energy increase at the core hot spot is limited to the extent that no fuel rods suffer significant damage following CEA ejection from full or zero power at the beginning or end of cycle. In addition, it has been calculated that the pressure surge associated with this excursion will not rupture the reactor coolant boundary.

Changes 2 through 8 The remaining changes proposed by this LAR are administrative in nature and comply with all FCS Design Criterion

LIC-05-0098 Page 13 5.2.3 Approved Methodologies See Section 7.0.

5.2.4 Analysis No analyses were performed to support this LAR.

5.2.5 Conclusion 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 EVALUATION A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

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

And The proposed amendment is confined to (i) changes to surety, insurance, and/or indemnity requirements, or (ii) changes to recordkeeping, reporting, or administrative procedures or requirements. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 5 1.22(~)(10).Therefore, pursuant to 10 CFR 5 1.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

7.1 Letter from NRC (A. B. Wang) to OPPD (R. T. Ridenoure) dated February 18, 2005, "Fort Calhoun Station, Unit No. 1. Issuance Of Amendment [23 11 Re:

Elimination Of Requirements To Provide Monthly Operating Reports And Annual Occupational Radiation Exposure Reports (TAC No. MC4306)" (NRC-05-0024) (ML050550061)

LIC-05-0098 Page 14 7.2 Letter from NRC (A. B. Wang) to OPPD (R. T. Ridenoure) dated July 23, 2004, "Fort Calhoun Station, Unit No. 1 - Issuance Of Amendment [228] Re: Various Administrative And Editorial Changes (TAC No. MC1535)" (NRC-04-0099)

(ML042110411) 7.3 NUREG-1432, Rev. 3, Standard Technical Speczfications Combustion Engineering Plants, June 2004 7.4 Letter from NRC (A. B. Wang) to OPPD (R. T. Ridenoure) dated May 8, 2003, "Fort Calhoun Station, Unit No 1 - Issuance of Amendment No. 218, Various Administrative and Editorial Changes TAC No. MB7498" (NRC-03-0091)

(ML031290081) 7.5 Letter from IWC (B. Benney) to OPPD (R. T. Ridenoure) dated February 26, 2003, "Fort Calhoun Station, Unit No 1 - Issuance of Amendment [216] RE:

Relocation of Technical Specification 3.5(5), Requirements for Testing Prestressed Concrete Containment Tendons, to The Updated Safety Analysis Report (TAC No. MB6472)" (NRC-03-0042) (ML030630727) 7.6 ANSUANS-19.6.1-1985, December 13, 1985 7.7 Draft General Design Criteria (32 FR 10213), July 11, 1967.

7.8 USAR Section 14.3, Boron Dilution Event

LIC-05-0098 Page 1 Examples of Amendments That Are Considered Not Likely To Involve Significant Hazards Considerations 48 FR 14864 Unless the speczfic circumstances of a license amendment request, when measured against the standards in Section 50.92, lead to a contrary conclusion then, pursuant to the procedures in Section 50.91, a proposed amendment to an operating license for a facility licensed under Section 50.21(b) or Section 50.22 or for a testing facility will likely be found to involve no signzficant hazards considerations, if operation of the facility in accordance with the proposed amendment involves only one or more of the following:

(i) A purely administrative change to technical speczfications: for example, a change to achieve consistency throughout the technical speczfications, correction of an error, or a change in nomenclature.

(ii) A change that constitutes an additional limitation, restriction, or control not presently included in the technical speczfications: for example, a more stringent surveillance requirement.

(iii) For a nuclear power reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies signzficantly difSerentfiom those found previously acceptable to the NRC for a previous core at the facility in question are involved. This assumes that no signzficant changes are made to the acceptance criteria for the technical speczfications, that the analytical methods used to demonstrate conformance with the technical speczfications and regulations are not signzficantly changed, and that NRC has previously found such methods acceptable.

(iv) A relief granted upon demonstration of acceptable operation from an operating restriction that was imposed because acceptable operation was not yet demonstrated. This assumes that the operating restriction and the criteria to be applied to a request for relief have been established in a prior review and that it is justzped in a satisfactory way that the criteria have been met.

(v) Upon satisfactory completion of construction in connection with an operating facility, a relief granted from an operating restriction that was imposed because the construction was not yet completed satisfactorily. This is intended to involve only restrictions where it is justzfied that construction has been completed satisfactorily.

