ML050330442

License Amendment 257, Modify License Condition 2.C.(2)(b) to Eliminate Main Steam Isolation Valve Closure Test for Extended Power Uprate

ML050330442
Person / Time
Site:	Duane Arnold
Issue date:	03/17/2005
From:	Spaulding D NRC/NRR/DLPM/LPD3
To:	Peifer M Nuclear Management Co
Beaulieu, David, NRR/DLPM, 415-3243
References
TAC MC2320
Download: ML050330442 (23)

Text

March 17, 2005 Mark A. Peifer Site Vice President Duane Arnold Energy Center Nuclear Management Company, LLC 3277 DAEC Road Palo, IA 52324-0351

SUBJECT:

DUANE ARNOLD ENERGY CENTER - ISSUANCE OF AMENDMENT RE: LICENSE AMENDMENT REQUEST TSCR-056, MODIFY LICENSE CONDITION 2.C.(2)(b) TO ELIMINATE MAIN STEAM ISOLATION VALVE CLOSURE TEST FOR EXTENDED POWER UPRATE (TAC NO. MC2320)

Dear Mr. Peifer:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 257 to Facility Operating License No. DPR-49 for the Duane Arnold Energy Center. This amendment consists of a change to the Operating License in response to your application dated February 27, 2004, as supplemented by letters dated August 9, 2004, and January 7, 2005.

The amendment modifies license condition 2.C.(2)(b) to remove the requirement to perform a full main steam isolation valve closure test associated with extended power uprate. In accordance with your request in letter dated January 7, 2005, licensee condition 2.C.(2)(b) to eliminate the requirement to perform a main generator load reject test is not included in this amendment and will be addressed by separate correspondence. Our review of this effort will now be performed under a separate TAC.

A copy of the Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely,

/RA/

Deirdre W. Spaulding, Project Manager, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-331

Enclosures:

1. Amendment No. 257 to License No. DPR-49

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

March 17, 2005 Mark A. Peifer Site Vice President Duane Arnold Energy Center Nuclear Management Company, LLC 3277 DAEC Road Palo, IA 52324-0351

SUBJECT:

DUANE ARNOLD ENERGY CENTER - ISSUANCE OF AMENDMENT RE: LICENSE AMENDMENT REQUEST TSCR-056, MODIFY LICENSE CONDITION 2.C.(2)(b) TO ELIMINATE MAIN STEAM ISOLATION VALVE CLOSURE TEST FOR EXTENDED POWER UPRATE (TAC NO. MC2320)

Dear Mr. Peifer:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 257 to Facility Operating License No. DPR-49 for the Duane Arnold Energy Center. This amendment consists of a change to the Operating License in response to your application dated February 27, 2004, as supplemented by letters dated August 9, 2004, and January 7, 2005.

The amendment modifies license condition 2.C.(2)(b) to remove the requirement to perform a full main steam isolation valve closure test associated with extended power uprate. In accordance with your request in letter dated January 7, 2005, licensee condition 2.C.(2)(b) to eliminate the requirement to perform a main generator load reject test is not included in this amendment and will be addressed by separate correspondence. Our review of this effort will now be performed under a separate TAC.

A copy of the Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely,

/RA/

Deirdre W. Spaulding, Project Manager, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-331

Enclosures:

1. Amendment No. 257 to License No. DPR-49

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

PUBLIC GHill (2)

THarris PDIII-1 R/F TBoyce DSpaulding WRuland GGrant DLPMDPR LRaghavan OGC FAkstulewicz MKotzalas ACRS DThatcher Amendment Accession No.: ML050330442 OFFICE PM:PD3-1 LA:PD3-1 SPSB-A/SC IPSB-A/SC SRXB-A/SC OGC SC:PD3-1 NAME DSpaulding THarris SJones - Not Needed DThatcher*

FAkstulewicz MHiggins LRaghavan DATE 03/11/05 03/11/05 12/21/04 Per Email 01/24/05 02/07/05 02/15/05 03/17/05

• ML0501902740 OFFICIAL RECORD COPY

Duane Arnold Energy Center cc:

Mr. John Paul Cowan Executive Vice President &

Chief Nuclear Officer Nuclear Management Company, LLC 700 First Street Hudson, MI 54016 John Bjorseth Plant Manager Duane Arnold Energy Center 3277 DAEC Road Palo, IA 52324 Steven R. Catron Manager, Regulatory Affairs Duane Arnold Energy Center 3277 DAEC Road Palo, IA 52324 U. S. Nuclear Regulatory Commission Resident Inspectors Office Rural Route #1 Palo, IA 52324 Regional Administrator, Region III U. S. Nuclear Regulatory Commission 2443 Warrenville Road, Suite 210 Lisle, IL 60532-4352 Jonathan Rogoff Vice President, Counsel & Secretary Nuclear Management Company, LLC 700 First Street Hudson, WI 54016 Bruce Lacy Nuclear Asset Manager Alliant Energy/Interstate Power and Light Company 3277 DAEC Road Palo, IA 52324 Daniel McGhee Utilities Division Iowa Department of Commerce Lucas Office Buildings, 5th floor Des Moines, IA 50319 Chairman, Linn County Board of Supervisors 930 1st Street SW Cedar Rapids, IA 52404 Craig G. Anderson Senior Vice President, Group Operations 700 First Street Hudson, WI 54016 November 2004

NUCLEAR MANAGEMENT COMPANY, LLC DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 257 License No. DPR-49

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

A.

