ML052020427

Issuance of License Amendment 169 Deletion or Revision of Licensing Conditions (Revision of TS Tables 5.5.9-2 and 5.5.9-3)

ML052020427
Person / Time
Site:	Callaway
Issue date:	10/25/2005
From:	Donohew J NRC/NRR/DLPM/LPD4
To:	Naslund C Union Electric Co
Taylor G, NRR/DLPM/LPD4
Shared Package
ML053000556	List: ML052020427 ML053000474
References
TAC MC5060
Download: ML052020427 (24)

Text

October 25, 2005 Mr. Charles D. Naslund Senior Vice President and Chief Nuclear Officer Union Electric Company Post Office Box 620 Fulton, MO 65251

SUBJECT:

CALLAWAY PLANT, UNIT 1 - ISSUANCE OF AMENDMENT RE: DELETION OR REVISION OF LICENSING CONDITIONS (TAC NO. MC5060)

Dear Mr. Naslund:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 169 to Facility Operating License No. NPF-30 for the Callaway Plant, Unit 1. The amendment consists of changes to the Operating License in response to your application dated October 27, 2004 (ULNRC-05070), as supplemented by letter dated June 17, 2005 (ULNRC-05162).

The amendment (1) deletes Conditions 2.C.(3), 2.C.(4), 2.C.(6) through 2.C.(14), Section 2.F, and Attachments 1 and 2, and (2) revises Conditions 2.C.(1) and 2.C.(5), to the facility operating license, to reflect completed requirements. In addition, the list of the attachments and appendices to the operating license is revised to reflect the deletion of Attachments 1 and 2. The proposed changes to Technical Specifications (TSs) Table 5.5.9-2, Steam Generator Tube Inspection, and Table 5.5.9-3, Steam Generator Repaired Tube Inspection, were also submitted in your application dated September 17, 2004 (ULNRC-05056), for the replacement steam generator project and were approved in Amendment No. 168, which was issued in our letter dated September 29, 2005, for TAC No. MC4437.

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/

Jack N. Donohew, Senior Project Manager, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-483

Enclosures:

1. Amendment No. 169 to NPF-30

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

October 25, 2005 Mr. Charles D. Naslund Senior Vice President and Chief Nuclear Officer Union Electric Company Post Office Box 620 Fulton, MO 65251

SUBJECT:

CALLAWAY PLANT, UNIT 1 - ISSUANCE OF AMENDMENT RE: DELETION OR REVISION OF LICENSING CONDITIONS (TAC NO. MC5060)

Dear Mr. Naslund:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 169 to Facility Operating License No. NPF-30 for the Callaway Plant, Unit 1. The amendment consists of changes to the Operating License in response to your application dated October 27, 2004 (ULNRC-05070), as supplemented by letter dated June 17, 2005 (ULNRC-05162).

The amendment (1) deletes Conditions 2.C.(3), 2.C.(4), 2.C.(6) through 2.C.(14), Section 2.F, and Attachments 1 and 2, and (2) revises Conditions 2.C.(1) and 2.C.(5), to the facility operating license, to reflect completed requirements. In addition, the list of the attachments and appendices to the operating license is revised to reflect the deletion of Attachments 1 and

2. The proposed changes to Technical Specifications (TSs) Table 5.5.9-2, Steam Generator Tube Inspection, and Table 5.5.9-3, Steam Generator Repaired Tube Inspection, were also submitted in your application dated September 17, 2004 (ULNRC-05056), for the replacement steam generator project and were approved in Amendment No. 168, which was issued in our letter dated September 29, 2005, for TAC No. MC4437.

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/

Jack N. Donohew, Senior Project Manager, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-483 DISTRIBUTION:

PUBLIC GHill (2)

Enclosures:

1. Amendment No. 169 to NPF-30 PDIV-2 Reading TBoyce

2. Safety Evaluation RidsNrrDlpmPdiv (HBerkow)

RidsNrrDlpmDpr RidsNrrDlpmPdiv2 cc w/encls: See next page RidsNrrPMJDonohew RidsNrrLALFeizollahi RidsOgcRp PKG.: ML053000556 RidsAcrsAcnwMailCenter TS: ML053000474 NRR-100 RidsRegion4MailCenter (BJones)

ACCESSION NO.: ML052020427(Letter) NRR-058 OFFICE PDIV-1/GE PDIV-2/PM PDIV-2/LA IROB-A/SC OGC PDIV-2/(A)SC NAME GTaylor JDonohew LFeizollahi TBoyce KWinsberg DCollins DATE 8/12/05 10/18/05 10/18/05 8/23/05 9/14/05 10/25/05 OFFICIAL RECORD COPY

UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT 1 DOCKET NO. 50-483 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 169 License No. NPF-30

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

A. The application for amendment by Union Electric Company (UE, the licensee) dated October 27, 2004, as supplemented by letter dated June 17, 2005, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's 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 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, the license is amended by changes to the Operating License as indicated in the attachment to this license amendment.

3. This amendment is effective as of its date of issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Daniel S. Collins, Acting Chief, Section 2 Project Directorate IV Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License Date of Issuance: October 25, 2005

ATTACHMENT TO LICENSE AMENDMENT NO. 169 FACILITY OPERATING LICENSE NO. NPF-30 DOCKET NO. 50-483 Replace the following pages of the Operating License with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT Operating License Operating License 3 3 4 4 5 5 6 6 7 ---

8 ---

9 ---

Attachment 1 Attachment 1 Attachment 2 Attachment 2

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 169 TO FACILITY OPERATING LICENSE NO. NPF-30 UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT 1 DOCKET NO. 50-483

1.0 INTRODUCTION

By application dated October 27, 2004 (Agencywide Document Access and Management System (ADAMS) Accession No. ML043140378), as supplemented by letter dated June 17, 2005 (ADAMS Accession No. ML052000083), Union Electric Company (UE or the licensee) requested changes to the Operating License (Facility Operating License No. NPF-30) for the Callaway Plant, Unit 1 (Callaway).

The proposed changes would (1) delete or revise certain license conditions from the operating license, and (2) revise Table 5.5.9-2, Steam Generator Tube Inspection, and Table 5.5.9-3, Steam Generator Repaired Tube Inspection, of the Technical Specifications (TSs). License Conditions 2.C.(3), 2.C.(4), 2.C.(6) through 2.C.(14), and Attachments 1 and 2 of the operating license are considered to have been completed and obsolete, or duplicate other license requirements, and are proposed to be deleted. License Conditions 2.C.(1) and 2.C.(5) are proposed to be revised to reflect completed items in these conditions. Section 2.F of the operating license is considered to duplicate the reporting requirements of Title 10 of the Code of Federal Regulations (10 CFR) Sections 50.72 and 50.73 and is proposed to be deleted. The reporting requirements in two Action Required columns of TS Tables 5.5.9-2 and 5.5.9-3 are also considered to duplicate the reporting requirements in 10 CFR 50.72 and 50.73 and are proposed to be deleted.

