ML041320667

5/7/04 Fort Calhoun, Unit 1 - Issuance of Amd. 226 Reactor Coolant System (RCS) Leakage Limits (Tac. MC0194)

ML041320667
Person / Time
Site:	Fort Calhoun
Issue date:	05/07/2004
From:	Wang A NRC/NRR/DLPM/LPD4
To:	Ridenoure R Omaha Public Power District
Wang A, NRR/DLPM, 415-1445
Shared Package
ML041390280	List: ML041320667 ML041390122
References
TAC MC0194
Download: ML041320667 (14)

Text

May 7, 2004 Mr. R. T. Ridenoure Division Manager - Nuclear Operations 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: REACTOR COOLANT SYSTEM (RCS) LEAKAGE LIMITS (TAC NO. MC0194)

Dear Mr. Ridenoure:

The Commission has issued the enclosed Amendment No. 226 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 July 25, 2003,as supplemented on December 5, 2003.

The amendment modifies TS 2.1.4, Reactor Coolant System (RCS) Leakage Limits, by (1) adding a requirement for no RCS pressure boundary leakage, (2) combining the existing RCS leakage limits into a format similar to the Improved Standard TS (ISTS), and (3) replacing the existing basis associated with this TS with a basis similar in format and content to the ISTS.

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, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-285

Enclosures:

1. Amendment No. 226 to DPR-40

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

May 7, 2004 Mr. R. T. Ridenoure Division Manager - Nuclear Operations 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: REACTOR COOLANT SYSTEM (RCS) LEAKAGE LIMITS (TAC NO. MC0194)

Dear Mr. Ridenoure:

The Commission has issued the enclosed Amendment No. 226 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 July 25, 2003, as supplemented on December 5, 2003.

The amendment modifies TS 2.1.4, Reactor Coolant System (RCS) Leakage Limits, by (1) adding a requirement for no RCS pressure boundary leakage, (2) combining the existing RCS leakage limits into a format similar to the Improved Standard TS (ISTS), and (3) replacing the existing basis associated with this TS with a basis similar in format and content to the ISTS.

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, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-285

Enclosures:

1. Amendment No. 226 to DPR-40

2. Safety Evaluation cc w/encls: See next page DISTRIBUTION: See attached page TS: ML041390122 NRR-100 PKG: ML041390280 ACCESSION NO.: ML041320667 NRR-058 OFFICE PDIV-2/PM PDIV-1/LA OGC PDIV-2/SC TSS:IRUB/SC NAME AWang:mp DBaxley TSmith SDembek TBoyce DATE 4/28/04 4/27/04 5/3/04 5/5/04 4/2/04 DOCUMENT NAME: C:\ORPCheckout\FileNET\ML041320667.wpd OFFICIAL RECORD COPY

Ft. Calhoun Station, Unit 1 cc:

Winston & Strawn Mr. Daniel K. McGhee ATTN: James R. Curtiss, Esq. Bureau of Radiological Health 1400 L Street, N.W. Iowa Department of Public Health Washington, DC 20005-3502 401 SW 7th Street, Suite D Des Moines, IA 50309 Chairman Washington County Board of Supervisors Mr. Richard P. Clemens P.O. Box 466 Division Manager - Nuclear Assessments Blair, NE 68008 Omaha Public Power District Fort Calhoun Station Mr. John Kramer, Resident Inspector P.O. Box 550 U.S. Nuclear Regulatory Commission Fort Calhoun, NE 68023-0550 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 Cententiall 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. John B. Herman Manager - Nuclear Licensing Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.

P.O. Box 550 Fort Calhoun, NE 68023-0550

FORT CALHOUN STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT RE: REACTOR COOLANT SYSTEM LEAKAGE LIMITS (TAC NO. MC0194)

Dated: May 7, 2004 DISTRIBUTION:

PUBLIC PDIV-2 Reading File RidsNrrDlpmLpdiv (HBerkow)

RidsNrrPMAWang RidsNrrLAEPeyton RidsNrrLADBaxley RidsOgcRp RidsAcrsAcnwMailCenter RidsRgn4MailCenter (CJohnson)

SDembek TBoyce GHill (2)