(vi) A change which either may result in some increase to the probability or consequences of a previously-analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component speczped in the Standard Review Plan: for example, a change resulting from the application of a small refinement of a previously used calculational model or design method.

(vii) A change to make a license conform to changes in the regulations, where the license change results in very minor changes tofacility operations clearly in keeping with the regulations.

(viii) A change to a license to reflect a minor adjustment in ownership shares among co-owners already shown in the license.

LIC-05-0098 Page 1 MARKUP OF TECHNICAL SPECIFICATION PAGES

TECHNICAL SPECIFICATION TABLE OF CONTENTS (Continued) 4.4 Fuel Storage 4.4.1 New Fuel Storage 4.4.2 Spent Fuel Storage 4.5 Seismic Design for Class I Systems 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.2 Organization 5.3 Facility Staff Qualifications 5.4 Training 5.5 5.6 5.7 Safety Limit Violation 5.8 Procedures 5.9 Reporting Requirements 5.9.1 5.9.2 5.9.3 Special Reports 5.9.4 Unique Reporting Requirements 5.9.5 Core Operating Limits Report 5.9.6 RCS Pressure-Temperature Limits 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 Environmental Monitoring 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 5.2 1 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. 3?34+$4,55,5?,

-!,!52,!57,!24j+85, a

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

d. With both PORVs inoperable in Modes 4 and vent the RCS through at least a 0.94 square inch or larger vent within the (5) Two power-operated relief valves (PORVs) and their associated block valves shall be operable in Modes l , 2 , and 3.

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

inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either to operable status or close its associated block valve and remove power from the PORV to operable status within the following HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

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

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

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

The TS 2.1.6(1) lift settings are determined during Surveillance Testing in accordance with ASME Code test methods. The ASME Code requires that valves in steam service use steam as the test medium for establishing the setpoint. The +I%/-3% tolerance range specified in TS 2.1.6(1) applies to opening pressures determined during Surveillance Testing. When the valves are installed in the system, the presence of a water-filled loop seal at the valve inlets may result in in-situ actuation at a pressure that differs from the actuation pressure with steam at the inlet. Comparative testing and analysis indicates that with a loop seal present, the opening pressure of these valves may be up to 1% lower than the opening pressure under normal test conditions. Opening pressures below the specified setpoints are not a concern with respect to the safety limit for RCS pressure. Analysis of loss of load case involving elevated PSV opening pressur indicated that RCS pressures remained below the 2750 psia Safety Limit with PSV opening pressures p a to 6% above nominal setpoints. The valves are set to a tolerance of 1% of setpoint using ASME Cod te methods before being returned to service after testing. This allows for some setpoint variance over the surveillance interval.

0 2.1 -Page 21 Amendment No. 54&%,!57,!6!,l-89, ZB

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) 2.10.2 Reactivity Control Systems and Core Phvsics Parameters (Continued)

2. The position of each trippable CEA required shall be determined at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and

3. Each CEA not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 7 days prior to reducing the Shutdown Margin to less than the limits of Specification 2.10.2(1).

(ii) If the shutdown margin specified in part (i) above is not available, immediately initiate and continue boration until the requirements of 2.10.2(1) are met.

c. Moderator Temperature Coefficient

0) The moderator temperature coefficient (MTC) requirements of 2.10.2(3) may be suspended during physics tests at less than lo"% of rated power.

(ii) If power exceeds lo-'% of rated power, either:

1. Reduce power to less than lo-'% of rated power within 15 minutes, or

2. Be in hot shutdown in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Basis Shutdown Margin A sufficient shutdown margin ensures that (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

Shutdown margin requirements vary throughout core life as a h c t i o n of fuel depletion, RCS boron concentration, and RCS T,,. The most restrictive condition occurs at EOL, with Tavgat no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown. In the analysis of this accident, a minimum shutdown margin as specified in the COLR is initially adequate to control the reactivity transient. Accordingly, 2.10 -Page 9 Amendment No. 32,?3,'! 7,70,! 48,442

TECHNICAL SPECIFICATIONS TABLE 3-13 STEAM GENl RATOR TUBE INSPECTION 1st Sample Inspection 2ndSample Inspection 3rd Sample Inspection Action Required Result Action Required Result Action Required None N/A N/A N/A N/A Plug or repair defective None tubes and inspect additional 600 tubes in C-2 Plug or repair defective tubes and this S.G. inspect additional 1200 tubes in this S.G.

defective tubes C-3 Perform action for C-3 result of first C-3 Perform action sample for C-3 result of first sample Inspect all tubes in this None

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.8 Procedures (Continued)

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 Standing Orders as the responsible department head for that procedure within 14 days of implementation.