The application for amendment by Nuclear Management Company, LLC (NMC) dated February 27, 2004, as supplemented by letters dated August 9, 2004, and January 7, 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 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 paragraph 2.C.(2)(b) of Facility Operating License No. DPR-49 is hereby amended to read as follows:

(b)

The licensee will perform the generator load reject transient test required by the General Electric Licensing Topical Report for Extended Power Uprate (NEDC-32424P-A) - ELTR-1, including the allowances described in Section L.2.4 (2) of ELTR-1 regarding credit for unplanned plant transient events, using the thermal power level (1658 MWt) to establish the ELTR-1 power level limit.

The testing shall be performed at an initiating power level greater than the steady-state operation power level exceeding the ELTR-1 power level limit for the generator load reject transient.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

L. Raghavan, Chief, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Change to the Operating License Date of Issuance: March 17, 2005

ATTACHMENT TO LICENSE AMENDMENT NO. 257 FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 Replace the following page of the Facility Operating License DPR-49 with the attached revised page as indicated. The revised page is identified by order number and contains marginal lines indicating the area of change.

Remove Page Insert Page 4

4 (a)

For Surveillance Requirements (SRs) whose acceptance criteria are modified, either directly or indirectly, by the increase in authorized maximum power level in 2.C.(1) above, in accordance with Amendment No. 243 to Facility Operating License DPR-49, those SRs are not required to be performed until their next scheduled performance, which is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment No. 243.

(b)

The licensee will perform the generator load reject transient test l

required by the General Electric Licensing Topical Report for Extended Power Uprate (NEDC-32424P-A) - ELTR-1, including the allowances described in Section L.2.4 (2) of ELTR-1 regarding credit for unplanned plant transient events, using the thermal power level (1658 MWt) to establish the ELTR-1 power level limit. The testing l

shall be performed at an initiating power level greater than the steady-state operation power level exceeding the ELTR-1 power level l

limit for the generator load reject transient.

l (3) Fire Protection NMC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the Duane Arnold Energy Center and as approved in the SER dated June 1, 1978, and Supplement dated February 10, 1981, subject to the following provision:

NMC may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

(4) The licensee is authorized to operate the Duane Arnold Energy Center following installation of modified safe-ends on the eight primary recirculation system inlet lines which are described in the licensee letter dated July 31, 1978, and supplemented by letter dated December 8, 1978.

(5) Physical Protection NMC shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "Nuclear Management Company Duane Arnold Energy Center Physical Security Plan, Revision 0" submitted by letter dated October 18, as supplemented by letter dated October 21, 2004.

Amendment No. 43, 47, 50, 63, 65, 74, 112, 152, 190, 198, 214, 223, 232, 243, 257 Revised by Letter Dated October 28, 2004 Revised by letter dated December 10, 2004 l

Revised by letter dated March 17, 2005 l

l

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 257 TO FACILITY OPERATING LICENSE NO. DPR-49 NUCLEAR MANAGEMENT COMPANY, LLC DUANE ARNOLD ENERGY CENTER DOCKET NO. 50-331

1.0 INTRODUCTION

By application dated February 27, 2004, as supplemented by letters dated August 9, 2004, and January 7, 2005, the Nuclear Management Company, LLC (NMC or the licensee), requested a change to Facility Operating License No. DPR-49 for the Duane Arnold Energy Center (DAEC).

The proposed change was to remove license condition 2.C.(2)(b) which requires that two specific large transient tests (LTTs) be performed at specified reactor thermal power levels, as part of power ascension testing for the extended power uprate (EPU) project at the DAEC. In a letter dated February 27, 2004, NMC requested approval of this change prior to March 1, 2005, as modifications were planned for the upcoming refuel outage at the DAEC which will allow the reactor power level to reach the license condition for performing the first of the two LTTs, the full main steamline isolation valve (MSlV) closure test. However, these planned modifications will not allow the reactor to achieve the thermal power level required to invoke the second of the two LTTs required by the license condition, namely the main generator load reject test.

Given the staggered nature of the plant modifications in the DAEC EPU project, NMCs letter dated January 7, 2005, requested that the U. S. Nuclear Regulatory Commission (NRC) to issue separate license amendments, one for each of the two LTTs.

The supplemental letters contained clarifying information and did not change the initial no significant hazards consideration determination and did not expand the scope of the original Federal Register notice.

The NRC staff reviewed the licensees submittals and prepared this safety evaluation (SE) that addresses the MSIV closure test provision of the DAEC Operating License. The main generator load reject test provision will be addressed in separate correspondence.

DAEC provided supplemental information concerning the elimination of license condition 2.C.(2)(b) for performance of large transient tests for EPU in a letter dated August 9, 2004, in response to an NRC staff request for additional information (RAI). In addition, the NRC staff reviewed the relevant portions of the documents listed in Section 3 of this SE. NRC staff guidance for reviewing EPU test programs is described in NUREG-0800, Standard Review Plan (SRP) 14.2.1, Generic Guidelines for EPU Testing Programs, and provides reasonable assurance that the proposed testing program verifies those plant structures, systems, and components (SSCs) that are affected by the proposed power uprate will perform satisfactorily in service at the proposed power uprate level. The NRC staff review focused on the licensee adequately addressing the applicable portions of the guidance described in SRP 14.2.1 related to LTT.

In a letter dated November 6, 2001, the NRC issued Amendment No. 243 that approved the EPU for DAEC. This amendment consisted of changes to the operating license and Technical Specifications (TSs) to allow an increase in the maximum power level at DAEC from 1658 Megawatts thermal (MWt) to 1912 MWt, representing a power increase of 15.3 percent.