KKavanagh

OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 226 License No. DPR-40

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by the Omaha Public Power District (the licensee) dated July 25, 2003, as supplemented by letter dated December 5, 2003, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions 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. 226, 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 90 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Stephen Dembek, Chief, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: May 7, 2004

ATTACHMENT TO LICENSE AMENDMENT NO. 226 TO 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 INSERT Definitions - Page 8 Definitions - Page 8 2.1 - Page 13 2.1 - Page 13 2.1 - Page 14 2.1 - Page 14 2.1 - Page 15 2.1 - Page 15

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 226 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-40 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. 1 DOCKET NO. 50-285

1.0 INTRODUCTION

By letter dated July 25, 2003 (Reference 1), as supplemented by letter dated December 5, 2003 (Reference 2), Omaha Public Power District (OPPD or the licensee) submitted a request to revise Technical Specification (TS) 2.1.4, Reactor Coolant System (RCS) Leakage Limits, for the Fort Calhoun Station, Unit 1 (FCS). Specifically, the proposed changes add a requirement for no RCS pressure boundary leakage, combine the existing RCS leakage limits into a format similar to the Improved Standard TS (ISTS) NUREG-1432, Revision 2, Standard Technical Specifications - Combustion Engineering Plants, and replace the existing bases associated with this TS with a bases similar in format and content of the ISTS. The licensee stated that the proposed change will assure that the design criteria of no RCS pressure boundary leakage is maintained as recommended by the Nuclear Regulatory Commission (Commission or NRC)

Davis Besse Lessons Learned Report (Reference 3). In addition, the proposed changes will provide clearer identification of allowed leakage rates, allowed outage time or time to restore operability, and required actions for various conditions of inoperability. The licensee is also proposing to add a definition of "Leakage" to the "Definitions" section of the FCS TS.

The supplemental letter dated December 5, 2003, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register on August 19, 2003 (68 FR 49818).

2.0 REGULATORY EVALUATION

The Commissions regulatory requirements related to the content of TS are set forth in Section 50.36 of Title 10 of the Code of Federal Regulations (10 CFR). This regulation requires that the TS include items in five specific categories. These categories include: (1) safety limits, limiting safety system settings and limiting control settings, (2) limiting conditions for operation (LCOs), (3) surveillance requirements (SRs), (4) design features, and (5) administrative controls. However, the regulation does not specify the particular TS to be included in a plants license.

Additionally, 10 CFR 50.36(c)(2)(ii) sets forth four criteria to be used in determining whether a LCO is required to be included in the TS. These criteria are as follows:

1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

3. A structure, system, or component that is part of the primary success path which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

4. A structure system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

Existing LCOs and related surveillances included as TS requirements which satisfy any of the criteria stated above must be retained in the TS. RCS leakage limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

3.0 TECHNICAL EVALUATION

The NRC staff has reviewed the licensees regulatory and technical analyses in support of its proposed license amendment which are described in Sections 5.0 and 6.0 of the licensees submittal. The detailed evaluation below will support the conclusion 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 Commissions 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.

3.1 Addition of Definition of LEAKAGE The licensee has proposed to add the following definition of LEAKAGE to the DEFINITIONS section of the FCS TS:

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal leakoff), that is captured and conducted to collection systems or a sump or collecting tank,

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE, or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator (SG) to the Secondary System,

b. Unidentified LEAKAGE All LEAKAGE (except RCP seal leakoff) that is not identified LEAKAGE, and

c. Pressure Boundary LEAKAGE LEAKAGE (except SG LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

The FCS TS currently do not contain a definition of LEAKAGE because TS 2.1.4 currently does not differentiate between unidentified or identified LEAKAGE and does not contain a requirement for pressure boundary LEAKAGE, as discussed in Section 3.2. The proposed addition of the definition of LEAKAGE provides clarity since Identified LEAKAGE, Unidentified LEAKAGE, and Pressure Boundary LEAKAGE are terms that are proposed to be used in TS 2.1.4. The NRC staff considers the addition of the definition of LEAKAGE to be an administrative change which does not add or delete any requirements from the FCS TSs.

Therefore, the NRC staff finds the proposed addition of the definition of LEAKAGE to be acceptable. In addition, the proposed definition of LEAKAGE is consistent with the LEAKAGE definition in NUREG-1432, Revision 2.