The proposed changes to Technical Specifications (TSs) Table 5.5.9-2, Steam Generator Tube Inspection, and Table 5.5.9-3, Steam Generator Repaired Tube Inspection, were also submitted in the licensee's application dated September 17, 2004, for the replacement steam generator project and were approved in Amendment No. 168, which was issued in NRC letter dated September 29, 2005.

The additional information provided in the supplemental letter dated June 17, 2005, did not expand the scope of the application as noticed and does not change the Nuclear Regulatory Commission (NRC or Commission) staffs original proposed no significant hazards consideration determination published in the Federal Register on December 7, 2004 (69 FR 70723).

2.0 REGULATORY EVALUATION

In accordance with 10 CFR 50.10, a nuclear power plant may not be operated without a license.

In that license, conditions on the operation of the nuclear power plant may be specified and the license requires that the operator must operate the plant in accordance with these conditions. If such a license condition has been met, then the licensee may request that the license condition be deleted from the license because the condition is no longer valid and no longer needs to be enforced by the license.

3.0 TECHNICAL EVALUATION

The licensee has proposed to delete or revise certain conditions of the operating license, which are discussed below:

and Appendix 1 Condition 2.C.(1) states that:

UE is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal (100% power) in accordance with the conditions specified herein and in Attachment 1 to this license. The preoperational tests, startup tests and other items identified in Attachment 1 to this license shall be completed as specified. Attachment 1 is hereby incorporated into this license. states the following:

This attachment identifies items which must be completed to the Commissions satisfaction in accordance with the operational modes as identified below.

A. The licensee shall implement Radiation/Chemical Technician refresher training within six months following fuel load.

B. The licensee shall install a permanent area monitor on the manipulator crane prior to the entering Mode 6 (refueling mode).

The licensee stated that all items of Attachment 1 have been completed and it proposed to delete Attachment 1 from its license and revise License Condition 2.C.(1) to reflect this deletion as follows: UE is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal (100% power) in accordance with the conditions specified herein.

In its supplemental letter dated June 17, 2005, the licensee stated that training required by Item A of Attachment 1 was implemented on November 27, 1984, according to the Callaway Commitment Tracking System. Callaway was issued its fuel loading and up to 5 percent operating license on June 12, 1984. Since this training was completed within the 6-month time frame, as required in Item A of Attachment 1, the NRC staff concludes that this item is completed.

By letter dated February 7, 1986, the licensee informed the NRC that it planned to install a permanent area radiation monitor on the manipulator crane prior to entry into Mode 6. The letter further stated that the monitor will be installed in accordance with an approved design and will be calibrated and operable prior to the beginning of reactor vessel stud detensioning. The supplemental letter of June 17, 2005, stated that the installation of a permanent area radiation monitor on the manipulator crane was completed on March 7, 1986, as referenced by the control room Shift Supervisor Daily log, SD-RE-41. Based on the licensees statement that the area radiation monitor was installed prior to Callaway entering Mode 6 on March 9, 1986, the NRC staff concludes that Item B of Attachment 1 to the Callaway operating license is completed. Therefore, the NRC staff concludes that all items of Attachment 1 have been completed and no longer need to be stated in the license. The NRC staff further concludes that the proposed deletion of Attachment 1 and the proposed revision of License Condition 2.C.(1),

to reflect the deletion of Attachment 1, in the operating license, as follows:

UE is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal (100% power) in accordance with the conditions specified herein.

meet 10 CFR 50.10 and are, therefore, acceptable.

3.2 Condition 2.C.(3) - Environmental Qualification (Section 3.11, SSER#3)

Condition 2.C.(3) states that:

(a) Prior to November 30, 1985, UE shall environmentally qualify all electrical equipment according to the provisions of 10 CFR 50.49.

(b) Prior to restart following the first refueling outage, UE shall have qualified the reactor vessel level instrumentation system high volume sensor.

The licensee stated that it notified the NRC that the electric equipment required to be qualified under 10 CFR 50.49 had been evaluated and determined to be qualified per the provisions of 10 CFR 50.49 in its letters dated November 29, 1985 and January 17, 1986. The NRC responded by letter dated March 17, 1986, stating that License Condition 2.C.(3)(a) had been fulfilled.

Section 50.49 of 10 CFR, Environmental qualification of electric equipment important to safety for nuclear power plants, requires that licensees conform to Regulatory Guide (RG) 1.97 (Revision 2), in relation to certain post-accident monitoring equipment. As stated in a letter dated April 10, 1985, the NRC staff found the licensees conformance to RG 1.97 (Revision 2) acceptable. By letter dated April 4, 1986, the licensee notified the NRC that the reactor vessel level instrumentation system (RVLIS) high volume sensor, as well as all other RVLIS components, had been evaluated and determined to be qualified. As a result, the licensee stated Condition 2.C.(3)(b) had been fulfilled. In a supplemental letter dated June 17, 2005, the licensee stated that it began restarting Callaway on April 18, 1986, following its completion of Callaways first refueling outage.

Based on the NRC staffs letter of March 17, 1986, which concluded that Condition 2.C.(3)(a) had been fulfilled, along with the licensees conformance to RG 1.97 (Revision 2) and completion of qualifying the RVLIS high volume sensor prior to restart following Callaways first refueling outage, as specified in Condition 2.C.(3)(b), the NRC staff concludes that the requirements of Condition 2.C.(3) have been met and the license condition no longer needs to be stated in the operating license. Based on this, the NRC staff further concludes that the proposed deletion of this condition (i.e., replace the license condition requirement with the phrase deleted per amendment) meets 10 CFR 50.10 and is, therefore, acceptable.

3.3 Condition 2.C.(4) - Surveillance of Hafnium Control Rods (Section 4.2.3.1(10) SER

[Safety Evaluation Report] and SSER#2)

Condition 2.C.(4) states that:

UE shall perform a visual inspection of a sample of hafnium control rods during one of the first five refueling outages. A summary of the results of these inspections shall be submitted to the NRC.

