The FirstEnergy Nuclear Operating Company (FENOC/the licensee) is the holder of Facility Operating License No. NPF-73 that authorizes operation of the Beaver Valley Power Station, Unit 2. The license provides, among other things, that the licensee is subject to all rules, regulations, and orders of the U.S. Nuclear Regulatory Commission (NRC, the Commission) now or hereafter in effect.
The facility consists of a pressurized water reactor located in Shippingport, Beaver County, Pennsylvania.
Title 10 of the Code of Federal Regulations (10 CFR) Part 50, § 50.60(a), requires that “all light-water nuclear power reactors * * * must meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary set forth in appendices G and H to this part.” Appendix G to 10 CFR Part 50, 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, this regulation states that “[t]he appropriate requirements on * * * the pressure-temperature limits and the minimum permissible temperature must be met for all conditions.” Additionally, it specifies that the requirements for these limits are the American Society of Mechanical Engineers (ASME) Code, Section XI, Appendix G, Limits. This section of the ASME Code in turn specifies that RPV P/T limits be developed using the KIa fracture toughness curve of ASME Section XI, Appendix G, Figure G-2210-1, as the lower bound for fracture toughness.
Pressurized water reactor licensees have installed low temperature overpressure protection (LTOP) systems in order to protect the reactor coolant pressure boundary (RCPB) from being operated outside of the boundaries established by the P/T limit curves and to provide pressure relief of the RCPB during low temperature overpressurization events. The licensee is required by the Beaver Valley Unit 2 Technical Specifications (TSs) to update and submit the changes to its LTOP setpoints whenever the licensee is requesting approval for amendments to the P/T limit curves in the Beaver Valley Unit 2 TSs.
In order to address provisions of amendments to the TS P/T limits and LTOP curves, the licensee requested in its submittal dated June 17, 1999, that the staff exempt Beaver Valley Unit 2 from application of specific requirements of 10 CFR Part 50, § 50.60(a), and 10 CFR Part 50, appendix G, and substitute the use of ASME Code Case N-640. It should be noted that, as a result of ASME Code committee action, the original designation for this Code Case (N-626) was changed to N-640. Therefore, Code Case N-640 will be discussed below rather than Code Case N-626, which is the designation referenced in Attachments C and D of the submittal. Code Case N-640 is an alternate reference for fracture toughness for reactor vessel materials for use in determining the P/T limits.
The proposed action is in accordance with the licensee's application for exemption contained in a submittal dated June 17, 1999, and is needed to support the TS amendment that is contained in the same submittal. The proposed amendment will revise the P/T limits of TS 3/4.4.9 for Beaver Valley Unit 2 related to the heatup, cooldown, and inservice test limitations for the reactor coolant system (RCS) to 15 Effective Full Power Years (EFPYs). It will also revise the section of the TSs that relates to the overpressure protection system (OPPS) to reflect the revised P/T limits of the reactor vessels.
Code Case N-640 (formerly Code Case N-626)
The proposed amendment to revise the P/T limits for Beaver Valley Unit 2, relies, in part, on the requested exemption. In accordance with Code Case N-640, these revised P/T limits have been developed using the KKIc fracture toughness curve shown in ASME Section XI, Appendix A, Figure A-2200-1, in lieu of the KKIa 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 KIC curve in determining the lower bound fracture toughness in the development of the P/T operating limits curve is more technically correct than the KIa curve. The KIC curve appropriately implements the use of static initiation fracture toughness behavior to evaluate the controlled heatup and cooldown process of a reactor vessel. The use of the initial conservatism of the KIa 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, however, additional knowledge has been gained about RPV materials, which demonstrates that the lower bound on fracture toughness provided by the KIa 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 KIC 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. Current OPPS setpoints produce operational constraints by limiting the P/T range available to the operator for heatup or cooldown of the plant. The operating window through which the operator heats up and cools down the RCS is established 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 (RCP) seals adjusted for OPPS overshoot and instrument uncertainties. The operating window becomes more restrictive with continued reactor vessel service.
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 Beaver Valley Unit 2 with these P/T curves without the relief provided by ASME Code Case N-640 would unnecessarily restrict the P/T operating window, especially at low temperature conditions. Reducing this operating window could potentially have an adverse safety impact by increasing the possibility of inadvertent OPPS actuation due to pressure surges Start Printed Page 55307associated with normal plant evolutions such as RCP start and swapping operating charging pumps with the RCS in a water-solid condition.
Additionally, the impact on the P/T limits and OPPS setpoints has been evaluated for an increased service period to 15 EFPYs based on ASME Section XI, Appendix G, requirements. The results indicate that OPPS would significantly restrict the ability to perform plant heatup and cooldown, create an unnecessary burden to plant operations, and challenge control of plant evolutions required with OPPS enabled. 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 CFR50.12(a)(2)(ii), the underlying purpose of the ASME Code and the NRC regulations to ensure an acceptable margin of safety.
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. Special circumstances are present whenever, according to 10 CFR 50.12(a)(2)(ii), “Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule * * *.”
The underlying purpose of 10 CFR part 50, section 50.60(a), and 10 CFR part 50, appendix G, is to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. This is accomplished through these regulations that, in part, specify fracture toughness requirements for ferritic materials of the reactor coolant pressure boundary. The NRC staff accepts the licensee's determination that an exemption would be required to approve the use of Code Case N-640.
The NRC staff examined the licensee's rationale to support the exemption request. Based upon a consideration of the conservatism that is explicitly incorporated into the methodologies of 10 CFR part 50, Appendix G; ASME Section XI, appendix G; and Regulatory Guide 1.99, Revision 2, the NRC staff finds that the application of Code Case N-640 will provide results which are sufficiently conservative to ensure the integrity of the reactor coolant pressure boundary and, thus, meet the underlying intent of 10 CFR part 50, section 50.60(a), and 10 CFR part 50, appendix G. This is also consistent with determinations that the NRC staff has reached for other licensees under similar conditions, and based on the same considerations. Therefore, the NRC staff finds that special circumstances set forth in 10 CFR 50.12(a)(2)(ii) are present and that the methodology of Code Case N-640 may be used to revise the P/T limits and the LTOP setpoints for the Beaver Valley Unit 2 RCS.
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 FENOC an exemption from the requirements of 10 CFR part 50, section 50.60(a), and 10 CFR part 50, appendix G, for the Beaver Valley Unit 2 reactor coolant system.
This exemption is effective upon issuance.Start Signature
Dated at Rockville, Maryland, this 6 day of September 2000.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 00-23526 Filed 9-12-00; 8:45 am]
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