3.2 TS 2.1.4, RCS Leakage Limits The FCS TS are custom specifications and are not consistent with the ISTS content and format.

As such, the LCO 2.1.4 has four specifications that must be met whenever the reactor coolant temperature (Tcold) is greater than 210 degrees Fahrenheit. The current TS 2.1.4 requires the following:

(1) If the reactor coolant system leakage exceeds 1 gpm and the source of the leakage is not identified within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the reactor shall be placed in the hot shutdown condition. If the source leakage exceeds 1 gpm and is not identified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be placed in the cold shutdown condition.

(2) If leakage exceeds 10 gpm, the reactor shall be placed in the hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the leakage exceeds 10 gpm for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor shall be placed in the cold shutdown condition.

(3) Primary to secondary leakage through the steam generator tubes shall be limited to 150 gallons per day per steam generator and 300 gallons per day total for both steam generators.

When primary to secondary leakage has been determined to be in excess of the limit, the leakage rate shall be reduced to within limits in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or the reactor shall be placed in the cold shutdown condition within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

(4) To determine leakage to the containment, a containment atmosphere radiation monitor (gaseous or particulate) or dew point instrument, and a containment sump level instrument must be operable.

a. With no containment sump level instrument operable, verify that a containment atmosphere radiation monitor is operable, and restore the containment sump level instrument to operable status within 30 days.

b. With no containment atmosphere radiation monitor and no dewpoint instrument operable, restore either a radiation monitor or dewpoint instrument to operable status within 30 days.

c. With only the dewpoint instrument operable, or with no operable instruments, enter Specification 2.0.1 immediately.

The licensee proposed to replace the content of TS 2.1.4 Specifications 1 through 3 in their entirety as stated below:

(1) RCS operational LEAKAGE shall be limited to:

a. No Pressure Boundary LEAKAGE,

b. 1 gpm unidentified LEAKAGE,

c. 10 gpm identified LEAKAGE,

d. 150 gallons per day primary to secondary LEAKAGE through any one SG [Steam Generator].

(2) If RCS LEAKAGE limits of (1), above, are not met for reasons other than Pressure Boundary LEAKAGE, then reduce LEAKAGE to meet limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

(3) If the Required Action and associated completion time of (2), above, is not met, OR Pressure Boundary LEAKAGE exists, then be in MODE 3, Hot Shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND be in MODE 4, Cold Shutdown, within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The revised TS 2.1.4 Specification (1) will clearly state the limits on RCS operational LEAKAGE. The leakage limits for Unidentified and Identified leakage are consistent with the current TS and therefore, the proposed format change is acceptable. The leakage limits for primary to secondary leakage through any one SG is also consistent with the current TS.

However, the licensee proposed to delete the leakage limit of 300 gallons per day for both SGs.

The staff finds this deletion to be acceptable because the leakage limit of 300 gallons per day for both SGs is redundant. In addition, the safety analysis assumes a 1 gallon per minute (gpm) primary to secondary leak as the initial condition. As such, the proposed TS leakage limit of 150 gallons per day primary to secondary leakage per SG is significantly less than the initial condition for the safety analysis. Therefore, the staff finds the proposed primary to secondary leakage limit per SG to be acceptable. The licensee has also proposed to add a requirement for No Pressure Boundary LEAKAGE in TS 2.1.4 Specification (1).

Pressure Boundary Leakage is indicative of material deterioration and could cause further deterioration, resulting in higher leakage. Since the purpose of this LCO is to protect the reactor coolant pressure boundary from degradation and the core from inadequate cooling, the NRC staff finds the addition of the requirement for No Pressure Boundary LEAKAGE to be a more restrictive change and acceptable.

