ML052700523

Fort Calhoun, License Amendment, Deletion of Shutdown Margin During Low Power Testing from TS and Various Other Editorial and Administrative Changes

ML052700523
Person / Time
Site:	Fort Calhoun
Issue date:	02/01/2006
From:	Wang A B Plant Licensing Branch III-2
To:	Ridenoure R T Omaha Public Power District
Wang A B,/NRR/DLPM/415-1445
References
TAC MC8095
Download: ML052700523 (12)

Text

February 1, 2006 Mr. R. T. Ridenoure

Vice President - Chief Nuclear Officer

Omaha Public Power District

Fort Calhoun Station FC-2-4 Adm.

Post Office Box 550

Fort Calhoun, NE 68023-0550

SUBJECT:

FORT CALHOUN STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT RE:

DELETION OF THE SHUTDOWN MARGIN SURVEILLANCE REQUIREMENT (TAC NO. MC8095)

Dear Mr. Ridenoure:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 237 to Renewed Facility Operating License No. DPR-40 for the Fort Calhoun Station, Unit No. 1. The amendment consists of changes to the Technical Specifications (TS) in

response to your application dated August 11, 2005.

The amendment deletes TS 2.10.2(9)b(iii). This surveillance requirement (SR) required the licensee to verify the shutdown margin on every 8-hour 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 Specifications Combustion Engineering Plants, Revision 3." In addition, the

Containment Structural Tests Report is deleted from TS 5.9.3c and several administrative and editorial changes were approved.

A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.Sincerely,/RA/Alan B. Wang, Project Manager Plant Licensing Branch IV

Division of Operating Reactor Licensing

Office of Nuclear Reactor Regulation Docket No. 50-285

Enclosures:

1. Amendment No. 237 to DPR-40

2. Safety Evaluation cc w/encls: See next page February 1, 2006 Mr. R. T. Ridenoure

Vice President - Chief Nuclear Officer

Omaha Public Power District

Fort Calhoun Station FC-2-4 Adm.

Post Office Box 550

Fort Calhoun, NE 68023-0550

SUBJECT:

FORT CALHOUN STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT RE:

DELETION OF THE SHUTDOWN MARGIN SURVEILLANCE REQUIREMENT (TAC NO. MC8095)

Dear Mr. Ridenoure:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 237 to Renewed Facility Operating License No. DPR-40 for the Fort Calhoun Station, Unit No. 1. The amendment consists of changes to the Technical Specifications (TS) in

response to your application dated August 11, 2005.

The amendment deletes TS 2.10.2(9)b(iii). This surveillance requirement (SR) required the licensee to verify the shutdown margin on every 8-hour 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 Specifications Combustion Engineering Plants, Revision 3." In addition, the

Containment Structural Tests Report is deleted from TS 5.9.3c and several administrative and editorial changes were approved.

A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.Sincerely,/RA/Alan B. Wang, Project Manager Plant Licensing Branch IV

Division of Operating Reactor Licensing

Office of Nuclear Reactor Regulation Docket No. 50-285

Enclosures:

1. Amendment No. 237 to DPR-40

2. Safety Evaluation cc w/encls: See next page DISTRIBUTION

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ACCESSION NO.: ML052700523OFFICEPDIV-2/PMPDIV-2/LASXRB/AIROB-AOGCLPLIV/BCNAMEAWangLFeizollahiNakowskiTBoyceJZornDTerao DATE1/9/061/9/0612/21/051/19/061/6/061/25/06 OFFICIAL RECORD COPY OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 237 License No. DPR-401.The Nuclear Regulatory Commission (the Commission) has found that:A.The application for amendment by the Omaha Public Power District (the licensee) dated August 11, 2005, complies with the standards and requirements

of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's

rules and regulations set forth in 10 CFR Chapter I;B.The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;C.There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the

public, and (ii) that such activities will be conducted in compliance with the

Commission's regulations;D.The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. 2.Accordingly, Renewed Facility Operating License No. DPR-40 is amended by changes to the Technical Specifications as indicated in the attachment to this license

amendment, and paragraph 3.B. of Facility Operating License No. DPR-40 is hereby

amended to read as follows:B.Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 237, are hereby incorporated in the license. The licensee shall

operate the facility in accordance with the Technical Specifications.3.The license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/David Terao, Branch Chief Plant Licensing Branch IV

Division of Operating Reactor Licensing

Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: February 1, 2006 ATTACHMENT TO LICENSE AMENDMENT NO. 237 RENEWED FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical

lines indicating the areas of change.