5.8.3 Written 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 report in^ 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.

5.0 - Page 6 Amendment No. 9,!?,35,?5,!4944%

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.9 Reporting Requirements (Continued) 5.9.2 Not Used 5.9.3 Special Reports Special reports shall be submitted to the appropriate NRC Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification where appropriate:

a. In-service inspection report, reference 3.3.

e. DELETED

f. DELETED

g. Materials radiation surveillance specimens reports, reference 3.3.

h. DELETED

i. Post-accident monitoring instrumentation, reference 2.21

j. Electrical systems, reference 2.7(2).

TECHNICAL SPECIFICATIONS TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.9.6 Reactor Coolant System (RCS) Pressure - Temperature Limits Report (PTLR)

a. RCS pressure and temperature limits for heatup, cooldown, low temperature overpressure protection, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for Technical Specifications 2.1.1 and 2.1.2.

b. The analytical methods used in the PTLR shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

CE NPSD-683-A, Revision 6, "Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications," April 200 1.

WCAP-15443, Revision 0, "Fast Neutron Fluence Evaluations for the Fort Calhoun Unit 1 Reactor Pressure Vessel," July 2000.

Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment Number 199 to Facility Operating License DPR-40 Omaha Public Power District Fort Calhoun Station, Unit Number 1, dated June 7,2001.

CEN-636, Revision 2, "Evaluation of Reactor Vessel Surveillance Data Pertinent to the Fort Calhoun Reactor Vessel Beltline Materials, dated July 2000.

FC06876, Revision 0, "Performance of Low Temperature Overpressure Protection System Analyses Using RELAPS: Methodology Paper."

FC06877, "Low Temperature Overpressure Protection (LTOP) Analysis, Revision 1."

Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment Number 207 to Facility Operating License Number DPR-40 Omaha Public Power District Fort Calhoun Station, Unit Number 1, dated April 22,2002.

Letter LTR-CI-01-25, Revision 0 from Westinghouse Electric Company (S.T. Byrne) to OPPD (J. Jensen), "Assessment of Extended Beltline Limit for Fort Calhoun Station Reactor Pressure Vessel," dated December 18,2001.

WCAP-15741, Revision 0, "Reactor Vessel Surveillance Program Withdrawal Schedule Modifications," dated September 2001.

Letter from NRC (A. B. Wang) to Omaha Public Power District (R. T. Ridenoure), Fort Calhoun Station - Unit 1, Exemption from the Requirements of Appendix G to 10 CFR 50 (TAC No. MB8237), dated July 30,2003.

Letter from Information Systems Laboratories (William Arcieri) to OPPD (J. Jensen),

"WCA-09-2002: Transmittal of RELAP5/MOD3.2d," dated August 2,2002.

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period (i.e., the number of EFPY used in the P-T 1imitILTOP analysis) and for any revision or supplement thereto.

Amendment No.

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.10 Record Retention 5.10.1 Records shall be retained as described in the Quality Assurance Program.

5.1 1 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.

5.1 1.1 In lieu of the "control device" required by paragraph 20.160 1(a) of 10 CFR Part 20, and as an alternative method allowed under 8 20.1601(c), each high radiation area (as defined in 8 20.1601) in which the intensity of radiation is 1000 me& 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.

5.1 1.2 The requirements of 5.1 1.1, above, shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 me&** but less than 500 rads/hr*** (Restricted 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 andlor the MRP with the following exception:

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

• 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.

• • At 30 centimeters (1 2 inches) fiom the radiation source or fiom any surface penetrated by the radiation.

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

Amendment IVo. %,6! ,I 32,1443184;.

-, 203

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.1 1 Radiation Protection Promam (Continued)

1. roped off such that an individual at the rope boundary is exposed to 1000 mrem/hr or less, 11 conspicuously posted, and iii a flashing light shall be activated as a warning device.