Amendment No. 243 also added license condition 2.C.(2)(b) requiring the licensee to perform generator load reject and full MSIV closure transient tests at specified reactor thermal power levels. As discussed, the licensees February 27, 2004, application as supplemented, is seeking two amendments that would eliminate this license condition entirely with the first amendment eliminating only the full MSIV closure test. Although the NRC staff used SRP 14.2.1, the staff noted that SRP 14.2.1 covers the entire EPU test program and a review of the licensees overall EPU test program was performed in the SE for Amendment No. 243.

Therefore, the focus of this SE is on issues related to the elimination of the performance of the full MSIV closure transient test.

License condition 2.C.(2)(b) states, The licensee will perform the generator load reject and full main steam line isolation valve closure transients tests required by the General Electric Licensing Topical Report for Extended Power Uprate (NEDC-32424P-A)-ELTR-1, including the allowances described in Section L.2.4(2) of ELTR-1 regarding credit for unplanned plant transient events, using the thermal power level (1658 MWt) to establish ELTR-1 power level limits. The testing shall be performed at an initiating power level greater than the steady-state operation power level exceeding the respective ELTR-1 power level limit for each transient.

NEDC-32424P-A, Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate, is hereinafter referred to as ELTR-1. Following the issuance of DAEC Amendment No. 243, General Electric (GE) Company revised ELTR-1 to state that testing involving an automatic scram from a high power (which would include the DAEC generator load reject and MSIV closure tests) is not required. In a letter to GE dated March 31, 2003, the NRC took exception to GE's proposed elimination of large transient testing and stated that the NRC staff was preparing guidance to generically address the requirement for conducting large transient tests in conjunction with power uprates. The NRC subsequently provided this guidance in SRP 14.2.1. SRP 14.2.1 allows licensees to either perform the large transient tests (which would include the DAEC generator load reject and MSIV closure tests) or provide adequate technical justification for not performing the tests. To ensure consistency throughout this SE when power levels are discussed, the following table is included:

Power Level Date Related Information Original Rated Thermal Power 1593 MWt 1974 Initial plant licensed thermal power Current Rated Thermal Power (CRTP) 1658 MWt 1985 EPU Phase I 1790 MWt December 2001 EPU Phase II 1840 MWt Spring 2005 1840 MWt is planned. Final achievable power level to be determined.

EPU Phase III 1912 MWt Not yet scheduled Power Level in ELTR-1 for Main Steam Isolation Valve Closure Test 1823.8 MWt Power level in ELTR-1 for test (10% of 1658 MWt).

Power Level in ELTR-1 for Generator Load Reject Test 1906.7 MWt Power level in ELTR-1 for test (15% of 1658 MWt).

2.0 REGULATORY EVALUATION

The purpose of the EPU test program is to verify that SSCs will perform satisfactorily in service at the proposed EPU power level. The NRC staffs review covers (1) plans for the initial approach to the proposed maximum licensed thermal power level, including verification of adequate plant performance, (2) integrated plant systems testing, including transient testing, if necessary, to demonstrate that plant equipment will perform satisfactorily at the proposed increased maximum licensed thermal power level, and (3) the test programs conformance with applicable regulations. The NRC staffs acceptance criteria for the proposed EPU test program was based, in part, on (1) Appendix B to 10 CFR Part 50, Criterion XI, which requires establishment of a test program to demonstrate that SSCs will perform satisfactorily in service, (2) General Design Criterion 1, Quality Standards and Records, of Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, insofar as it requires that SSCs important to safety be tested to quality standards commensurate with the importance of the safety functions to be performed, (3) 10 CFR Part 50.34, Contents of Applications: Technical Information, which specifies requirements for the content of the original operating license application, including Final Safety Analysis Report (FSAR) plans for pre-operational testing and initial operations, and (4) Regulatory Guide (RG) 1.68, Appendix A, Section 5, Power Ascension Tests, which describes tests that demonstrate that the facility operates in accordance with design both during normal steady-state conditions, and, to the extent practical, during and following anticipated operational occurrences (AOOs). Specific review and acceptance criteria are contained in SRP 14.2.1.

3.0 TECHNICAL EVALUATION

3.1 SRP 14.2.1 Section III.A - Comparison of Proposed Test Program to the Initial Plant Test Program 3.1.1 Evaluation Criteria of SRP 14.2.1 Section III.A SRP 14.2.1 Section III.A, specifies the guidance and acceptance criteria that the licensee should use to compare the proposed EPU testing program to the original power ascension test program performed during initial plant licensing. The scope of this comparison should include (1) all initial power ascension tests performed at a power level of equal to or greater than 80 percent of the original licensed thermal power level, and (2) initial test program tests performed at lower power levels if the EPU would invalidate the test results. The licensee shall either repeat initial power ascension tests within the scope of this comparison or adequately justify proposed deviations from the initial power ascension test program. The following specific criteria should be identified in the EPU test program:

C all power ascension tests initially performed at a power level of equal to or greater than 80 percent of the original licensed thermal power level, C

all initial test program tests performed at power levels lower than 80 percent of the original licensed thermal power level that would be invalidated by the EPU, and C

differences between the proposed EPU power ascension test program and the portions of the initial test program identified by the previous criteria.

3.1.2 NRC Staff Evaluation Using SRP 14.2.1 Section III.A.

The NRC staff reviewed the licensees Plant Uprate Safety Analysis Report for testing recommended in ELTR-1. The licensee compared the initial startup test program, and consistent with the NRC-approved generic EPU guidelines in ELTR-1, the EPU was determined to require only a limited subset of the original startup test program. As applicable to this plants design, testing for the EPU is consistent with the description in ELTR-1. Specifically, the following testing was performed for Phase I and will be performed for Phases II and III during the power ascension steps of the EPU.

C Testing will be performed in accordance with the TS surveillance requirements on the instrumentation that requires re-calibration for the EPU conditions.