The licensee stated that it submitted, in its letter dated November 30, 1989, the results of the eddy current testing and visual inspections performed during the examination of the rod cluster control assemblies (RCCAs) during the second refueling outage in the fall of 1987. Further, the licensee communicated that after their hafnium control rod inspections, during Cycle-3 of operation, swelling of hafnium RCCAs was identified at other plants with Westinghouse supplied nuclear steam supply systems. Although swelling was not identified at Callaway during the second refueling outage eddy current inspections, the licensee chose to replace all hafnium RCCAs with Silver-Indium-Cadmium RCCAs during Callaways third refueling outage.

The use of Silver-Indium-Cadmium RCCAs was found to be acceptable by the NRC staff via license Amendment No. 41 issued to Callaway on February 14, 1989, and the licensee completed replacement of all hafnium RCCAs with Silver-Indium-Cadmium RCCAs in the spring of 1989.

Since the results of the visual inspection of hafnium control rods was submitted by the licensee in its letter of November 30, 1989, the NRC staff concludes that the licensee has met the requirements of Condition 2.C.(4) and the requirements in this license condition no longer need to be stated in the operating license. Based on this conclusion, the NRC staff concludes that the proposed deletion of the condition (i.e., replace the license condition requirement with the phrase deleted per amendment) meets 10 CFR 50.10 and is, therefore, acceptable.

3.4 Condition 2.C.(5) - Fire Protection (Section 9.5.1.7 SER and Section 9.5.1.8, SSER#3)

Condition 2.C.(5) states that:

(a) Within 60 days of acquisition of the 100% power data for thermal and dynamic testing, UE shall have operable the Halon systems in the north electrical penetration room (fire area A-18).

(b) Prior to restart following the first extended outage of known duration greater than two weeks occurring after February 15, 1985 or prior to

restart following the first refueling outage which ever occurs first, UE shall have completed the installation of the five new isolation switches and modification to the four existing isolation switches identified in the August 23, 1984 SNUPPS [standardized nuclear unit power plant system]

letter.

The licensee stated, in its letter dated February 21, 1985, that the Halon concentration test for the north electrical penetration room was completed on February 5, 1985 and the results were approved by the Onsite Review Committee on February 8, 1985. Since Callaway startup testing was completed on December 19, 1984, this testing was completed within 60 days of acquisition on the 100 percent power data for thermal and dynamic testing. Based on the licensees statements that the Halon system was operable by February 5, 1985, the NRC staff concludes Condition 2.C.(5)(a) has been completed.

In a letter dated February 20, 1986, the licensee informed the NRC that the switch installations and modifications were completed in compliance with License Condition 2.C.(5)(b). The installation of five new isolation switches and the modification of four existing isolation switches enhance the ability to shut the plant down with a postulated fire in the Main Control Room, as described in the August 23, 1984, SNUPPS letter. Based on the licensee's February 20, 1986, notification letter regarding License Condition 2.C.(5)(b), the NRC staff concludes that this condition has been satisfied.

The NRC staff has reviewed the actions taken by the licensee to satisfy Callaway License Conditions 2.C.(5)(a) and 2.C.(5)(b) and has found them to be satisfactory. Therefore, the NRC staff concludes that these two license conditions do not need to be stated in the operating license. Based on this, the NRC staff further concludes that the proposed deletion of these license conditions (i.e., replace the license condition requirements with the phrase deleted per amendment) meet 10 CFR 50.10 and are, therefore, acceptable.

3.5 Condition 2.C.(6) - Qualification of Personnel Condition 2.C.(6) states that:

(a) UE shall have on each shift operators who meet the requirements described in Attachment 2 [of the license].

(b) UE shall have a senior individual with previous operating experience on a commercial PWR [pressurized-water reactor] assigned to assist the Plant Manager as an advisor during the startup test program and for one year following full power operation. , which lists operating staff experience requirements, states the following:

UE shall have a licensed senior operator on each shift who has had at least six months of hot experience on a same type plant, including at least six weeks at power levels greater than 20% of full power, and who has had startup and shutdown experience. For those shifts where such an individual is not available

on the plant staff, an advisor shall be provided who has had at least four years of power plant experience, including two years of nuclear plant experience, and who has had at least one year of experience on shift as a licensed senior operator at a similar type facility. Use of advisors who were licensed only at the RO level will be evaluated on a case-by-case basis. Advisors shall be trained on plant procedures, technical specifications and plant systems, and shall be examined on these topics at a level sufficient to assure familiarity with the plant.

For each shift, the remainder of the shift crew shall be trained in the role of the advisors. The training of the advisors and remainder of the shift crew shall be completed prior to exceeding 5% power. Prior to exceeding 5% power, UE shall certify to the NRC the names of the advisors who have been examined and have been determined to be competent to provide advice to the operating shifts.

These advisors shall be retained until the experience levels identified in the first sentence above have been achieved. The NRC shall be notified at least 30 days prior to the date UE proposes to release the advisors from further service.

The licensee stated that it notified the NRC of its intent to release the shift advisors from further service on May 31, 1985, in its letter dated April 15, 1985. It stated further that TS 5.3 and Final Safety Analysis Report (FSAR) Chapter 13 provide the shift staffing and qualification requirements for operations personnel. The regulations in 10 CFR 50.54(m) provide the minimum requirements for on-site staffing of nuclear power plants for licensed operators and senior operators.

In its letter of April 15, 1985, the licensee stated its response to the Advisory Committee on the Reactor Safeguards Report on Callaway that included a related commitment to have shift advisors for approximately one year. This one-year period was to begin approximately one month prior to fuel load and to include operation from initial criticality until the attainment of nominal full power. Callaway shift advisors were assigned to shift on April 22, 1984. The Callaway Plant attained initial criticality on October 2, 1984, and the plant was declared fully operational on December 19, 1984. In its letter of May 16, 1984, the licensee submitted the names and resumes of shift advisors who had been examined and determined competent to provide advice to the operating shifts. The licensee further stated, in its April 15, 1985 letter, that licensed personnel have accumulated sufficient hot experience to alleviate the need for shift advisors. Also, the letter notifying the NRC that the licensee would be releasing the shift advisors from further service was sent at least 30 days prior to the date they were released, which was required by Condition 2.C.(6)(a).

Based on the above discussion, the NRC staff concludes that Attachment 2 of the Callaway license has been completed and the proposed deletion of Attachment 2 for the license is acceptable. Also, License Condition 2.C.(6)(a) is no longer needed to allow advisors to assist the licensed operators and the required notification to the NRC of the change has been met.

Based on these findings, the NRC staff concludes that Condition 2.C.(6)(a) has been satisfied and does not need to be stated in the operating license.

