Source: https://www.federalregister.gov/documents/2001/04/24/01-10095/nuclear-management-company-llc-duane-arnold-energy-center-environmental-assessment-and-finding-of-no
Timestamp: 2018-02-26 02:27:09
Document Index: 493208133

Matched Legal Cases: ['art 50', 'art 50', 'art 50', 'art 50', 'art 50', 'art 50']

Federal Register :: Nuclear Management Company, LLC; Duane Arnold Energy Center; Environmental Assessment and Finding of No Significant Impact
Nuclear Management Company, LLC; Duane Arnold Energy Center; Environmental Assessment and Finding of No Significant Impact
A Notice by the Nuclear Regulatory Commission on 04/24/2001
66 FR 20692
01-10095
https://www.federalregister.gov/d/01-10095 https://www.federalregister.gov/d/01-10095
The U.S. Nuclear Regulatory Commission (NRC) is considering issuance of an exemption from certain requirements of 10 CFR part 50, appendix G, for Facility Operating License No. DPR-49, issued to Nuclear Management Company, LLC (NMC, or the licensee) for operation of the Duane Arnold Energy Center (DAEC), located in Linn County, Iowa.
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, “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 requirements for these limits are the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), Section XI, Appendix G Limits.
To address provisions of amendments to the technical specifications (TS) P-T limits, the licensee requested in its submittal dated October 16, 2000, that the staff exempt NMC from application of specific requirements of 10 CFR part 50, appendix G, and substitute use of ASME Code Case N-640. The license amendment request is being addressed as a separate action. Code Case N-640 permits the use of an alternate reference fracture toughness (KIc fracture toughness curve instead of KIa fracture toughness curve) for reactor vessel materials in determining the P-T limits. Since the KIc fracture toughness curve shown in ASME Section XI, Appendix A, Figure A-2200-1 (the KIc fracture toughness curve) provides greater allowable fracture toughness than the corresponding KIa fracture toughness curve of ASME Section XI, Appendix G, Figure G-2210-1 (the KIa fracture toughness curve), using Code Case N-640 for establishing the P-T limits would be less conservative than the methodology currently endorsed by 10 CFR part 50, appendix G and, therefore, Start Printed Page 20693an exemption to apply the Code Case would be required by 10 CFR 50.60(b).
The proposed exemption is needed to allow the licensee to implement ASME Code Case N-640 in order to revise the method used to determine the reactor coolant system (RCS) P-T limits, because continued use of the present curves unnecessarily restricts the P-T operating window. 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 would unnecessarily require the RPV to maintain a temperature exceeding 212 degrees Fahrenheit in a limited operating window during the pressure test. Consequently, steam vapor hazards would continue to be one of the safety concerns for personnel conducting inspections in primary containment. Implementation of the proposed P-T curves, as allowed by ASME Code Case N-640, does not significantly reduce the margin of safety and would eliminate steam vapor hazards by allowing inspections in primary containment to be conducted at a lower coolant temperature.
In the associated exemption, the staff has determined that, pursuant to 10 CFR 50.12(a)(2)(ii), the underlying purpose of the regulation will continue to be served by the implementation of this Code Case.
With regard to potential nonradiological environmental impacts, the proposed action does not involve any historic sites. It does not affect nonradiological plant effluents and has no other environmental impact. Therefore, there are no significant nonradiological impacts associated with the proposed action.
This action does not involve the use of any resources not previously considered in the “Final Environmental Statement Relating to the Operation of the Duane Arnold Energy Center,” dated March 1973.
In accordance with its stated policy, on March 26, 2001, the staff consulted with the Iowa State official, Mr. D. McGhee of the Department of Public Health, regarding the environmental impact of the proposed action. The State official had no comments.
For further details with respect to the proposed action, see the licensee's letter dated October 16, 2000. Documents may be examined, and/or copied for a fee, at the NRC's Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible electronically from the ADAMS Public Library component on the NRC Web site, http://www.nrc.gov (the Electronic Reading Room).
Dated at Rockville, Maryland, this 17th day of April 2001.
[FR Doc. 01-10095 Filed 4-23-01; 8:45 am]