C Steady-state data will be taken at points from 90 percent up to the previous reactor thermal power so that system performance parameters can be projected for the EPU before the previous power rating is exceeded.

C Power increases beyond the previous reactor thermal power level will be made in increments of equal to or less than 5 percent power. Steady-state operating data, including fuel thermal margin, will be taken and evaluated at each step. Routine measurements of reactor and system pressures, flows, and vibration will be evaluated from each measurement point prior to the next power increment.

C Control system tests will be performed for the feedwater/reactor water level controls and pressure controls. These operational tests will be made at the appropriate plant conditions for each test and at each power increment above the previous rated power condition to show acceptable adjustments and operational capability. The same performance criteria will be used as in the original power ascension tests.

C A test specification will identify the EPU tests, the associated acceptance criteria, and the appropriate test conditions. All testing will be done in accordance with Appendix B to 10 CFR Part 50, Criterion XI.

The licensees test plan follows the guidance of ELTR-1 and satisfies the applicable requirements in Appendix B to 10 CFR Part 50; therefore, the NRC staff found the test plan acceptable.

The staff reviewed the power ascension testing performed as part of the original plan described in the DAEC Updated Final Safety Analysis Report (UFSAR) Table 14.2-3. The basis for testing was described in UFSAR Section 14.2.1.3. The startup testing requirements for the original DAEC test program were listed in Specification 22A2569, General Electric Startup Test Specification. By letter dated August 9, 2004, the licensee provided a comparison of the EPU test program with the original plant startup test program, as described in DAEC UFSAR Section 14.2. Additionally, the licensee provided a matrix of these tests versus the thermal power levels at which testing was performed for Phase I and future phases of the EPU program. The NRC staff found that essentially, the test plans were similar in scope. However, the EPU plans do not include a full MSIV closure test (or main generator load reject test).

The NRC staff reviewed the following EPU test plan information provided by the licensee in order to verify that the initial EPU license amendment submittal, supplemental information provided in response to NRC staff RAIs, and applicable sections of TSs and the UFSAR addressed the specific criteria for an adequate EPU test program as described in SRP 14.2.1.

Specifically, the following documents were reviewed during the NRC staffs evaluation:

C FSAR Section 14, Initial Test Program - Provided a detailed description of the licensees initial startup test programs (1) administrative controls (2) scope of testing (systems tested), and (3) the overall test objectives, methods, and acceptance criteria.

C DAEC letter NG-05-0010, Request for Segmented Review of License Amendment Request (TSCR-056), dated January 7, 2005 - Provided a description of the revised request of the proposed change to the operating license, which would eliminate the MSIV closure test as part of the EPU.

C DAEC letter NG-04-011, License Amendment Request (TSCR-056): Elimination of License Condition 2.C.(2)(b) for Performance of Large Transient Tests for Extended Power Uprate, dated February 27, 2004 - Provided a description of the proposed change, the supporting technical analysis, and evaluation of the No Significant Hazards Consideration for removing the license condition to perform large transient testing as part of the EPU.

C DAEC letter NG-04-0478, Response to Request for Additional Information Regarding License Amendment Request (TSCR-056): Elimination of License Condition 2.C.(2)(b) for Performance of Large Transient Tests for Extended Power Uprate, dated August 9, 2004 - Provided responses to NRC staff questions for (1) a comparison of the EPU test program to the initial plant test program, (2) modifications and the associated post-modification tests (PMTs) that were performed and are planned for the EPU, and (3) the licensees response on how SRP 14.2.1 was addressed.

C DAEC letter NG-01-764, Response to Request for Additional Information (RAI) to Technical Specification Change Request TSCR-042 - Extended Power Uprate, dated June 11, 2001 - Provided licensee responses to RAIs on (1) proposed implementation of the power uprate phases, (2) types of high power startup tests performed, (3) recent transient events that could be an indicator of plant response to the EPU, and (4) post-scram evaluation of applicable transient events.

C DAEC letter NG-01-1198, Final Typed Pages for Technical Specification Change Request TSCR-042 - Extended Power Uprate, dated October 17, 2001 - Provided inclusion of the commitment to perform certain transient testing during power ascension to the new licensed power level.

C DAEC letter NG-02-0187, Startup Test Report for Extended Power Uprate - Phase 1, dated March 4, 2002 - Provided a summary of the startup testing performed at DAEC following implementation of the first phase of the EPU, which increased thermal power 8 percent from 1658 MWt (CRTP) to 1790 MWt (Phase I).

C Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 243 to Facility Operating License No. DPR-49 Nuclear Management Company, LLC Duane Arnold Energy Center Docket No. 50-331, dated November 6, 2001 - Provided an NRC safety evaluation of the licensees proposed amendment request to allow an increase of the authorized operating power level from 1658 MWt (CRTP) to 1912 MWt (Phase III). The change represented an increase of 15.3 percent power above the current rated thermal power and therefore, was considered an EPU.

As part of this SE, the NRC staff reviewed the previous staff assessment of the EPU test program done for Amendment No. 243. Amendment No. 243 authorized operation up to 1912 MWt. Actual implementation of the EPU is being conducted in phases that support the licensees modification schedule. Refer to the table in Section 1 of this SE for the power levels associated with the EPU phases.

As part of the licensees review of the original test program, the following additional tests were evaluated for applicability to the EPU and added.

C Steady-State Data Collection: Key nuclear steam supply system and balance of plant parameters were recorded to ensure proper plant equipment performance.

C Power Conversion System Piping Vibration Monitoring: Main steam and feedwater (FW) piping was instrumented and monitored for unacceptable flow-induced vibrations.