5.12 Environmental Oualification Deleted 5.13 Secondary Water Chemistrv A secondary water chemistry monitoring program to inhibit steam generator tube degradation shall be implemented. This program shall be described in the station chemistry manual and shall include:

1. Identification of a sampling schedule for the critical parameters and control points for these parameters;

2. Identification of the procedures used to measure the values of the critical parameters;

3. Identification of process sampling points;

4. Procedures for the recording and management of data;

5. Procedures defining corrective actions for off control point chemistry conditions; and

6. A procedure identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective actions.

5.14 Svstems Integrity A program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels shall be implemented. This program shall include the following:

1. Provisions establishing preventive maintenance and periodic visual inspection requirements, and

2. Integrated leak test requirements for each system at a frequency not to exceed refbeling cycle intervals.

Amendment No. m , W

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.15 Post-Accident Radiological Sampling and Monitoring The following programs shall be implemented and maintained to ensure the capability to accurately monitor and/or sample and analyze radiological effluents and concentrations in a post-accident condition:

1. A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. (Any space which will require occupancy to permit an operator to aid in mitigation of, or recovery from, an accident is designated as vital.)

2. A program which will ensure the capability to obtain and analyze radioactive iodines and particulates in plant gaseous effluents.

These programs shall include the following:

1. Training of personnel.

2. Procedures for monitoring and/or sampling and analysis.

3. Provisions for maintenance of sampling and analysis equipment.

5.16 Radiological Effluents and Environmental monitor in^ 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 operating 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 operability of radioactive liquid and gaseous radiation monitoring instrumentation including operability 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 20.100 1-20.2401, Appendix B, Table 2, Column 2. For dissolved or entrained noble gases, the concentration shall be limited to 2.0 E-04 pCi/ml total activity.

5.0 - P Amendment No. !52,!6?,288,X

TECHIVICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.16 Radiological Effluents and Environmental Monitoring Programs (Continued)

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 fiom 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 with the ODCM on a quarterly basis.

f. Limitations on the operability 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 fiom radioactive material, other than noble gases, released in gaseous effluents to unrestricted areas times 10 CFR 20.100 1-20.2401, Appendix B, Table 2, Column 1. For shall be limited to five times 10 CFR 20.1001-20.2401,

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

1. Limitations on the annual and quarterly doses to an individual beyond the site boundary fiom Iodine-131, 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 fiom uranium fie1 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 the effluent monitoring program and modeling of environmental exposure pathways. 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:

5.0 - Amendment No. 432+4, X 2

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.16 Radiological Effluents and Environmental Monitoring Programs (Continued)

a. Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with 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 checks 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.

5.17 Offsite Dose Calculation Manual (ODCM)

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 with the appropriate analyses or evaluations justifying the change(s) and

2. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302,40 CFR Part 190,10 CFR 50.36a, and Appendix 1 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 Review Committee and the approval of the plant manager.

c. 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 which 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., monthlyear) the change was implemented.

5.18 Process Control Program (PCP)

Changes to the PCP:

Amendment No. -H&k64,W

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.18 Process Control Program (PCP) (Continued)

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 with the appropriate analyses or evaluationsjustifying the change(s) and

2. A determination that the change will maintain the overall conformance of the solidified waste program to existing requirements of federal, state, or other applicable regulations.

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

c. Temporary changes to the PCP 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 PCP as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the PCP 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.

5.19 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J. Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program, dated September 1995," as modified by the following exceptions:

(1) If the Personnel Air Lock (PAL) is opened during periods when containment integrity is not required, the PAL door seals shall be tested at the end of such periods and the entire PAL shall be tested within 14 days afier RCS temperature Tcold> 2 1O°F.

(2) Type A tests may be deferred for penetrations of the steel pressure retaining boundary where the nominal diameter does not exceed one inch.

(3) Elapsed time between consecutive Type A tests used to determine performance shall be at least 24 months or refueling interval.

(4) The first Type A test performed afier the November 1993 Type A test shall be no later than November 2008.

The containment design accident pressure (Pa) is 60 psig.

TECHNICAL SPECIFICATIONS ADMINISTRATIVE CONTROLS 5.19 Containment Leakage Rate Testing Promam (Continued)

The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.1% of containment air weight per day.

Leakage Rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is < 1.0 La. During unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 5 0.60 La Maximum Pathway Leakage Rate (MXPLR) for Type B and C tests and < 0.75 La for Type A tests.

b. Personnel Air Lock testing acceptance criteria are:

(1) Overall Personnel Air Lock leakage is < 0.1 La when tested at 2 Pa.