REMOVE INSERTTOC - Page 3TOC - Page 32.1 - Page 212.1 - Page 21 2.10 - Page 92.10 - Page 9 Table 3-13Table 3-13 5.0 - Page 65.0 - Page 6 5.0 - Page 75.0 - Page 7 5.0 - Page 105.0 - Page 10 5.0 - Page 115.0 - Page 11 5.0 - Page 125.0 - Page 12 5.0 - Page 135.0 - Page 13 5.0 - Page 145.0 - Page 14 5.0 - Page 155.0 - Page 15 5.0 - Page 165.0 - Page 16 5.0 - Page 175.0 - Page 17 5.0 - Page 185.0 - Page 18 5.0 - Page 19 -------

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 237 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-40 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. 1 DOCKET NO. 50-28

51.0INTRODUCTION

By application dated August 11, 2005 (A gencywide Documents Access and Management System Accession No. ML052240078), Omaha Pub lic Power District (OPPD) requested changes to the Technical Specifications (TSs)(Appendix A to Renewed Facility Operating

License No. DPR-40) for the Fort Calhoun Station, Unit No. 1 (FCS).

The proposed amendment will delete TS 2.10.2(9)b(iii). This surveillance requirement (SR) required the licensee to verify the shutdown margin (SDM) on every 8-hour 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 Specifications Combustion Engineering Plants, Revision 3."

In addition, OPPD proposes to delete the Containment Structural Tests Report from TS 5.9.3c

and make several administrative and editorial changes.

2.0REGULATORY EVALUATION

The SDM Test exemption of TS 2.10.2(9)b provides that a minimum amount of control element assembly (CEA) worth is immediately availabl e for reactivity control when physics tests are performed for CEA worth measurement. This will ensure that the operators can respond

promptly to unexpected increases in core reactivity during the tests. This special test exemption is needed to permit the periodic verification of the actual versus predicted core reactivity

condition occurring because of fuel burnup or fuel cycling conditions. The following criteria

apply to the SDM requirements:Section XI of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix B,"Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," requires

that a test program be established to ensure that structures, systems, and components will perform satisfactorily in service. All func tions 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.

Appendix I, "Startup Manual," of the FCS Updated Safety Analysis Report (USAR) defines the requirements for initial testing of the facility, including physics tests. Requirements for reload

fuel cycle physics tests are defined in ANSI/A NS-19.6.1-1985 (Reference 7.6 of the licensee's submittal). Although these physics tests are generally accomplished within the limits of all

limiting conditions for operations (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 Linear Heat Rate (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 a forced reactor coolant flow accident

are specified in Chapter 14.6 of the FCS USAR. Operation within the LHR limit preserves the

LOCA criteria; operation within the departure from nucleate boiling (DNB) parameter limits

preserve the loss of flow criteria.

The remaining changes proposed by this LAR are administrative or editorial in nature.

3.0TECHNICAL EVALUATION

3.1Deletion of SDM SR TS 2.10.2(9)b(iii)

Physics 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.

TS 2.10.2(9)b provides an exemption to the SD M required by TS 2.10.2(1). During the low power physics testing for CEA worth and SDM, the SDM requirement may be reduced provided

the following conditions are met: 1.The SDM can 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;2.During low power physics testing, the position of each CEA required to be trippable is determined at least once every two hours; and 3.Each CEA not fully inserted must be demonstrated capable of full insertion when tripped from at least the 50 percent withdrawn position within 7 days of reducing SDM to less

than the limits of TS 2.10.2(1).