C Turbine Combined Intermediate Valve (CIV) and Turbine Control Valve (TCV)

Surveillance Testing: Testing similar to original testing for the turbine stop valve was conducted on the CIVs and TCVs. The purpose of the testing was to establish the proper level for conducting on-line surveillance testing of the CIVs and TCVs.

C General Service Water (GSW) Heat Exchanger Performance Monitoring: GSW piping size was increased for the EPU to provide additional cooling to key components. This monitoring program will confirm adequate design cooling.

Phase I Test Program During performance of the Phase I test program, some acceptance criteria needed to be modified, as the original FSAR startup testing requirements were no longer applicable to the existing plant configuration. A problem in the FW level control system was discovered that required maintenance and re-performance of those tests at 1658 MWt. Also, based upon review of test data at lower power levels, the test matrix at high power was simplified and some tests were not performed, as they would not have provided useful data.

The completed testing at the Phase I target power level of 1790 MWt demonstrated stable plant operation. Changes in plant chemistry and radiological conditions were minor, vibration monitoring of main steam and FW piping was normal, and no plant equipment anomalies were noted.

The NRC staff found that all tests described in the initial startup test program were addressed in the description of the Phase I EPU test program. The NRC resident staff observed portions of the Phase I testing. No significant deficiencies were noted.

Phase II Test Program The NRC staff reviewed the proposed testing for Phase II, which will increase power to approximately 1840 MWt. Specifically, the NRC staff reviewed the changes to the test program for Phase II that differ from the NRC staff review performed for Amendment No. 243. The licensee is herein proposing to eliminate the following test discussed below:

C Test No. 25b, MSIVs - Full MSIV Closure Test: This test was not required as part of EPU Phase I testing, as the required power level per the license condition is 1823.8 MWt (ELTR-1 power level for the MSIV closure test), which was not reached in Phase I. This test is currently required to be performed as part of Phase II testing. However, the purpose of this license amendment request is to not perform this test as part of EPU testing.

3.

1.3 NRC Staff Conclusion

s Related to SRP 14.2.1 Section III.A.

The NRC staff concludes, through comparison of the documents referenced above, a review of test results from Phase I referenced in the FSAR, and a review of the test commitments proposed for Phase II, that the proposed EPU test program adequately identified (1) all initial power ascension tests performed at a power level of equal to or greater than 80 percent of the original licensed thermal power level, and (2) differences between the proposed EPU power ascension test program and the portions of the initial test program.

3.2 SRP 14.2.1 Section III.B.- Post Modification Testing Requirements for SSCs Important to Safety Impacted by EPU-Related Plant Modifications 3.2.1 Evaluation Criteria of SRP 14.2.1 Section III.B SRP 14.2.1 Section III.B., specifies the guidance and acceptance criteria which the licensee should use to assess the aggregate impact of the EPU plant modifications, setpoint adjustments, and parameter changes that could adversely impact the dynamic response of the plant to AOOs. AOOs include those conditions of normal operation that are expected to occur one or more times during the life of the plant and include events such as loss of all offsite power, tripping of the main turbine generator set, and loss of power to all reactor coolant pumps. The EPU test program should adequately demonstrate the performance of SSCs important to safety that meet all of the following criteria (1) the performance of the SSC is impacted by EPU-related modifications, (2) the SSC is used to mitigate an AOO described in the plant-specific design-basis, and (3) involves the integrated response of multiple SSCs. The following should be identified in the EPU test program as it pertains to the above paragraph:

plant modifications and setpoint adjustments necessary to support operation at power uprate conditions, and changes in plant operating parameters (such as reactor coolant temperature, pressure, reactor pressure, flow, etc.) resulting from operation at EPU conditions.

3.2.2 NRC Staff Evaluation Using SRP 14.2.1 Section III.B The NRC staff reviewed the planned EPU modifications and their potential effect on SSCs as documented in the DAEC letter NG-04-0478. The PMTs listed in the attachment to that letter were the acceptance tests to demonstrate design function performance and integration with the existing plant. The NRC staff also reviewed the basis for the licensees conclusions that the modifications did not change the design function of the SSCs or the methods of performing or controlling their functions. The following modifications and PMT descriptions were reviewed by the NRC staff.

The following modifications were completed in May 2001 for Phase I (operation to 1790 MWt):

C Changes to the main turbine included (1) the high pressure turbine was replaced, (2) turbine control valve operation was converted to partial arc admission, and adjustments made to the electro-hydraulic control (EHC) system.

C Changes to the main generator included (1) new hydrogen coolers with increased cooling capacity, and (2) new GSW piping of increased capacity to support the larger hydrogen coolers.

C Larger main transformer coolers were installed.

C New temperature sensors to monitor isophase buss temperature were installed.

C A capacitor bank was installed to increase plant volts-ampere reactive capability and enhance grid stability.

C Changes to the FW heaters included (1) adjustment to FW heater level control settings to new heat balance, (2) trim on FW heater level control valves to allow higher flow, and (3) installation of a bypass around FW heaters 5A/B to maintain extraction steam flow at pre-EPU values for heater tube vibration concerns.

C Tube stakes were installed on the high and low pressure condenser tubes for vibration dampening.

C Instrumentation upgrades included (1) re-calibration of the local power range monitors and average power range monitors to the new 100 percent power, (2) trip reference cards installed for the maximum extended load-line limit analysis (MELLLA) operating domain on the power-to-flow map, (3) new main steamline high flow trip instruments installed and re-calibrated to new setpoint, (4) turbine first stage pressure (reactor protection system and end-of-cycle recirculation pump trip bypass) were re-calibrated to new setpoints, based upon operating characteristics of the new high pressure turbine, (4) revised alarm setpoint for the standby liquid control system tank volume alarm, (5) control room indications respanned to new ranges, and (6) the process computer re-programmed to new instrument ranges.