The revised TS 2.1.4 Specification (2) provides a required condition, required action, and completion time for the condition of RCS LEAKAGE limits above those allowed in Specification (1), except for Pressure Boundary LEAKAGE. The proposed required condition, required action, and completion time are consistent with current TS 2.1.4 Specification 3 for primary to secondary leakage through the SG tubes and, therefore, is acceptable for meeting the LEAKAGE limits specified in Specification (1)d. The existing FCS specifications do not have a requirement to reduce identified and unidentified LEAKAGE to meet limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. For unidentified LEAKAGE, the NRC staff concludes that the required action to reduce LEAKAGE to meet limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is a more restrictive change and, therefore, acceptable for meeting the LEAKAGE limits specified in Specification (1)b. For identified LEAKAGE, the current Specification (2) requires that the reactor be placed in the hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if leakage exceeds 10 gpm. The proposed 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> completion time to reduce LEAKAGE within limits is within the current completion time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to place the reactor in hot shutdown and, therefore, is a more restrictive change. In addition, identified LEAKAGE is from known sources that do not interfere with the detection of unidentified LEAKAGE and is well within the capability of the RCS makeup system, therefore, the staff concludes that the proposed change is acceptable for meeting the LEAKAGE limits specified in Specification (1)c.

The revised TS 2.1.4 Specification (3) provides required conditions, required actions and completion times for the condition of the Required Action and completion times of Specification (2) not met OR Pressure Boundary LEAKAGE exists. The current specification does not have a requirement to place the reactor in hot shutdown if primary to secondary LEAKAGE is not reduced to meet the limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. In addition, the current specification allows up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to place the reactor in hot shutdown if unidentified and identified LEAKAGE exceed limits. In these conditions, the proposed requirement to place the reactor in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is a more restrictive change which would reduce LEAKAGE and prevent further degradation of the RCS. Therefore, the staff concludes that the proposed required action to place the reactor in hot shutdown within six hours is acceptable. The current specification allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for the reactor to be placed in cold shutdown if unidentified, identified or primary to secondary LEAKAGE exceeds limits. The licensee has proposed to allow 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to place the reactor in cold shutdown if unidentified, identified, or primary to secondary LEAKAGE exceeds limits. The proposed required actions for Specification (3) is an AND statement, therefore both actions of placing the reactor in hot shutdown and cold shutdown are required. These actions are more conservative than the current FCS requirements which do not contain the AND statement. Therefore, the staff concludes that the proposed change to place the reactor in cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> is acceptable.

As discussed above, the current specification does not have a requirement for Pressure Boundary LEAKAGE. Therefore, the proposed required actions and associated completion times for the condition that Pressure Boundary LEAKAGE exists is a more restrictive change which provides added protection against reactor coolant pressure boundary degradation and inadequate core cooling. As such, the staff concludes that the proposed change for Pressure Boundary LEAKAGE is acceptable.

The NRC staff notes that the proposed changes for Specifications (1), (2), and (3) are consistent to the content of Specification 3.4.13 of NUREG-1432, Revision 2. In addition, the licensee did not propose any changes to the associated SRs. The licensee stated that SRs for FCS are contained in Section 3 of the TSs and are related by association but not directly applicable to the LCO statements. Therefore, the licensee did not propose any changes to the

existing SRs with this proposed change. The staff evaluated the existing SRs for FCS and has determined that they are similar to those in NUREG-1432, Revision 2.

The NRC staff has reviewed the licensees application with the supporting documentation.

Based on its review, the NRC staff concludes that the proposed TS changes are acceptable because the proposed changes provide clarity, are generally more restrictive in nature, and meet the regulations specified in 10 CFR 50.36. Additionally, the NRC staff concludes that there is reasonable assurance that plant operation in this manner poses no undue risk to the health and safety of the public.

4.0 STATE 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.

5.0 ENVIRONMENTAL 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 (68 FR 49818 dated August 19, 2004). 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 environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

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.

7.0 REFERENCES

1. Ridenoure, R. T., OPPD, to USNRC, Fort Calhoun Station Unit No. 1 License Amendment Request, Reactor Coolant System Leakage Limits, July 25, 2003.

2. Ridenoure, R. T., OPPD, to USNRC, Response to NRC Request for Additional Information on Fort Calhoun Stations Reactor Coolant System Leakage Limits Amendment Request, December 5, 2003.

3. Howell, A. T., USNRC, to W. F. Kane, USNRC Degradation of the Davis Besse Nuclear Power Station Reactor Pressure Head Lessons-Learned Report, September 30, 2002.

Principal Contributor: K. Kavanagh Date: May 7, 2004