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Commonwealth Edison Company (Dresden Nuclear Power Station, Units 2 and 3); Exemption

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I

The Commonwealth Edison Company (ComEd, the licensee) is the holder of Facility Operating Licenses Nos. DPR-19 and DPR-25 which authorize operation of the Dresden Nuclear Power Station, Units 2 and 3 (Dresden). The licenses provide, 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 boiling water reactors (Units 2 and 3) located on the licensee's Dresden site in Grundy County, Illinois. This exemption refers to both units.

II

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 PT 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, in its submittal of February 23, 2000, the licensee requested that the staff exempt Dresden from application of specific requirements of 10 CFR Part 50, Section 50.60(a) and Appendix G, and substitute use of ASME Code Cases N-588 and N-640. Code Case N-588 permits the postulation of a circumferentially-oriented flaw (in lieu of an axially-oriented flaw) for the evaluation of the circumferential welds in RPV P-T limit curves. Since the pressure stresses on a circumferentially-oriented flaw are lower than the pressure stresses on an axially-oriented flaw by a factor of two, using Code Case N-588 for establishing the P-T limits would be less conservative than the methodology currently endorsed by 10 CFR Part 50, Appendix G and, therefore, an exemption to apply the Code Case would be required by 10 CFR 50.60(a). Code Case N-640 permits the use of an alternate reference fracture toughness (K1c fracture toughness curve instead of K1a fracture toughness curve) for reactor vessel materials in determining the P-T limits. Since the K1c fracture toughness curve shown in ASME Code, Section XI, Appendix A, Figure A-2200-1 provides greater allowable fracture toughness than the corresponding K1a fracture toughness curve of ASME Code, Section XI, Appendix G, Figure G-2210-1 (the K1a 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, an exemption to apply the Code Case would also be required by 10 CFR 50.60(a).

Code Case N-588

The licensee has proposed an exemption to allow the use of ASME Code Case N-588 in conjunction with ASME Code, Section XI, 10 CFR 50.60(a) and 10 CFR Part 50, Appendix G, to determine the P-T limits.

The proposed amendments to revise the P-T limits for Dresden rely, in part, on the requested exemption. These proposed P-T limits have been developed using the postulation of a circumferentially-oriented reference flaw as the limiting flaw in a RPV circumferential weld in lieu of an axially-oriented flaw required by the 1989 Edition of ASME Code, Section XI, Appendix G.

Postulating the Appendix G (axially-oriented flaw) reference flaw in a circumferential weld is physically unrealistic and overly conservative because the length of the flaw would extent well beyond the girth of the circumferential weld and into the adjoining base metal material. Industry experience with the repair of weld indications found during preservice inspection, and data taken from destructive examination of actual vessel welds, confirms that any remaining flaws are small, laminar in nature, and do not transverse the weld bead orientation. Therefore, any potential defects introduced during the fabrication process, and not detected during subsequent nondestructive examinations, would only be expected to be oriented in the direction of weld fabrication. A defect with a circumferential orientation is, therefore, postulated for circumferential welds.

An analysis provided to the ASME Code's Working Group on Operating Plant Criteria (WGOPC) (in which Code Case N-588 was developed) indicated Start Printed Page 53230that if an axial flaw is postulated on a circumferential weld, then based on the correction factors for membrane stress (Mm) given in the Code Case for the inside diameter circumferential (0.443) and axial (0.926) flaw orientations, it is equivalent to applying a safety factor of 4.18 on the pressure loading under normal operating conditions. Appendix G requires a safety factor of two on the contribution of the pressure load in the case of an axially-oriented flaw in an axial weld, shell plate, or forging. By postulating a circumferentially-oriented flaw on a circumferential weld and using the appropriate stress magnification factor, the margin of two (1.5 for pressure testing condition) is maintained for the contribution of the pressure load to the integrity calculation of the circumferential weld. Consequently, the staff determined that the posulation of an axially-oriented flaw on a circumferential RPV weld is a level of conservatism that is not required to establish P-T limits to protect the reactor coolant system (RCS) pressure boundary from failure during pressure testing and normal operations, including heatup, cooldown, and anticipated operational transients.

In summary, the ASME Code, Section XI, Appendix G, pocedure was developed for axially-oriented flaws, which is physically unrealistic and overly conservative for postulation flaws of this orientation to exist in circumferential welds. Hence, the NRC staff concurs that relaxation of the ASME Code, Section XI, Appendix G, requirements by application of ASME Code Case N-588 is acceptable and would maintain, 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.

Code Case N-640 (formerly Code Case N-626)

The licensee has proposed an exemption to allow the use of ASME Code Case N-640 in conjunction with ASME Code, Section XI; 10 CFR 50.60(a); and 10 CFR Part 50, Appendix G, to determine P-T limits.

The proposed amendment to revise the P-T limits for Dresden rely in part on the requested exemption. These revised P-T limits have been developed using the K1c fracture toughness curve, in lieu of the K1a fracture toughness curve, as the lower bound for fracture toughness.

Use of the K1c curve in determining the lower bound fracture toughness in the development of P-T operating limits curve is more technically correct that use of the K1a curve since the rate of loading during a heatup or cooldown is slow and is more representative of a static condition than a dynamic condition. The K1c curve appropriately implements the use of static initiation fracture toughness behavior to evaluate the controlled heatup and cooldown process of a reactor vessel. The staff has required use of the initial conservatism or the K1a curve since 1974 when the curve was codified. 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 K1a 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 K1c curve would enhance overall plant safety by opening the P-T operating window with the greatest safety benefit in the region of low temperature opertions.

Since the RCS P-T operating window is defined by the P-T operating and test limit curves developed in accordance with the ASME Code, Section XI, Appendix G, continued operation of Dresden with these P-T curves without the relief provided by ASME Code Case N-640 would unnecessarily require that the RPV maintain a temperature exceeding 212 degrees Fahrenheit in a limited operating window during pressure tests. 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 lower coolant temperature. 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 Code, Section XI, Appendix G, procedure was conservatively developed based on the level of knowledge existing tin 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 Code, 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.

III

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 the exemption would be required to approve the use of Code Cases N-588 and N-640. The staff examined the licensee's rationale to support the exemption requests and concurred that the use of the code cases would 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 1.99, Revision 2, the staff concludes that application of the code cases as described would provide an adequate margin of safety against brittle failure of the RPV and that application of the specific requirements of 10 CFR 50.60(a) and Appendix G in these circumstances is not necessary to achieve the underlying purpose of the rules. This is also consistent with the determination that the staff has reached for other licensees under similar conditions based on the same considerations (Quad Cities Nuclear Power Station, Units 1 and 2, exemption dated February 4, 2000). Therefore, the staff concludes that requesting an exemption under the special circumstances of 10 CFR 50.12(a)(2)(ii) is appropriate and that the methodology of Code Cases N-588 and N-640 may be used to revise the P-T limits for Dresden Nuclear Power Station, Units 2 and 3.

IV

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, and that special circumstances are present. Therefore, the Commission hereby grants Commonwealth Edison Company an exemption from the requirements of 10 CFR Part 50, Section 50.60(a) and 10 CFR Part 50, Appendix G, for Dresden Nuclear Power Station, Units 2 and 3.

Pursuant to 10 CFR 51.32, an environmental assessment and finding of no significant impact has been prepared and published in the Federal Start Printed Page 53231Register (65 FR 51344). 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 25th day of August 2000.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

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

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[FR Doc. 00-22498 Filed 8-31-00; 8:45 am]

BILLING CODE 7590-01-U