In its letter of May 16, 1984, the licensee submitted the resume of the Advisor to the Plant Manager, who became the Advisor to the Manager for Callaway in February 1982. License Amendment No. 16 approved deletion of the Advisor to the Manager position because the required license condition 2.6.(C)(b) was satisfied on December 15, 1985, one year after full

power operation at Callaway. Since a qualified Advisor to the Manager was present during the startup test program and for one year following full power operation, the NRC staff concludes that Condition 2.C.(6)(b) has been satisfied and does not need to be stated in the operating license.

Based on the above, the NRC staff has reviewed the actions regarding License Condition 2.C.(6) taken by the licensee and has found that they have been satisfied and no longer need to be stated in the operating license. Based on this, the NRC staff further concludes that the proposed deletion of Condition 2.C.(6) (i.e., replace the license condition requirements with the phrase deleted per amendment) meets 10 CFR 50.10 and is, therefore, acceptable.

3.6 Condition 2.C.(7) - NUREG-0737 Conditions Condition 2.C.(7) states that:

UE shall complete the following conditions to the satisfaction of the NRC. These conditions reference the appropriate items in Section 22.2, TMI Action Plan Requirements for Applicants for Operating Licenses, in the Safety Evaluation Report and Supplements 1, 2, 3 and 4 NUREG-0830.

(a) Detailed Control Room Design Review (I.D.1, SSER #4)

Prior to May 1, 1985, UE shall submit for review and approval by the NRC staff, the results of the function and task analysis. For those Human Engineering Discrepancies (HEDs) identified by this analysis that require correction, the submittal shall include the proposed correction and implementation schedule and for those HEDs for which no planned correction is proposed, a basis for that determination shall be documented.

(b) Emergency Response Capabilities (Generic Letter 82-33, Supplement 1 to NUREG-0737)

Prior to restart following the first refueling outage, UE shall have a fully functional Technical Support Center and Emergency Operations Facility and a fully operable Emergency Response Facilities Information System (ERFIS).

(c) Regulatory Guide 1.97 (Section 7.5.2.3, SSER #3)

Prior to restart following the first refueling outage, UE shall have installed and operable the following instrumentation.

1) Source range instrumentation qualified to post-accident conditions

2) Reactor vessel water level instrumentation

3) Subcooling monitors

4) Radiation monitors for releases from steam generator safety/relief valves or atmospheric dump valves, and

5) Auxiliary feedwater pump turbine exhaust monitor On April 26, 1985, the licensee submitted the Final Report for the Task Analysis for SNUPPS Detailed Control Room Design Review (DCRDR), and in so doing, it considered all human

factor issues addressed and License Condition 2.C.(7)(a) complete. By letter dated August 27, 1985, the NRC staff stated that they had found that the requirements of License Condition 2.C.(7)(a) of the full power license satisfied with respect to the DCRDR, but the task analysis results were still under review as they relate to the upgrade of emergency operating procedures. The NRC staff completed its review and comparison of the Callaway emergency operating procedures against the Westinghouse Emergency Response Guidelines and found Callaways operating procedures to be acceptable. The NRC transmitted its finding to the licensee in a letter dated March 4, 1987, which stated that the NRC staff had completed its review of the Callaway emergency operating procedures and found them acceptable. The letter went on to say that, based upon this review, the NRC staff concluded that License Condition 2.C.(7)(a) had been satisfied. Based on this conclusion, the NRC staff concludes that the proposed deletion of Item a of Condition 2.C.(7) is acceptable.

For the emergency response capabilities in (b) above, the licensee stated its commitment to have the emergency response facilities and information systems of this condition completed prior to restart following the first refueling outage, in its letter dated April 8, 1986. The letter further stated that, in order to assess the operability of the emergency response facilities for Callaway, an Emergency Response Facility Assessment Program had been established and with this program, the licensee had determined that the subject facilities and information systems were fully functional. However, training of emergency response personnel on the ERFIS was being conducted and would be completed on April 9, 1986. Callaway began restart from the first refueling outage on April 18, 1986. Based on the information provided in the April 8, 1986, letter stating that the Technical Support Center, the Emergency Operations Facility, and information systems were determined to be fully functional, which was prior to restart following the first refueling outage and the licensees commitment to train appropriate response personnel by April 9, 1986, the NRC staff concludes that the licensee has satisfied License Condition 2.C.(7)(b), and that the proposed deletion of Item b of Condition 2.C.(7) is acceptable.

For the RG 1.97 requirement above, the licensee informed the NRC, in its letter dated April 7, 1986, that the instruments required to be operable via License Condition 2.C.(7)(c) were ready to be declared operable at the time of startup for Callaway Cycle 2. The letter went on to say that the instruments had been incorporated into the appropriate procedures and the operating staff had been trained as to their proper use. Callaway began restart from the first refueling outage on April 18, 1986. Based on statements from the licensee that the instruments required to be operable by April 18, 1986, were operable on April 7, 1986, the NRC staff concludes that License Condition 2.C.(7)(c) has been completed and that the proposed deletion of Item c of Condition 2.C.(7) is acceptable.

The NRC staff concludes that the licensee has successfully completed all conditions related to the Detailed Control Room Design Review (Item a), Emergency Response Capabilities (Item b),

and Regulatory Guide 1.97 (Item c). Therefore, all items of this condition have been completed and no longer need to be stated in the operating license. Based on this, the NRC staff concludes that the proposed deletion of Condition 2.C.(7) (i.e., replace the license condition requirements with the phrase deleted per amendment) meets 10 CFR 50.10 and is, therefore, acceptable.

3.7 Condition 2.C.(8) - Post-Fuel-Loading Initial Test Program Condition 2.C.(8) states that:

UE shall conduct the post-fuel-loading initial test program described in Chapter 14 of the FSAR, as amended, without making any major modifications unless such modifications have prior NRC approval. Major Modifications are defined as:

(a) elimination of any safety-related test*

(b) modification of objectives, test method, or acceptance criteria for any safety-related test (c) performance of any safety-related test at a power level different from that stated in the FSAR by more than 5 percent of rated power (d) failure to satisfactorily complete the entire initial start-up test program by the time core burn up equals 120 effective full power days (e) deviation from initial test program administrative procedures or quality assurance controls described in the FSAR (f) delays in test program in excess of 30 days (14 days if power level exceeds 50 percent), concurrent with power operation. If continued power operation is desired during a delay, the licensee shall provide justification that adequate testing has been performed and evaluated to demonstrate that the facility can be operated at the planned power level with reasonable assurance that the health and safety of the public will not be endangered.