These requirements ensure that SDM is sufficient at all times to enable the reactor to be quickly shutdown, if necessary, during low power physi cs testing. These requirements are similar to those of TS 3.1.7, "Special Test Excepti ons (STE) - SHUTDOWN MARGIN (SDM)" of NUREG-1432.

OPPD has proposed to delete TS 2.10.2(9)b(iii). During the SDM low power physics testing, TS 2.10.2(9)b(iii) states that SDM "shall be verified every 8-hour shift." The purpose of TS 2.10.2(9) is to permit relaxation of existing 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 is

needed to permit the periodic verification of the actual versus predicted core reactivity condition

occurring because of fuel burnup or fuel cycling conditions. This special test exemption assures

that a minimum amount of CEA worth is immedi ately available for reactivity control when physics tests are performed for CEA worth measurement.

Reactivity control during low power physics te sting is provided primarily by trippable 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. Requiring that shutdown reactivity equivalent to at least the highest

estimated CEA worth (of those CEAs actually withdrawn) be available for trip insertion from the

operable CEA provides a high degree of assurance that shutdown capability is maintained for

the most challenging postulated accident, a stuck CEA. If, however, the SDM requirement of

TS 2.10.2 is suspended, there is not the same degree of assurance during this test that the

reactor would always be shut down if the highest worth CEA was stuck out and calculational

uncertainties or the estimated highest CEA worth was not as expected (i.e., the single-failure

criterion is not met). However, the NRC staff concludes that this situation is acceptable

because specified acceptable fuel limits are still met. The risk of experiencing a stuck CEA and

subsequent criticality is reduced during this Physics Test exception by the requirements to

determine CEA positions every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />; by the trip of each CEA to be withdrawn 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior

to suspending the SDM; and by ensuring that shutdown reactivity is available, equivalent to the

reactivity worth of the estimated highest worth withdrawn CEA. Thus, the TS 2.10.2(9)b(iii)

requirement to verify SDM every 8-hour shift is 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. In addition, 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 and/or 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. Therefore, for the hot standby, hot shutdown, cold shutdown, and refueling modes, the maximum reactivity addition due to the

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.

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. As CEA positions will be verified at least once per 2-hour interval

during low power physics testing and provide adequate SDM, the 8-hour SDM SR is a

redundant requirement. In addition, it would be unlikely that the SDM surveillance of

TS 2.10.2(9)b(iii), which is performed at an 8-hour interval, would result in the discovery that

SDM is insufficient before the control room is alerted by an increase in the count rate or makeup

controller alarm. Should the SDM provided by the CEAs be unavailable, boration is initiated

immediately and continued until the SDM required by the COLR is met. Based on the above, the Nuclear Regulatory Commission (NRC) staff concludes that TS 2.10.2(9)b(iii) may be

deleted from the TSs. This change is consistent with SR 3.1.7 in NUREG-1432, which does not

include this SR. 3.2Deletion of Containment Structural Tests Report From TS 5.9.3c, "Special Reports" OPPD has proposed to delete "Containment Structur al Tests Report" from TS 5.9.3c, "Special Reports." Amendment 216 relocated TS 3.5(5), which contained requirements for submitting the Containment Structural Test report, to the FCS USAR. Thus, although the Containment

Structural Tests Report is listed in TS 5.9.3c, there is no TS requirement to submit it. Therefore, the Containment Structural Tests Report of TS 5.9.3c can be deleted from the TSs as a special

report that must be submitted to the NRC. In addition, 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 the American Society of Mechanical Engineers (ASME),Section XI, Subsection IWA-6000 requirements.

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 ISI Summary Report required by ASME,Section XI, Subsection IWA-6000, and TS 5.9.3a. Therefore, the information provided in this report is redundant to the

requirements of 10 CFR 50.55a and TS 5.9.3a, and will continue to be provided to the NRC

staff.3.3Table of Contents

Amendment 228 revised TSs 5.5, 5.6, and 5.9.2 to show that they are "Not Used" but did not revise the TOC accordingly. OPPD proposes to revise Page 3 of the Table of Contents (TOC)

to reflect the changes approved in Amendment 228 to show TSs 5.5, 5.6, and 5.9.2 as "Not

Used." Amendment 231 deleted the remaining routine reports (Monthly Operating Report and Annual Occupational Exposure Report) of TS 5.9.1. OPPD proposes to revise Page 3 of the TOC to reflect the changes approved in Amendment 231 to show TS 5.9.1 is "Not Used." The

NRC staff concludes that this change is administrative in nature and, therefore, is acceptable.