(2) For each PAL door, seal leakage rate is 1 0.01 La when pressurized to L 5.0 psig.

c. Containment Purge Valve (PCV-742AiBICD) testing acceptance criterion is:

For each Containment Purge Valve, leakage rate is < 18.000 SCCM when tested at L Pa.

d. If at any time when containment integrity is required and the total Type B and C measured leakage rate exceeds 0.60 La Minimum Pathway Leakage Rate (MNLPR), repairs shall be initiated immediately. If repairs and retesting fail to demonstrate conformance to this acceptance criteria within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, then containment shall be declared inoperable.

The provisions of Specification 3.0.1 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of Specifications 3.0.4 and 3.0.5 are applicable to the Containment Leakage Rate Testing Program.

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 without prior NRC approval provided the changes do not require either of the following:

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

2. A change to the USAR or uires NRC approval pursuant to 10 CFR 50.59.

Amendment No. MS+ZO2,2+5

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.20 Technical Specifications (TS) Bases Control Program (Continued)

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the USAR.

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 without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

5.2 1 Containment Tendon Testing Program This program provides controls for monitoring any tendon degradation in prestressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial operations. The Containment Tendon Testing Program, inspection fre accordance with Regulatory Guide 1.35, Revision ,1989 a d acceptance criteria shall be in The provisions of TS 3.0.1 and TS 3.0.5 are applicable to the Containment Tendon Testing Program inspection frequencies.

If the acceptance criteria are not met, an immediate investigation shall be made to determine the cause(s) and extent of the non-conformance to the criteria, and the results shall be reported to the Commission within 90 days via a special report in accordance with Technical Specification 5.9.3.

5.22 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1. An API gravity or an absolute specific gravity within limits,

2. A flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and

3. A clear and bright appearance with proper color, or a water and sediment content within limits;

b. Within 3 1 days following addition of the new fuel oil to storage tanks, verify that the properties of the new fuel oil, other than those addressed in a., above, are within limits for ASTM 2D fuel oil, and

c. Total particulate concentration of the fuel oil is 5 10 mgll when tested every 3 1 days.

The provisions of TS 3.0.1 and TS 3.0.5 are applicable to the Diesel Fuel Oil Testing Program test frequencies.

Amendment No. X4+24-6,229

LIC-05-0098 Page 1 PROPOSED TECHNICAL SPECIFICATION PAGES

TECHNICAL SPECIFICATION TABLE OF CONTENTS (Continued) 4.4 Fuel Storage 4.4.1 New Fuel Storage 4.4.2 Spent Fuel Storage 4.5 Seismic Design for Class I Systems 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.2 Organization 5.3 Facility Staff Qualifications 5.4 Training 5.5 Not Used 5.6 Not Used 5.7 Safety Limit Violation 5.8 Procedures 5.9 Reporting Requirements 5.9.1 Not Used 5.9.2 Not Used 5.9.3 Special Reports 5.9.4 Unique Reporting Requirements 5.9.5 Core Operating Limits Report 5.9.6 RCS Pressure-Temperature Limits 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 5.16.2 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 5.2 1 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. -,55,5?,

w > =

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

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

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

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

b. With one PORV inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to operable status or close its associated block valve and remove power from the block valve; restore the PORV to operable status within the following 72 I

hours or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN witlun the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

1

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

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

Basis The purpose of the two spring-loaded Pressurizer Safety Valves (PSV's) is to provide Reactor Coolant System (RCS) overpressure protection and thereby ensure that the Safety Limit for RCS pressure (i.e.,

2750 psia) is not exceeded for analyzed accidents. The maximum RCS pressure transient for an analyzed accident is associated with a Loss of Load event(2).

The TS 2.1.6(1) lift settings are determined during Surveillance Testing in accordance with ASME Code test methods. The ASME Code requires that valves in steam service use steam as the test medium for establishing the setpoint. The +I%/-3% tolerance range specified in TS 2.1.6(1) applies to opening pressures determined during Surveillance Testing. When the valves are installed in the system, the presence of a water-filled loop seal at the valve inlets may result in in-situ actuation at a pressure that differs from the actuation pressure with steam at the inlet. Comparative testing and analysis indicates that with a loop seal present, the opening pressure of these valves may be up to 1% lower than the opening pressure under normal test conditions. Opening pressures below the specified setpoints are not a concern with respect to the safety limit for RCS pressure. Analysis of loss of load case involving elevated PSV opening pressures indicated that RCS pressures remained below the 2750 psia Safety Limit with PSV opening pressures up to 6% above nominal setpoints. The valves are set to a tolerance off 1% of setpoint 1 using ASME Code test methods before being returned to service after testing. This allows for some setpoint variance over the surveillance interval.