• Safety-related tests are those tests which verify the design, construction, and operation of safety-related structures, and equipment.

The licensee stated that this condition is obsolete since the Initial Test Program is complete and the unit is currently in Operating Cycle 14. The NRC staff received the licensee's startup report for Callaway on March 19, 1985, which contained the methods and results of the Initial Test Program. Because Callaway is currently in operating cycle 14 and the Initial Test Program, as described in FSAR Chapter 14, is complete with the startup tests prior to the first operating cycle, the NRC staff concludes that Condition 2.C.(8) regarding the post-fuel-loading initial test program has been met and no longer needs to be stated in the operating license. Based on this, the NRC staff further concludes that the proposed deletion of Condition 2.C.(8) (i.e.,

replace the license condition requirements with the phrase deleted per amendment) meets 10 CFR 50.10 and is, therefore, acceptable.

3.8 Condition 2.C.(9) - Inservice Inspection Program (Section 5.2.4 and 6.6, SER)

Condition 2.C.(9) states that:

Within nine months of the date of this license, UE shall submit for staff review and approval, the inservice inspection program which conforms to the ASME

[American Society of Mechanical Engineers] Code [Boiler and Pressure Vessel Code] in effect 12 months prior to the date of issuance of this license.

The licensee submitted its First 10-Year Interval Inservice Inspection (ISI) program plan (including relief requests) by letter dated July 16, 1985, as supplemented by letters dated November 7, 1986, March 3, April 2, 1987, and April 11, 1988. An NRC letter to the licensee dated December 14, 1988, stated that the NRC staff had evaluated the programs compliance with the requirements of the 1980 Edition through Winter 1981 Addenda of Section XI of the ASME Code, 10 CFR 50.55a, and commitments made prior to granting the facilitys operating license. The letter also stated that the NRC staff determined that certain Section XI Code requirements were impractical to perform at Callaway, and relief from those requirements may be granted as requested. The letter concluded by stating that the NRC staff found Callaway's first 10-year ISI program plan acceptable. The NRC staff concludes, based on the above discussion and receipt of the initial submittal on July 16, 1985, that Condition 2.C.(9) has been met and no longer needs to be stated in the operating license. Based on this, the NRC staff further concludes that the proposed deletion of Condition 2.C.(9) (i.e., replace the license condition requirements with the phrase deleted per amendment) meets 10 CFR 50.10 and is, therefore, acceptable.

3.9 Condition 2.C.(10) - Emergency Planning Condition 2.C.(10) states that:

In the event that the NRC finds that the lack of progress in completion of the procedures in the Federal Emergency Management Agencys Final rule, 44 CFR Part 350, is an indication that a major substantive problem exists in achieving or maintaining an adequate state of emergency preparedness, the provisions of 10 CFR Section 50.54(s)(2) will apply.

The licensee stated that this condition duplicates the requirements of 10 CFR 50.54(s) that are applicable to Callaway and are enforceable; therefore, deleting this license condition would not reduce any requirements on the plant.

The NRC staff received the finding of the Federal Emergency Management Agency (FEMA) on 44 CFR 350 for the Missouri State and local emergency plans for Callaway in a letter dated July 9, 1985. The State and local emergency plans were found to be adequate to protect the health and safety of the public in that there is reasonable assurance that the appropriate protective measures can be taken in the event of a radiological emergency. However, this approval was conditional on FEMAs verification of the Alert and Notification (AN) system in accordance with the criteria of Appendix 3 of NUREG-0654/FEMA-REP-1, Revision 1, and the Standard Guide for the Evaluation of Alert and Notification Systems for Nuclear Power Plants (FEMA-43). In FEMA's letter to the licensee dated May 18, 1987, FEMA determined that the AN system installed around Callaway satisfied the criteria of NUREG-0654/FEMA-REP-1, Revision 1 and FEMA-43, and that there is reasonable assurance that the system is adequate to promptly alert and notify the public in the event of a radiological emergency at the Callaway site. Therefore, FEMA found Callaways conformance with 44 CFR 350 acceptable.

The regulations in 10 CFR 50.54(s)(2)(ii) and 50.54(s)(3) state the following, respectively:

If after April 1, 1981, the NRC finds that the state of emergency preparedness does not provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency ... and if the deficiencies ... are not corrected within four months of that finding, the Commission will determine whether the reactor shall be shut down until such deficiencies are remedied or whether other enforcement action is appropriate.

The NRC will base its finding on a review of the FEMA findings and determinations as to whether State and local emergency plans are adequate and capable of being implemented, and on the NRC assessment as to whether the licensees emergency plans are adequate and capable of being implemented.

Based on the above, the NRC staff concludes that, because Condition 2.C.(10) duplicates the regulations in 10 CFR 50.54(s) and the proposed deletion of the license condition would not remove any requirements from the plant, Condition 2.C.(10) does not need to be stated in the operating license. Based on this, the NRC staff further concludes that the proposed deletion of Condition 2.C.(10) (i.e., replace the license condition requirements with the phrase deleted per amendment) meets 10 CFR 50.10 and is, therefore, acceptable.

3.10 Condition 2.C.(11) - Steam Generator Tube Rupture (Section 15.4.4, SSER#3)

Condition 2.C.(11) states that:

Prior to restart following the first refueling outage, UE shall submit for NRC review and approval an analysis which demonstrates that the steam generator single-tube rupture (SGTR) analysis presented in the FSAR is the most severe case with respect to the release of fission products and calculated doses.

Consistent with the analytical assumptions, the licensee shall propose all necessary changes to Appendix A [of the TSs] to this license.

The licensee stated that it provided a report which demonstrated that the SGTR analysis presented in the FSAR is the most severe case with respect to the release of fission products and calculated doses in its letter of January 8, 1986. The licensee also submitted a license amendment request, in its letter of January 14, 1986, to incorporate a limiting condition for operation and surveillance requirements into the TSs for the steam generator atmospheric relief valves to assure the availability of mitigating equipment assumed in the SGTR analysis. The TS amendment request was approved in Amendment No. 45, dated May 16, 1989. The NRC Safety Evaluation Report for the Callaway SGTR was issued on August 6, 1990 and concluded that the licensee could successfully mitigate a design-basis SGTR accident as shown in its accident analysis. The SGTR analysis is described in Chapter 15 of the FSAR, and any changes to the analysis would be reviewed in accordance with the criteria in 10 CFR 50.59.