C Sensors and a data collection system were installed for the main steam and FW piping vibration monitoring system.

C The main steam reheater cross-around relief valve capacity was increased (phased upgrade - one valve planned for each outage over four refueling outages).

All of the Phase I modifications have been installed, tested (performance monitoring, calibrations and startup testing) and are currently in operation. The NRC resident staff observed several of the PMTs performed for the above modifications. Also, portions of the Phase I power ascension were also observed. In addition, during the ensuing plant operation since EPU implementation, several plant events have occurred, including manual scrams from intermediate power levels, as well as a dual main recirculation pump runback event. In none of these actual events has the plants dynamic response been abnormal. The NRC staff found the PMTs and subsequent observed equipment performance acceptable for the modifications performed in Phase I.

The following modifications are scheduled to be completed in the spring of 2005 for Phase II (operation to approximately 1840 MWt):

C The condensate pumps and motors will be upgraded to allow higher flow rate and their electrical protective relay settings adjusted. The PMT will include (1) factory acceptance testing (full flow performance test with motor), (2) pump and motor vibration baseline measurements, and (3) performance monitoring.

C FW heater upgrades will continue with replacement of the 3A/B, 4A/B and 5A/B FW heaters. The PMT will include (1) factory acceptance testing (eddy-current testing and non-destructive examination of welds), (2) In-service leak testing, (3) thermal performance testing, and (4) FW heater level controller adjustments.

The Phase II modifications are primarily to address current FW and condensate system flow capacity limitations. The modifications will bring system capacity up to that needed to achieve a target power level of approximately 1840 MWt. Because modifications are focused on the FW and condensate system, testing will target this equipment, in addition to the general testing required during power ascension. These modifications will not significantly change the overall plant dynamic response to the anticipated initiating events described in the UFSAR. The NRC staff found the proposed PMTs acceptable for the modifications to be conducted in Phase II.

3.

2.3 NRC Staff Conclusion

s Related to SRP 14.2.1 Section III.B The NRC staff concludes, based on review of each planned modification, the associated PMT, and the basis for determining the appropriate test, that the EPU test program will adequately demonstrate the performance of SSCs important to safety; included in this analysis are those SSCs (1) impacted by EPU-related modifications, (2) used to mitigate an AOO described in the plant design basis, and (3) supported a function that relied on integrated operation of multiple systems and components.

The NRC staff concludes that the proposed test program adequately identified plant modifications and setpoint adjustments necessary to support operation at the uprated power level and changes in plant operating parameters (such as reactor coolant temperature, pressure, reactor pressure, flow, etc.) resulting from operation at EPU conditions. Additionally, the NRC staff determines there are no unacceptable system interactions because of modifications to the plant.

3.3 SRP 14.2.1 Section III.C - Justification for Elimination of EPU Power Ascension Tests 3.3.1 Evaluation Criteria Using SRP 14.2.1 Section III.C SRP 14.2.1 Section III.C., specifies the guidance and acceptance criteria the licensee should use to provide justification for a test program that does not include all of the power ascension testing that should be considered for inclusion in the EPU test program pursuant to the review criteria of Sections 1 and 2 above. The proposed EPU test program shall be sufficient to demonstrate that SSCs will perform satisfactorily in service. The following factors should be considered, as applicable, when justifying elimination of power ascension tests:

C previous operating experience, C

introduction of new thermal-hydraulic phenomena or identified system interactions, C

facility conformance to limitations associated with analytical analysis methods, C

plant staff familiarization with facility operation and trial use of operating and emergency operating procedures, C

margin reduction in safety analysis results for anticipated operational occurrences, and C

guidance contained in vendor topical reports C

risk implications.

3.3.2 NRC Staff Evaluation Using SRP 14.2.1 Section III.C The NRC staff focused the review on information regarding the following exception to original startup testing contained in the licensee RAI response letters NG-04-0478 and NG-01-0764.

C Test No. 25b, MSIVs - Full MSIV Closure Test: This test was not required as part of EPU Phase I testing, as the required power level per the license condition is 1823.8 MWt (ELTR-1 power level for MSIV closure test), which was not reached in Phase I. As part of the license condition, this test is currently required to be performed as part of Phase II testing. However, the purpose of this license amendment request is to not perform this test as part of EPU testing.

The NRC staff reviewed the licensees response in NG-01-0764 regarding previous operating experience. The DAEC experienced unplanned events at approximately 1658 MWt (CRTP),

which provided data for the MSIV closure test. In the first event, when the reactor was operating at approximately 1658 MWt, one MSIV unexpectedly closed due to a failed solenoid.

Reactor pressure and reactor power increased and steam flow through the remaining three steamlines increased, until a full isolation of the main steamlines was initiated on high steam flow. No significant anomalies in the plant response were observed. In the second event, with the same reactor power, the main generator backup lockout differential current trip resulted in a turbine control valve fast closure event. The primary source signal for the reactor scram was the pressure switches on the EHC system that signal the fast closure of the turbine control valve. Again, no significant anomalies in the plant response were observed, with one exception.

The FW controls allowed reactor level to increase to greater than the FW pump trip setpoint.

While the Level 2 criterion (licensee established criterion for FW level control) was not met, the Level 1 criterion that the steamlines not flood was met. There is no safety consequence to the level 2 criterion not being met. Normal reactor water level control was subsequently established. The NRC resident staff observed the FW control troubleshooting. The licensee adequately resolved the FW control setpoint issue.