2.1 -Page 21 Amendment No. 5?,!644-89,W

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.10 Reactor Core (Continued) 2.10.2 Reactivity Control Systems and Core Physics Parameters (Continued)

2. The position of each trippable CEA required shall be determined at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and

3. Each CEA not hlly inserted shall be demonstrated capable of h l l insertion when tripped from at least the 50%

withdrawn position within 7 days prior to reducing the Shutdown Margin to less than the limits of Specification 2.10.2(1).

(ii) If the shutdown margin specified in part (i) above is not available, immediately initiate and continue boration until the requirements of 2.10.2(1) are met.

c. Moderator Temperature Coefficient (0 The moderator temperature coefficient (MTC) requirements of 2.10.2(3) may be suspended during physics tests at less than lo-'%

of rated power.

(ii) If power exceeds lo-'% of rated power, either:

1. Reduce power to less than lo-'% of rated power within 15 minutes, or

2. Be in hot shutdown in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Basis Shutdown Margin A sufficient shutdown margin ensures that (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

Shutdown margin requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS Tavg. The most restrictive condition occurs at EOL, with Tavgat no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown. In the analysis of this accident, a minimum shutdown margin as specified in the COLR is initially adequate to control the reactivity transient. Accordingly, 2.10 - Page 9 Amendment No. 3,2,?3,?7,19, M,%

TECHNICAL SPECIFICATIONS TABLE 3-13 STEAM GENERATOR TUBE INSPECTION I I I 1st Sample Inspection I 2nd Sample Inspection I 3rd Sample Inspection 1 Sample Size Result Action Required Result Action Required Result Action Required A minimum of C- 1 None N/A N/A N/A N/A 300 tubes per S.G. C-2 Plug or repair C- 1 None N/A N/A defective tubes and inspect additional 600 C-2 Plug or repair defective C- 1 None tubes in this S.G. tubes and inspect additional 1200 tubes in C-2 Plug or repair this S.G. defective tubes C-3 Perform action for C-3 C-3 Perform action for result of first sample C-3 result of first sample C-3 Inspect all tubes in this The second None N/A N/A

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.8 Procedures (Continued)

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 Standing Orders as the responsible department head for that procedure within 14 days of implementation.

5.8.3 Written 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 Amendment No. 9,!?,35,75,! 4994%

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.9 Reporting Requirements (Continued) 5.9.2 Not Used I

5.9.3 Special Reports Special reports shall be submitted to the appropriate NRC Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification where appropriate:

In-service inspection report, reference 3.3.

Tendon surveillance, reference 5.2 1.

DELETED DELETED DELETED DELETED Materials radiation surveillance specimens reports, reference 3.3.

DELETED Post-accident monitoring instrumentation, reference 2.21 Electrical systems, reference 2.7(2).

5.0 - Page 7 Amendment No. 9,2?,35,3E,46%%8%%

1 1 1 1 1 A',A-'J,A 7 1 4 7 1 C 7, 1 0 11111 1 A",L-'b,xJ' 1 Q < '7n3 1 1 5 2 1 0 1 1 Q

> uJ7rwb,r , L'>L.hu>

234

TECHIVICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.9.6 Reactor Coolant System (RCS) Pressure - Temperature Limits Report (PTLR)

a. RCS pressure and temperature limits for heatup, cooldown, low temperature overpressure protection, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for Technical Specifications 2.1.1 and 2.1.2.

b. The analytical methods used in the PTLR shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

CE NPSD-683-A, Revision 6, "Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits and LTOP Requirements from the Technical Specifications," April 200 1.

WCAP-15443, Revision 0, "Fast Neutron Fluence Evaluations for the Fort Calhoun Unit 1 Reactor Pressure Vessel," July 2000.

Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment Number 199 to Facility Operating License DPR-40 Omaha Public Power District Fort Calhoun Station, Unit Number 1, dated June 7,2001.

CEN-636, Revision 2, "Evaluation of Reactor Vessel Surveillance Data Pertinent to the Fort Calhoun Reactor Vessel Beltline Materials, dated July 2000.