The licensee concluded that the requirements of the license condition have been met, and therefore, the license condition is obsolete and can be deleted.

In Amendment No. 45, the NRC staff stated that the SGTR analysis was submitted in the licensees letter dated January 8, 1986, and supplemented with additional letters. The analysis

takes credit for the operation of Atmospheric Steam Dumps (ASDs) to mitigate the consequences of an SGTR accident. Since the ASDs have not previously been relied upon to mitigate postulated accidents and transients, there were no requirements relating to operability of the ASDs in the Callaway TSs. Therefore, by letter dated January 14, 1986, the licensee submitted proposed TSs that require ASDs operability.

By letter dated March 21, 1989, the NRC staff provided the results of the review of the licensees submittal of January 14, 1986. In summary, the NRC staff did not find the proposed TSs to be acceptable. Areas of concern included the time interval allowed for inoperability of a single ASD and operability with an ASD isolated (via closure of the upsteam block valve) due to excessive leakage. By letter dated April 14, 1989, the licensee responded to the NRC staff position with revised TSs for the ASDs. Based on the application, as amended on May 16, 1989, TSs on the ASDs for the SGTR accident were approved.

In the safety evaluation for the SGTR analysis, dated August 6, 1990, the NRC staff agreed that the licensee could successfully mitigate a design-basis SGTR accident. Additionally, the NRC staff reviewed the licensee's responses regarding operator action times during an SGTR, and concluded that the licensee had satisfactorily verified the times assumed in the SGTR analysis for Callaway. The NRC staff concluded that the issuance of Amendment No. 45 and the safety evaluation dated August 6, 1990, completed the NRC staffs action on Condition 2.C.(11) and the condition had been satisfied. Therefore, based on the NRC staffs letter of August 6, 1990, the requirements of Condition 2.C.(11) regarding SGTR analysis have been met and do not need to be stated in the operating license. Based on this, the NRC staff further concludes that the proposed deletion of Condition 2.C.(11) (i.e., replace the license condition requirements with the phrase deleted per amendment) meets 10 CFR 50.10 and is, therefore, acceptable.

3.11 Condition 2.C.(12) - Low Temperature Overpressure Protection (Section 15, SSER#3)

Condition 2.C.(12) states that:

By January 1, 1985, UE shall submit for NRC review and approval a description of equipment modifications to the residual heat removal system (RHRS) suction isolation valves and to closure circuitry which conform to the applicable staff requirements (SRP 5.2.2). Within one year of receiving NRC approval of the modifications, UE shall have the approved modifications installed. Alternately, by January 1, 1985, UE shall provide acceptable justification for reliance on administrative means alone to meet the staffs RHRS isolation requirements, or otherwise, propose changes to Appendix A to this license which remove reliance on the RHRS as a means of low temperature overpressure protection.

The licensee stated that it responded to License Condition 2.C.(12), regarding low-temperature overpressure protection (LTOP), in its letter dated December 28, 1984. This letter described plant modifications to be completed within one year of receiving NRC approval. The modification included adding an alarm circuit to reactor coolant system (RCS) and residual heat removal (RHR) system valves as a control room indication that the RHR system was properly isolated from the RCS when the plant returned to operating pressure following use of the RHR relief valves for LTOP. The NRC letter dated July 30, 1985, approved the proposed

modification to satisfy Condition 2.C.(12) and concluded by stating that the submittal requirement of Condition 2.C.(12) had been met. In its letter of May 7, 1986, the licensee notified the NRC that the modifications had been completed and the requirements of Condition 2.C.(12) had been met. Since the licensee notified the NRC that the equipment modifications had been completed within one year after the NRC approved the modifications, as stated in the condition, the NRC staff concludes that Condition 2.C.(12) has been satisfied and no longer needs to be stated in the operating license. Based on this, the NRC staff further concludes that the proposed deletion of Condition 2.C.(12) (i.e., replace the license condition requirements with the phrase deleted per amendment) meets 10 CFR 50.10 and is, therefore, acceptable.

3.12 Condition 2.C.(13) - LOCA [loss-of-coolant accident] Reanalysis (Section 15, SSER#3)

Condition 2.C.(13) states that:

Prior to restart following the first refueling outage, UE shall submit for NRC review and approval a reanalysis for the worst large break LOCA using an approved ECCS [emergency core cooling system] evaluation model. At this time that model is the 1981 Westinghouse model. A modified version of the 1981 model which includes the BART computer code may be used.

The licensee stated that it transmitted an application for a reload license amendment for Callaway Cycle 2 on November 15, 1985, to the NRC, which included a large break LOCA analysis based on the BASH model. In a subsequent letter dated January 28, 1986, the licensee submitted a new large break LOCA analysis based on the Westinghouse BART model, which was the most recent Westinghouse large break LOCA model approved by the NRC. The NRC staff approved the reload amendment (Amendment No. 15) and, in the SER for the amendment, the NRC staff concluded that the license condition requiring reanalysis of the worst large break LOCA was met. The licensee concluded that the requirements of the condition have been met, and, therefore the license condition is obsolete and should be deleted.

In its SER of April 8, 1986, the NRC staff concluded, based on its review of the licensees reload amendment request letter of November 15, 1985, along with its supplements, that the reanalysis of the worst large break LOCA for Callaway was acceptable and the corresponding license condition had been met. The NRC staffs conclusions were based on the following: (1) the licensees analysis was performed using methodologies and codes which have been previously approved by the NRC and which satisfy the criteria of Appendix K to 10 CFR Part 50 and (2) the results using the analysis are within the acceptance criteria of 10 CFR 50.46.

Therefore, based on the NRC staffs SER of April 8, 1986, the NRC staff concludes that the requirements of Condition 2.C.(13) have been met and do not need to be stated in the operating license. Based on this, the NRC staff further concludes that the proposed deletion of Condition 2.C.(13) (i.e., replace the license condition requirements with the phrase deleted per amendment) meets 10 CFR 50.10 and is, therefore, acceptable.

3.13 Condition 2.C.(14) - Generic Letter 83-28 Condition 2.C.(14) states that:

UE shall submit responses to and implement the requirements of Generic Letter 83-28 on a schedule which is consistent with that given in its May 21, 1984 letter.