The licensee also cited Hatch Nuclear Plant, Unit 2, as an example of a similar plant which had an event subsequent to their EPU. Plant Hatch, Unit 2, is a boiling-water reactor (BWR) 4 with a Mark I containment of essentially the same design as the DAEC, including the key balance of plant area of turbine generator control logic. Hatch Nuclear Plant, Unit 2, had an unplanned event which resulted in a generator load reject from their full uprated power level. No anomalies were seen in the plants response to this event. In addition, Plant Hatch, Unit 1, has experienced one turbine trip and one generator load reject event subsequent to its uprate.

Again, the primary safety systems performed as expected. No new plant behaviors have been observed that would indicate that the analytical models being used are not capable of modeling plant behavior at the EPU conditions. A turbine trip and generator load reject event result in a pressurization transient similar to an MSIV closure event.

In response to the possible introduction of new thermal-hydraulic phenomena or identified system interactions, the licensee responded that none of the modifications implemented should have an impact in this area. The major EPU modification to the DAEC was to modify the main steam flow path from the reactor to the turbine generator to accommodate the higher steam flow due to the EPU. A new, more efficient high pressure turbine was installed and the TCVs were converted to partial arc mode. However, neither of these modifications introduced new thermal-hydraulic phenomena in the plant, nor do they introduce new or different system interactions that would warrant performing a pressurization transient test. The conversion to partial arc admission lessens the severity of a pressurization transient from operation in full arc admission.

In addition, no instrument setpoints were modified that initiate equipment relied upon to mitigate this event.

Specifically, MSIV stroke times were not changed, nor were the opening settings of the safety/relief valves (S/RVs). No instrument setpoints were modified that initiate equipment relied upon to mitigate this event, such as the MSIV closure signal that initiates a reactor scram.

The MSIV closure is a pressurization transient caused by a fast shutoff of steam flow from the reactor vessel, from closure of the MSIVs. The transient severity is primarily determined by the initial operating pressure and rate of pressure increase (i.e., valve closure time). Rated reactor power (i.e., rated steam flow), has a noticeable, but secondary effect on the rate of pressure increase. NMC has implemented the DAEC EPU without a reactor pressure increase (commonly referred to as a constant pressure power uprate), or change in the shutoff valve stroke times. In addition, no modifications to the major SSCs used to mitigate this transient, such as the S/RVs or turbine bypass valves, have been made. Only rated steam flow has been affected by the EPU.

The NRC staff reviewed the licensees response in NG-04-0111 to the introduction of new thermal-hydraulic phenomena or identified system interactions. The major EPU modification to the plant was to modify the main steam flow path from the reactor to the turbine generator to accommodate the higher steam flow due to the EPU. A new, more efficient high pressure turbine was installed and the turbine control valves were converted to partial arc mode.

However, neither of these modifications introduced new thermal-hydraulic phenomena in the plant, nor do they introduce new or different system interactions that would warrant performing the MSIV closure test. As noted above, the conversion to partial arc admission lessens the severity of a pressurization transient from operation in full arc admission.

The NRC staff reviewed Section 3.7 of the Nuclear Reactor Regulation (NRR) SE for the DAEC EPU. Section 3.7 discussed the assessment of the effects of the EPU on the MSIV closure times. The original SE indicated that the NRC staff accepted the generic assessment on the MSIVs, which was documented in Section 4.7 of Supplement 1 to ELTR-2. The generic evaluation covered the effects of the power uprate changes on (1) the capability of the MSIVs to meet pressure boundary structural requirements, and (2) the safety function of the MSIVs.

The NRC staff accepted the generic assessment that the MSIV closure time can be maintained as analyzed and specified in the TSs. In addition, various surveillances require routine monitoring of MSIV closure time and leakage to ensure that the licensing basis for the MSIVs is preserved.

Based on the review of the evaluation and rationale, the NRC staff agreed with the conclusion that EPU operation would remain bounded by the generic evaluation in Section 4.7 of ELTR-2 and that the plant operation at the EPU level will not affect the ability of the MSIVs to perform their safety function.

The NRC staff reviewed the licensees response in NG-04-0111 to facility conformance to limitations associated with analytical analysis methods. The licensee used General Electrics analytical model for analyzing transients (ODYN) and associated methods (GEMINI), which have been proven to acceptably predict plant behavior during a pressurization transient, including the DAEC, even at EPU conditions (e.g., Hatch). These methods are routinely used in the analysis of core reloads that form the basis for the core operating limit requirements. No new limitations on these methods have been imposed as a result of EPU implementation.

The NRC staff reviewed plant staff familiarization with facility operation and trial use of operating and emergency operating procedures. The NRC staff has previously reviewed and approved NMCs process for updating the plant operating procedures (normal and off-normal), training (including plant simulator), and human factors aspects of the DAECs EPU implementation.

The NRC staff also noted that in describing and justifying test exceptions or deviations, the licensee adequately considered previous operating experience, the possible introduction of new thermal-hydraulic phenomena or system interactions, and margin reduction in safety analysis results for AOOs. Other factors used to determine the EPU test elimination included use of baseline operational data, updated computer modeling analyses, and industry experience.

Risk informed justifications for not performing a transient test was considered, as described in Section 10.4 of the SE for Amendment No. 243, but was not the sole factor in determining elimination of those tests. Previous operating experience, the initial startup test program report, computer model analyses and surveillance requirements were the major factors on those decisions.

3.