FC06876, Revision 0, "Performance of Low Temperature Overpressure Protection System Analyses Using RELAPS: Methodology Paper."

FC06877, "Low Temperature Overpressure Protection (LTOP) Analysis, Revision 1."

Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment Number 207 to Facility Operating License Number DPR-40 Omaha Public Power District Fort Calhoun Station, Unit Number 1, dated April 22, 2002.

Letter LTR-CI-01-25, Revision 0 from Westinghouse Electric Company (S.T. Byrne) to OPPD (J. Jensen), "Assessment of Extended Beltline Limit for Fort Calhoun Station Reactor Pressure Vessel," dated December 18,2001.

WCAP- 15741, Revision 0, "Reactor Vessel Surveillance Program Withdrawal Schedule Modifications," dated September 200 1.

Letter from NRC (A. B. Wang) to Omaha Public Power District (R. T. Ridenoure), Fort Calhoun Station - Unit 1, Exemption from the Requirements of Appendix G to 10 CFR Part 50 (TAC No. MB8237), dated July 30,2003.

Letter from Information Systems Laboratories (William Arcieri) to OPPD (J. Jensen),

"WCA-09-2002: Transmittal of RELAP5/MOD3,2d," dated August 2,2002.

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period (i.e., the number of EFPY used in the P-T limit/LTOP analysis) and for any revision or supplement thereto.

5.0 - Page 10 Amendment No. 2X

TECHNICAL SPECIFICATIONS 5.O ADMINISTRATIVE CONTROLS 5.10 Record Retention 5.10.1 Records shall be retained as described in the Quality Assurance Program.

5.1 1 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.

5.1 1.1 In lieu of the "control device" required by paragraph 20.160 1(a) of 10 CFR Part 20, and as an alternative method allowed under 5 20.160 1(c), each high radiation area (as defined in 5 20.160 1) 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.

5.1 1.2 The requirements of 5.1 1.1, 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*** (Restricted 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 with the following exception:

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

• 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.

• • 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.

5.0 - Page 11 Amendment No. 28,61,132,154+l-8%

-,=

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.1 1 Radiation Protection Promam (Continued)

1. roped off such that an individual at the rope boundary is exposed to 1000 m r e m h or less, 11 conspicuously posted, and iii a flashing light shall be activated as a warning device.

5.12 Environmental Oualification Deleted 5.13 Secondam Water Chemistrv A secondary water chemistry monitoring program to inhibit steam generator tube degradation shall be implemented. This program shall be described in the station chemistry manual and shall include:

1. Identification of a sampling schedule for the critical parameters and control points for these parameters;

2. Identification of the procedures used to measure the values of the critical parameters;

3. Identification of process sampling points;

4. Procedures for the recording and management of data;

5. Procedures defining corrective actions for off control point chemistry conditions; and

6. A procedure identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective actions.

5.14 Systems Intemitv A program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels shall be implemented. This program shall include the following:

1. Provisions establishing preventive maintenance and periodic visual inspection requirements, and

2. Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.

5.0 - Page 12 Amendment No. 53,88,89,93,202

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.15 Post-Accident Radiological Sampling and Monitoring The following programs shall be implemented and maintained to ensure the capability to accurately monitor and/or sample and analyze radiological effluents and concentrations in a post-accident condition:

1. A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. (Any space which will require occupancy to permit an operator to aid in mitigation of, or recovery from, an accident is designated as vital.)

2. A program which will ensure the capability to obtain and analyze radioactive iodines and particulates in plant gaseous effluents.

These programs shall include the following:

1. Training of personnel.

2. Procedures for monitoring and/or sampling and analysis.

3. Provisions for maintenance of sampling and analysis equipment.

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 operating 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 operability of radioactive liquid and gaseous radiation monitoring instrumentation including operability 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 pCi/ml total activity.

5.0 - Page 13 Amendment No. !52,!5?,288,202

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.16 Radiological Effluents and Environmental Monitoring Programs (Continued)

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 with the ODCM on a quarterly basis.

f. Limitations on the operability 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 20.100 1-20.2401, 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.

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.

1. Limitations on the annual and quarterly doses to an individual beyond the site boundary from Iodine-13 1, 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 the effluent monitoring program and modeling of environmental exposure pathways. 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:

5.0 - Page 14 Amendment No. 422+4,2@2

TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.16 Radiological Effluents and Environmental Monitoring Programs (Continued)

a. Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with 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 checks 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.