The licensee provided a summary of the responses and the NRC review of the requirements of Generic Letter (GL) 83-28. The licensee listed the following items of the GL and the NRC letter accepting the licensees response to that item of the GL:

Item NRC Response Accepting Item Item 1.1 - Post Trip Review Program NRC letter dated May 7, 1985 Description and Procedure Item 1.2 - Post Trip Review Data and NRC letter dated July 24, 1986 Information Capability Items 2.1.1 and 2.1.2 - Equipment NRC letters dated July 21, 1986 and Classification and Vendor Interface (Reactor December 22, 1986 Trip System Components)

Item 2.2.1 - Equipment Classification NRC letter dated April 10, 1987 (Programs for all Safety-Related Components)

Item 2.2.2 - Vendor Interface (Programs for NRC letters dated April 10, 1987 and all Safety-Related Components) December 3, 1990 Items 3.1.1 and 3.1.2 - Post Maintenance NRC letters dated June 25, 1985 and Testing (Reactor Trip System Components) July 3, 1985 Item 3.1.3 - Post Maintenance Testing - NRC letter dated October 7, 1986 Changes to Test Requirements (Reactor Trip System Components)

Items 3.2.1 and 3.2.2 - Post Maintenance NRC letter dated June 25, 1985 Testing (all other safety-related components)

Item 3.2.3 - Post Maintenance Testing - NRC letter dated October 7, 1986 Changes to Test Requirements (all other safety-related components)

Item 4.1 - Reactor Trip System Reliability NRC letters dated June 25, 1985 and (Vendor-Related Modifications) July 3, 1985 Items 4.2.1 and 4.2.2 - Reactor Trip System NRC letter dated October 28, 1985 Reliability - Maintenance and Testing

Item NRC Response Accepting Item Items 4.2.3 and 4.2.4 - Reactor Trip System NRC letter dated October 7, 1992 Reliability - Live Cycle Testing of Reactor Trip Breakers Item 4.3 - Reactor Trip System Reliability NRC letter dated July 18, 1984 (Automatic Actuation of Shunt Trip NRC Amendment No. 19 (April 3, 1987)

Attachment for Westinghouse Plants) (Not March 3, as stated in the application)

NRC Amendment No. 34 (February 17, 1988)

Item 4.4 - Reactor Trip System Reliability Not Applicable because Callaway is not a (Improvements in Maintenance and Test B&W Plant.

Procedures for B&W Plants)

Item 4.5.1 - Reactor Trip System Reliability NRC letters dated June 25, 1985 and (System Functional Testing) July 3, 1985 Items 4.5.2 and 4.5.3 - Reactor Trip System NRC letter dated June 12, 1989 Reliability (On-line System Functional Testing)

For Item 1.1 of GL 83-28, the NRC staff concluded in its letter of May 7, 1985, that the Post-Trip Review Program and Procedures for Callaway were acceptable.

For Item 1.2 of GL 83-28, the NRC staff concluded in its letter of July 24, 1986, that the Post-Trip Review Data and Information Capability for Callaway were acceptable.

For Items 2.1.1 and 2.1.2 of GL 83-28, the NRC staff concluded in its letter of July 21, 1986, that it had reviewed the equipment classification for reactor trip system components at Callaway and found them to be acceptable, which completed the NRC staff's review of Item 2.1.1 of GL 83-28. The letter further stated that Item 2.1.2 of GL 83-28, Vendor Interface for Reactor Trip System Components, had not been resolved at that time. In a letter dated December 22, 1986, the NRC staff stated that Item 2.1.2 of GL 83-28 was acceptable for Callaway and the NRC staff's actions for Item 2.1 (Parts 1 and 2) were complete.

For Item 2.2.1 of GL 83-28, the NRC staff concluded in its letter of April 10, 1987, that the program for classifying safety-related components and controlling safety related activities for Callaway were acceptable.

For Item 2.2.2 of GL 83-28, the NRC staff concluded in its letter of April 10, 1987, that the program to ensure vendor information for safety-related components for Callaway were acceptable. On March 20, 1990 the NRC staff issued GL 90-03 relating to relaxation of GL 83-28 Item 2.2.2, allowing the Vendor Equipment Technical Information Program to meet the intent of Item 2.2.2 of GL 83-28. The licensee responded to GL 90-03 in its letter of September 21, 1990, and the NRC staff accepted the licensees response in its letter dated December 3, 1990, stating that the licensees compliance with GL 90-03 was acceptable as long as the licensee informed the NRC within 30 days of implementing its commitments. The licensee notified the NRC by its letter dated January 10, 1991, that the program had been implemented. Based on

the NRC staffs letter of April 10, 1987, and the licensees conformance with GL 90-03, the NRC staff concludes that Item 2.2.2 fo GL 83-28 for Callaway is complete.

For Items 3.1.1 and 3.1.2 of GL 83-28, the NRC staff concluded in its letter of June 25, 1985, that the Post Maintenance Testing for Reactor Trip System Components for Callaway was acceptable.

For Items 3.2.1 and 3.2.2 of GL 83-28, the NRC staff concluded in its letter of June 25, 1985, that the Post Maintenance Testing for all other safety-related components for Callaway was acceptable.

For Items 3.1.3 and 3.2.3 of GL 83-28, the NRC staff concluded in its letter of October 7, 1986, that the Post Maintenance Testing for Callaway was acceptable.

For Item 4.1 of GL 83-28, the NRC staff concluded in its letter of June 25, 1985, that the Reactor Trip System Reliability for Vendor-Related Modifications for Callaway were acceptable.

For Items 4.2.1 and 4.2.2 of GL 83-28, the NRC staff concluded in its letter of October 28, 1985, that the Maintenance and Trending Programs for Reactor Trip Breakers for Callaway were acceptable.

For Items 4.2.3 and 4.2.4 of GL 83-28, the NRC staff issued a GL supplement dated October 7, 1992, which informed the licensee that the actions of items 4.2.3 (life testing) and 4.2.4 (periodic replacement of breakers or components) of GL 83-28 were no longer needed.

For Item 4.3 of GL 83-28, the NRC staff concluded in its letter of July 18, 1984, that the shunt trip modifications relating to GL 83-28 were acceptable for implementation and that the licencee was required to submit proposed TSs once the modifications were implemented, which the licensee transmitted to the NRC staff in its January 29, 1985 letter. Subsequently, the NRC staff issued GL 85-09 on May 23, 1985, which provided guidance to all Westinghouse licensees for submitting proposed TSs regarding the modification requirements of Item 4.3 of GL 83-28.

By letter dated January 9, 1986, the licensee submitted proposed TSs per the guidance of GL 85-09 and the NRC staff approved the proposed TSs as Amendment No. 19, dated April 3, 1987, which completed all but one requirement of GL 85-09. On February 17, 1988, the NRC issued Amendment No. 34 to Callaway's operating license that completed the remaining TSs change requirement of GL 85-09. Therefore, the NRC staffs letter of July 18, 1984, Amendment No. 19, and Amendment No. 34 completed the licensees actions for Item 4.3 of GL 83-28 for Callaway.

