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Nuclear Management Company, LLC Duane Arnold Energy Center; Exemption

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1.0 Background

Nuclear Management Company, LLC (NMC, the licensee) is the holder of Facility Operating License No. DPR-49 which authorizes operation of the Duane Arnold Energy Center (DAEC). The license provides, among other things, that the facility is subject to all rules, regulations, and orders of the U.S. Nuclear Regulatory Commission (the Commission) now or hereafter in effect.

The facility consists of a boiling water reactor located on NMC's DAEC site, which is located in Linn County, Iowa.

2.0 Purpose

Title 10 of the Code of Federal Regulations (10 CFR) part 50, Appendix G requires that pressure-temperature (P-T) limits be established for reactor pressure vessels (RPVs) during normal operating and hydrostatic or leak rate testing conditions. Specifically, 10 CFR part 50, appendix G states that, “The appropriate requirements on both the pressure-temperature limits and the minimum permissible temperature must be met for all conditions.” Appendix G of 10 CFR part 50 specifies that the P-T limits must meet the safety margin requirements specified in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), Section XI, Appendix G.

To address provisions of the proposed amendments to the technical specification (TS) P-T limits, the licensee requested in its submittal dated October 16, 2000, that the staff exempt DAEC from application of specific requirements of 10 CFR part 50, § 50.60(a) and 10 CFR part 50, Appendix G, and substitute use of ASME Code Case N-640. Code Case N-640 permits the use of an alternate reference fracture toughness (Klc fracture toughness curve instead of Kla fracture toughness curve) for reactor vessel materials in determining the P-T limits. The proposed action is in accordance with the licensee's application for exemption contained in the October 16, 2000, submittal, and is needed to support the TS amendment request that is contained in the same submittal. The proposed amendment will revise the P-T limits for heatup, cooldown, and inservice test limitations for the reactor coolant system (RCS) to 25 and 32 effective full power years (EFPYs).

Code Case N-640

The licensee has proposed an exemption to allow use of ASME Code Case N-640 in conjunction with ASME Section XI, 10 CFR 50.60(a) and 10 CFR part 50, Appendix G, to determine that the P-T limits meet the underlying intent of the Nuclear Regulatory Commission (NRC) regulations.

The proposed amendment to revise the P-T limits for DAEC relies in part on the requested exemption. These revised P-T limits have been developed using the Klc fracture toughness curve shown in ASME Section XI, Appendix A, Figure A-2200-1, in lieu of the Kla fracture toughness curve of ASME Section XI, Appendix G, Figure G-2210-1, as the lower bound for fracture toughness. The other margins involved with the ASME Section XI, Appendix G process of determining P-T limit curves remain unchanged.

Use of the Klc curve in determining the lower bound fracture toughness in the development of P-T operating limits curve is more technically correct than the Kla curve. The Klc curve appropriately implements the use of static initiation fracture toughness behavior to evaluate the controlled heatup and cooldown process of a reactor vessel. The licensee has determined that the use of the initial conservatism of the Kla curve when the curve was codified in 1974 was justified. This initial conservatism was necessary due to the limited knowledge of RPV materials. Since 1974, additional knowledge has been gained about RPV materials, which demonstrates that the lower bound on fracture toughness provided by the Kla curve is well beyond the margin of safety required to protect the public health and safety from potential RPV failure. In addition, P-T curves based on the Klc curve will enhance overall plant safety by opening the P-T operating window with the greatest safety benefit in the region of low temperature operations. The operating window through which the operator heats up and cools down the RCS is determined by the difference between the maximum allowable pressure determined by Appendix G of ASME Section XI, and the minimum required pressure for the reactor coolant pump seals adjusted for instrument uncertainties.

Since the RCS P-T operating window is defined by the P-T operating and test limit curves developed in accordance with the ASME Section XI, Appendix G procedure, continued operation of DAEC with these P-T curves without the relief provided by ASME Code Case N-640 may unnecessarily restrict the P-Start Printed Page 22620T operating window, especially at low temperature conditions. The operating window becomes more restrictive with continued reactor vessel service. Implementation of the proposed P-T curves, as allowed by ASME Code Case N-640, does not significantly reduce the margin of safety. Thus, pursuant to 10 CFR 50.12(a)(2)(ii), the underlying purpose of the regulation will continue to be served.

In summary, the ASME Section XI, Appendix G procedure was conservatively developed based on the level of knowledge existing in 1974 concerning RPV materials and the estimated effects of operation. Since 1974, the level of knowledge about these topics has been greatly expanded. The NRC staff concurs that this increased knowledge permits relaxation of the ASME Section XI, Appendix G requirements by application of ASME Code Case N-640, while maintaining, pursuant to 10 CFR 50.12(a)(2)(ii), the underlying purpose of the ASME Code and the NRC regulations to ensure an acceptable margin of safety.

3.0 Discussion

Pursuant to 10 CFR 50.12, the Commission may, upon application by any interested person or upon its own initiative, grant exemptions from the requirements of 10 CFR part 50, when (1) the exemptions are authorized by law, will not present an undue risk to public health or safety, and are consistent with the common defense and security; and (2) when special circumstances are present. The staff accepts the licensee's determination that an exemption would be required to approve the use of Code Case N-640. The staff examined the licensee's rationale to support the exemption request and concurred that the use of the code case would also meet the underlying intent of these regulations. Based upon a consideration of the conservatism that is explicitly incorporated into the methodologies of 10 CFR part 50, Appendix G; Appendix G of the ASME Code; and regulatory guide (RG) 1.99, Revision 2, the staff concluded that application of the code case as described would provide an adequate margin of safety against brittle failure of the RPV. This is also consistent with the determination that the staff has reached for other licensees under similar conditions based on the same considerations. Therefore, the staff concludes that requesting the exemption under the special circumstances of 10 CFR 50.12(a)(2)(ii) is appropriate and that the methodology of Code Case N-640 may be used to revise the P-T limits for the DAEC RCS.

4.0 Conclusion

Accordingly, the Commission has determined that, pursuant to 10 CFR 50.12(a), the exemption is authorized by law, will not endanger life or property or common defense and security, and is, otherwise, in the public interest. Therefore, the Commission hereby grants NMC an exemption from the requirements of 10 CFR part 50, § 50.60(a) and 10 CFR part 50, Appendix G, for the DAEC.

Pursuant to 10 CFR 51.32, an environmental assessment and finding of no significant impact has been prepared and published in the Federal Register (66 FR 20692). Accordingly, based upon the environmental assessment, the Commission has determined that the granting of this exemption will not result in any significant effect on the quality of the human environment.

This exemption is effective upon issuance.

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Dated at Rockville, Maryland, this 27th day of April, 2001.

For the Nuclear Regulatory Commission.

Cynthia A. Carpenter,

Acting Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.

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[FR Doc. 01-11277 Filed 5-3-01; 8:45 am]