5.17 Offsite Dose Calculation Manual (ODCM)

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 with the appropriate analyses or evaluations justifying the change(s) and

2. A determination that the change will 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 Review Committee and the approval of the plant manager.

c. 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 which 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., monthlyear) the change was implemented.

5.18 Process Control Program (PCP)

Changes to the PCP:

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TECHNICAL SPECIFICATIONS 5.0 ADMINISTRATIVE CONTROLS 5.18 Process Control Program (PCP) (Continued)

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 with the appropriate analyses or evaluationsjustifying the change(s) and

2. A determination that the change will maintain the overall conformance of the solidified waste program to existing requirements of federal, state, or other applicable regulations.

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

c. Temporary changes to the PCP 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 PCP as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the PCP 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., monthlyear) the change was implemented.

5.19 Containment Leakage Rate Testing Prowam A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J. Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program, dated September 1995," as modified by the following exceptions:

(1) If the Personnel Air Lock (PAL) is opened during periods when containment integrity is not required, the PAL door seals shall be tested at the end of such periods and the entire PAL shall be tested within 14 days after RCS temperature Tcold> 210°F.

(2) Type A tests may be deferred for penetrations of the steel pressure retaining boundary where the nominal diameter does not exceed one inch.

(3) Elapsed time between consecutive Type A tests used to determine performance shall be at least 24 months or reheling interval.

(4) The first Type A test performed after the November 1993 Type A test shall be no later than November 2008.

The containment design accident pressure (Pa) is 60 psig.

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TECHNICAL SPECIFICATIONS ADMINISTRATIVE CONTROLS 5.19 Containment Leakage Rate Testing Proaam (Continued)

The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.1% of containment air weight per day.

Leakage Rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is I 1.0 La. During unit startup following testing in accordance with this program, the leakage rate acceptance criteria are I 0.60 La Maximum Pathway Leakage Rate (MXPLR) for Type B and C tests and F 0.75 La for Type A tests.

b. Personnel Air Lock testing acceptance criteria are:

(1) Overall Personnel Air Lock leakage is I 0.1 La when tested at 2 Pa, (2) For each PAL door, seal leakage rate is I 0.01 La when pressurized to > 5.0 psig.

c. Containment Purge Valve (PCV-742A/B/C/D)testing acceptance criterion is:

For each Containment Purge Valve, leakage rate is < 18.000 SCCM when tested at 2 Pa.

d. If at any time when containment integrity is required and the total Type B and C measured leakage rate exceeds 0.60 La Minimum Pathway Leakage Rate (MNLPR), repairs shall be initiated immediately. If repairs and retesting fail to demonstrate conformance to this acceptance criteria within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, then containment shall be declared inoperable.

The provisions of Specification 3.0.1 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of Specifications 3.0.4 and 3.0.5 are applicable to the Containment Leakage Rate Testing Program.

5.20 Technical Specifications (TS) Bases Control Promam 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 without prior NRC approval provided the changes do not require either of the following:

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

2. A change to the USAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

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TECHNICAL SPECIFICATIONS ADMINISTRATIVE CONTROLS 5.20 Technical Specifications (TS) Bases Control Program (Continued)

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the USAR.

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 without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

5.2 1 Containment Tendon Testing Program This program provides controls for monitoring any tendon degradation in prestressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial operations. The Containment Tendon Testing Program, inspection frequencies, and acceptance criteria shall be in accordance with Regulatory Guide 1.35, Revision 3, 1990. 1 The provisions of TS 3.0.1 and TS 3.0.5 are applicable to the Containment Tendon Testing Program inspection frequencies.

If the acceptance criteria are not met, an immediate investigation shall be made to determine the cause(s) and extent of the non-conformance to the criteria, and the results shall be reported to the Commission within 90 days via a special report in accordance with Technical Specification 5.9.3.

5.22 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1. An API gravity or an absolute specific gravity within limits,

2. A flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and

3. A clear and bright appearance with proper color, or a water and sediment content within limits;

b. Within 3 1 days following addition of the new fuel oil to storage tanks, verify that the properties of the new fuel oil, other than those addressed in a., above, are within limits for ASTM 2D fuel oil, and

c. Total particulate concentration of the fuel oil is 5 10 mgll when tested every 3 1 days.

The provisions of TS 3.0.1 and TS 3.0.5 are applicable to the Diesel Fuel Oil Testing Program test frequencies.

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