Item 4.4 of GL 83-28, is not applicable since Callaway uses a 4-loop Westinghouse system.

For Item 4.5.1 of GL 83-28, the NRC staff concluded in its letter dated June 25, 1985, that the Reactor Trip System Reliability for System Functional Testing for Callaway was acceptable.

For Items the NRC staff concluded in its letter dated June 12, 1989, that the existing intervals for on-line functional testing at Callaway are consistent with achieving high reactor trip system availability, and therefore, the NRC staff considered Item 4.5.3 of GL 83-28 to be complete for Callaway. The letter also noted that the Callaway plant is designed to permit

on-line functional testing of the reactor trip system, including testing of the diverse trip features of the reactor trip breakers (undervoltage and shut trip attachments), and the NRC staff concluded that Item 4.5.2 of GL 83-28 is not applicable for Callaway. Therefore, the NRC staff letter of June 12, 1989, completed the licensee's actions for Items 4.5.2 and 4.5.3 of GL 83-28.

Based on the above evaluation on how the licensee has completed the items of GL 83-28 for Condition 2.C.(14), the NRC staff concludes that Condition 2.C.(14) has been satisfied and no longer needs to be stated in the operating license. Based on this, the NRC staff further concludes that the proposed deletion of Condition 2.C.(14) (i.e., replace the license condition requirements with the phrase deleted per amendment) meets 10 CFR 50.10 and is, therefore, acceptable.

3.14 Section 2.F Section 2.F states the following:

With the exception of 2.C.(2) UE shall report any violations of the requirements contained in Section 2.C, of this license within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Initial notification shall be made in accordance with the provisions of 10 CFR 50.72 with written follow-up in accordance with the procedures described in 10 CFR 50.73(b), (c), (d), and (e).

The licensee stated that the reporting requirements of Section 2.F of the license are adequately addressed by the reporting requirements identified in 10 CFR 50.72 and 10 CFR 50.73. The deviations from Maximum Power Level (Condition 2.C.(1)), Technical Specifications and Environmental Protection Plan (Condition 2.C.(2)), Fire Protection (Condition 2.C.(5)), and other license conditions (Condition 2.C.(15)) are adequately governed by the requirements of 10 CFR 50.72 and 50.73. The remaining license conditions of Condition 2.C.(3), Condition 2.C.(4), and Conditions 2.C.(6) through 2.C.(14) are proposed to be deleted. The licensee concluded, therefore, that Section 2.F should be deleted from the license. The licensee pointed out in its application that the NRC staff had previously approved such a deletion in Amendment Nos. 220 and 97 for Beaver Valley Power Station, Units 1 and 2, respectively, issued on March 26, 1999, and Amendment No. 141 for Wolf Creek Generating Station, issued on September 24, 2001.

The NRC staff agrees that the requirements for immediate notification with written follow-up of events at operating nuclear plants have been incorporated into the regulations of 10 CFR 50.72 and 50.73. Therefore, the NRC staff concludes that the requirements of Section 2.F are redundant to the requirements in these regulations and do not need to be stated in the operating license. Based on this, the NRC staff further concludes that the proposed deletion of Condition 2.F (i.e., replace the license condition requirements with the phrase deleted per amendment) meets 10 CFR 50.10 and is, therefore, acceptable.

3.15 Conclusion Based on the above evaluations, the NRC staff has concluded that the proposed changes to revise or delete conditions in the operating license are acceptable. Therefore, the NRC staff further concludes that the amendment is acceptable.

4.0 REGULATORY COMMITMENT In Attachment 6 to its application, the licensee provided the following regulatory commitment:

COMMITMENT DUE DATE/EVENT The proposed amendment will be implemented within 90 90 days following NRC approval days after approval.

Although the licensee stated in its application letter that it would implement the approved amendment within 60 days, the NRC staff concludes that an implementation period of 90 days for this amendment is reasonable and, therefore, concludes that it is acceptable.

5.0 STATE CONSULTATION

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

6.0 ENVIRONMENTAL CONSIDERATION

The amendment relates to changes in recordkeeping, reporting, or administrative procedures or requirements. 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 (69 FR 70723). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(10). 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.

7.0 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.

Principal Contributor: G. Taylor Date: October 25, 2005

Callaway Plant, Unit 1 cc:

Professional Nuclear Consulting, Inc. Mr. Rick A. Muench 19041 Raines Drive President and Chief Executive Officer Derwood, MD 20855 Wolf Creek Nuclear Operating Corporation P.O. Box 411 John ONeill, Esq. Burlington, KA 66839 Shaw, Pittman, Potts & Trowbridge 2300 N. Street, N.W. Mr. Dan I. Bolef, President Washington, D.C. 20037 Kay Drey, Representative Board of Directors Coalition for the Mr. Mark A. Reidmeyer, Regional Environment Regulatory Affairs Supervisor 6267 Delmar Boulevard Regulatory Affairs University City, MO 63130 AmerenUE P.O. Box 620 Mr. Lee Fritz, Presiding Commissioner Fulton, MO 65251 Callaway County Court House 10 East Fifth Street U.S. Nuclear Regulatory Commission Fulton, MO 65151 Resident Inspector Office 8201 NRC Road Mr. David E. Shafer Steedman, MO 65077-1302 Superintendent, Licensing Regulatory Affairs Mr. Les H. Kanuckel AmerenUE Manager, Quality Assurance P.O. Box 66149, MC 470 AmerenUE St. Louis, MO 63166-6149 P.O. Box 620 Fulton, MO 65251 Mr. Keith D. Young Manager, Regulatory Affairs Missouri Public Service Commission AmerenUE Governor Office Building P.O. Box 620 200 Madison Street Fulton, MO 65251 Jefferson City, MO 65102-0360 Mr. Scott Clardy, Director Regional Administrator, Region IV Section for Environmental Public Health U.S. Nuclear Regulatory Commission P.O. Box 570 611 Ryan Plaza Drive, Suite 400 Jefferson City, MO 65102-0570 Arlington, TX 76011-4005 Certrec Corporation Mr. Mike Wells, Deputy Director 4200 South Hulen, Suite 630 Missouri Department of Natural Resources Fort Worth, TX 76109 P.O. Box 176 Jefferson City, MO 65102 Director, Missouri State Emergency Management Agency P.O. Box 116 Jefferson City, MO 65102-0116