3.3 NRC Staff Conclusion

s Related to SRP 14.2.1 Section III.C The NRC staff concludes that, in justifying test eliminations or deviations, the licensee adequately addressed factors that included (1) previous operating experience, (2) introduction of new thermal-hydraulic phenomena or system interactions, and (3) staff familiarization with facility operation and use of operating and emergency operating procedures. The NRC staff determined that the licensee did not rely on analytical analysis as the sole basis for elimination of a power ascension test from the proposed EPU test program. Construction, installation and/or pre-operational testing for each modification will be performed in accordance with the plant design process procedures. The final acceptance tests will demonstrate that the modifications will perform their design function and integrate appropriately with the existing plant.

3.4 SRP 14.2.1 Section III.D - Adequacy of Proposed Testing Plans 3.4.1 Evaluation Criteria of SRP 14.2.1 Section III.D SRP 14.2.1 Section III.D, specifies the guidance and acceptance criteria the licensee should use to include plans for the initial approach to the increased EPU power level and testing that should be used to verify that the reactor plant operates within the values of EPU design parameters. The test plan should assure that the test objectives, test methods, and the acceptance criteria are acceptable and consistent with the design basis for the facility. The predicted testing responses and acceptance criteria should not be developed from values or plant conditions used for conservative evaluations of postulated accidents. During testing, safety-related SSCs relied upon during operation shall be verified to be operable in accordance with existing and Quality Assurance Program requirements. The following should be identified in the EPU test program:

the method in which initial approach to the uprated EPU power level is performed in an incremental manner including steady-state power hold points to evaluate plant performance above the original full-power level, appropriate testing and acceptance criteria to ensure that the plant responds within design predictions including development of predicted responses using real or expected values of items such as beginning-of-life core reactivity coefficients, flow rates, pressures, temperatures, response times of equipment, and the actual status of the

plant, contingency plans if the predicted plant response is not obtained, and a test schedule and sequence to minimize the time untested SSCs important to safety are relied upon during operation above the original licensed full-power level.

3.4.2 NRC Staff Evaluation Using SRP 14.2.1 Section III.D The NRC staff reviewed Attachment 6 of NG-00-1900, which outlined the licensees proposed EPU test plan. The NRC staff also reviewed the original NRR SEs conclusions on the adequacy of the startup test program. The NRC staff had concluded that the licensees test plan followed the guidelines of ELTR-1 and satisfied the applicable requirements in Appendix B to 10 CFR Part 50.

The licensee will conduct limited startup testing at the time of implementation of the proposed EPU. The tests will be conducted in accordance with the guidelines of ELTR-1 to demonstrate the capability of plant systems to perform their design functions under uprated conditions.

The tests will be similar to some of the original startup tests described in Table 14.2-3 and Section 14.2.1.3 of the DAEC UFSAR. Testing will be conducted with established controls and procedures which have been revised to reflect the uprated conditions.

The tests will consist essentially of steady-state, baseline tests between 90 and 100 percent of the currently licensed power level. Several sets of data will be obtained between 100 and 115.3 percent current power with no greater than 5 percent power increments between data sets. A final set of data at the proposed EPU power level will also be obtained. The tests will be conducted in accordance with a site-specific test procedure, currently being developed by the licensee. The test procedure will be developed in accordance with written procedures as required by 10 CFR Part 50, Appendix B, Criterion XI, Test Control.

The licensee indicated that the power increase test plan will have features as described in the Power Uprate Safety Analysis Report, Section 10.4, Required Testing. Initial power ascension testing is outlined in Section 2.B.1 of this SE.

The guidelines in ELTR-1, Section 5.11.9, specify that pre-operational tests will be performed for systems or components which have revised performance requirements. These tests will occur during the ascension to EPU conditions. The performance tests and associated acceptance criteria are based on DAECs original startup test specifications and previous General Electric BWR EPU test programs. The licensees performance tests are discussed in Section 2.B.2 of this SE.

The NRC staff noted that the results from the uprate test program will be used to revise the operator training program to more accurately reflect the effects of the proposed EPU.

In addition, the plant staff, through classroom and/or simulator training, will be familiarized with the operation of the plant under EPU conditions. The training will include (1) plant modification and parameter value changes, (2) implementation/execution of normal, abnormal, and emergency operating procedures, and (3) accident mitigation strategies.

3.

4.3 NRC Staff Conclusion

s Related to SRP 14.2.1 Section III.D The NRC staff concludes that the proposed test plan will adequately assure that the test objectives, test methods, and test acceptance criteria are consistent with the design-basis for the facility. Additionally, the NRC staff concludes that the test schedule would be performed in an incremental manner, with appropriate hold points for evaluation, and contingency plans exist if predicted plant response is not obtained.

3.5 Technical Evaluation Summary The NRC staff has reviewed the EPU test program in accordance with SRP Section 14.2.1.

This review included an evaluation of: (1) plans for the initial approach to the proposed Phase II thermal power level, including verification of adequate plant performance, (2) transient testing necessary to demonstrate that plant equipment will perform satisfactorily at the proposed Phase II thermal power level, and (3) the test programs conformance with applicable regulations. For the reasons set forth above, the NRC staff concludes that the proposed EPU test program provides reasonable assurance that the plant will operate in accordance with design criteria and that SSCs affected by the EPU or modified to support the proposed power uprate will perform satisfactorily while in service. On this basis, the NRC staff finds that the EPU testing program satisfies the requirements of 10 CFR Part 50, Appendix B, Criterion XI, Test Control.

Therefore, the NRC staff finds the licensees proposed license amendment request to modify license condition 2.C.(2)(b) to eliminate the requirement to perform the full MSIV closure test from the EPU test program acceptable.

4.0 STATE CONSULTATION

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

5.0 ENVIRONMENTAL CONSIDERATION

S 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 published April 13, 2004, (69 FR 19572). 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.

Principal Contributor: P. Prescott Date: March 17, 2005