3.4TS 2.1.6 and its Basis The changes proposed for TS 2.1.6 improve grammatical structure by inserting appropriate punctuation and revising sentences with missing, incomplete, or inappropriate words. The NRC

staff has reviewed these changes and agree that they are editorial in nature and, therefore, are

acceptable. The NRC staff does not review Bases changes as they are controlled by TS 5.20, "Technical Specification (TS) Bases Control Program."3.5Table 3-13, Steam Generator Tube Inspections

OPPD proposes to revise Table 3-13, "Steam Generator Tube Inspections," to delete two occurrences of the phrase "Prompt notification to the NRC pursuant to specification 5.6."

Amendment 228 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 can be deleted. The NRC staff has reviewed

these changes and agree that they are administrative in nature and, therefore, are acceptable. 3.6 Page 10 of TS 5.0

OPPD has proposed to delete Page 10 of TS 5.0 which is currently designated as "Not Used." With the deletion of this page all subsequent TS 5.0 pages are being renumbered accordingly.

The NRC staff has reviewed these changes and agree that they are administrative in nature

and, therefore, are acceptable. 3.7 TS 5.16.1g

OPPD has proposed to revise the last sentence of TS 5.16.1g to change the period following the words "For noble gases" to a comma. The NRC staff has reviewed this change and

concluded it is editorial in nature and therefore, is acceptable. 3.8 TS 5.21

OPPD has proposed to revise TS 5.21 to correct an error showing Revision 3 of Regulatory Guide 1.35, "Inservice Inspection of Ungrouted Tendons in Prestressed Concrete

Containments," as issued in 1989. The correct date of Regulatory Guide 1.35, Revision 3 is

1990. The NRC staff has reviewed this change and agrees that it is editorial in nature and, therefore, is acceptable.

4.0STATE CONSULTATION

In accordance with the Commission's regulations, the Nebraska State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has

determined that the amendment involves no significant increase in the amounts, and no

significant change in the types, of any effluents that may be released offsite, and that there is no

significant increase in individual or cumulative occupational radiation exposure. The

Commission has previously issued a proposed finding that the amendment involves no

significant hazards consideration, and there has been no public comment on such finding

(70 FR 56503; September 27, 2005). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)9. Pursuant to 10 CFR 51.22(b), no

environmental impact statement or environm ental assessment need be prepared in connection with the issuance of the amendment.

6.0CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (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.

Principal Contributor: A. Wang

Date: February 1, 2006 September 2005 Ft. Calhoun Station, Unit 1 cc: Winston & Strawn ATTN: James R. Curtiss, Esq.

1400 L Street, N.W.

Washington, DC 20005-3502 Chairman Washington County Board of Supervisors

P.O. Box 466

Blair, NE 68008 Mr. John Hanna, Resident Inspector U.S. Nuclear Regulatory Commission

P.O. Box 310

Fort Calhoun, NE 68023 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission

611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 Ms. Sue Semerera, Section Administrator Nebraska Health and Human Services Systems

Division of Public Health Assurance

Consumer Services Section

301 Centential Mall, South

P.O. Box 95007

Lincoln, NE 68509-5007 Mr. David J. Bannister, Manager Fort Calhoun Station

Omaha Public Power District

Fort Calhoun Station FC-1-1 Plant

P.O. Box 550

Fort Calhoun, NE 68023-0550 Mr. Joe L. McManis Manager - Nuclear Licensing

Omaha Public Power District

Fort Calhoun Station FC-2-4 Adm.

P.O. Box 550

Fort Calhoun, NE 68023-0550 Mr. Daniel K. McGhee Bureau of Radiological Health

Iowa Department of Public Health

Lucas State Office Building, 5th Floor

321 East 12th Street

Des Moines, IA 50319