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STP Nuclear Operating Company, et al., South Texas Project, Units 1 and 2; Exemption

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Information about this document as published in the Federal Register.

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1.0 Background

STP Nuclear Operating Company, et al. (STPNOC or the licensee) is the holder of Facility Operating License Nos. NPF-76 and NPF-80, which authorize operation of the South Texas Project, Units 1 and 2 (STP or the facilities). The licenses provide, among other things, that the licensee is subject to all rules, regulations, and orders of the U.S. Nuclear Regulatory Commission (NRC or the Commission) now or hereafter in effect.

The facilities consist of two pressurized-water reactors located at the licensee's site in Matagorda County, Texas.

2.0 Request/Action

Section 21.3 of Title 10 of the Code of Federal Regulations part 21 (10 CFR 21.3), provides the definition of basic component as it relates to the reporting of defects and nonconformances. By letter dated July 13, 1999, as supplemented, October 14 and 22, 1999, January 26, and August 31, 2000, and January 15, 18, 23, March 19, May 8 and 21, 2001, (hereinafter, the submittal) the licensee requested an exemption from the definition of basic component to exclude safety-related structures, systems, or components (SSCs) classified in accordance with its risk-informed categorization process as low safety significant (LSS) or non-risk significant (NRS) from the scope of the definition of basic component. STPNOC proposed that it would not apply procurement, dedication, and reporting requirements in 10 CFR part 21 to safety-related LSS and NRS SSCs. STPNOC stated that 10 CFR Part 21 imposes procurement and dedication requirements and requires the reporting of defects and noncompliances involving basic components whose failure could cause a substantial safety hazard. Also, STPNOC stated that reporting of defects and noncompliance involving safety-related LSS and NRS SSCs is not necessary to meet the purpose of 10 CFR part 21 because failure of such SSCs would not result in a substantial safety hazard.

3.0 Discussion

The Commission, pursuant to 10 CFR 21.7, may grant exemptions from the requirements of 10 CFR Part 21 as it determines are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest.

The U.S. Nuclear Regulatory Commission (NRC) has completed its evaluation of STPNOC's request for an exemption from the definition of basic component in 10 CFR 21.3. As it relates to nuclear power plants licensed pursuant to 10 CFR part 50, a basic component is defined as a SSC, or part thereof, that affects its safety function necessary to assure (1) the integrity of the reactor coolant pressure boundary; (2) the capability to shut down the reactor and maintain it in a safe shutdown condition; or (3) the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to those referred to in 10 CFR 50.34(a)(1) or 10 CFR 100.11. Further, a basic component is defined as an item designed and manufactured under a quality assurance program complying with 10 CFR part 50, Appendix B, or commercial-grade items which have successfully completed the dedication process. Finally, the definition of basic component includes the safety-related design, analysis, inspection, testing, fabrication, replacement of parts, or consulting services that are associated with the SSC hardware.

In the discussion of the purpose in 10 CFR 21.1, the need to identify the failure of SSCs to satisfy requirements Start Printed Page 41910(e.g., NRC regulations or Atomic Energy Act), or identify SSCs that contain defects, is related to conditions that could result in a substantial safety hazard. A substantial safety hazard is defined in 10 CFR 21.3 as meaning a loss of safety function to the extent that there is a major reduction in the degree of protection provided to public health and safety.

In the safety evaluation (SE), dated August 3, 2001, prepared in support of this exemption, the NRC describes its assessment of the attributes of the proposed treatment processes for LSS and NRS SSCs. The NRC determined that the proposed alternative treatment processes, if effectively implemented, will provide reasonable confidence that safety-related LSS and NRS SSCs remain capable of performing their safety functions under design-basis conditions. Also, as discussed in the SE, the NRC determined that the licensee's categorization process provides a reasonable method for determining that safety-related LSS and NRS SSCs have a small contribution to overall safety. Further, the sensitivity study conducted by the licensee demonstrates that for relatively large changes in availability of all of the safety-related LSS SSCs modeled in the probabilistic risk assessment, there is only a small change in the overall plant risk. Therefore, the NRC determined that it is acceptable to exclude LSS and NRS SSCs from the scope of the definition of basic component in 10 CFR 21.3 because the NRC concluded that defects in these components would not result in a substantial safety hazard and thus reporting of such defects is not necessary. On this basis, the NRC finds that the proposed exemption will not endanger life or property or the common defense and security.

The NRC also finds the proposed exemption is otherwise in the public interest since it focuses NRC and licensee attention on the most safety and risk significant SSCs. Further, the NRC finds that the proposed exemption is authorized by law. Thus, the NRC finds that the proposed exemption satisfies the criteria given in 10 CFR 21.7 and should be granted.

4.0 Conclusion

Accordingly, the Commission has determined that, pursuant to 10 CFR 21.7, the exemption is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest. Therefore, the Commission hereby grants, subject to the conditions described below, STPNOC the exemption from the definition of basic component in (1)(ii) of 10 CFR 21.3 for SSCs at STP categorized as LSS and NRS. As conditions of this exemption:

1.The licensee described the categorization, treatment, and oversight (evaluation and assessment) processes in its submittal dated July 13, 1999, as supplemented October 14 and 22, 1999, January 26 and August 31, 2000, and January 15, 18, 23, March 19, May 8 and 21, 2001. The licensee has documented these processes in a proposed Final Safety Analysis Report (FSAR) submittal dated May 21, 2001, found acceptable by the staff as the regulatory basis for granting this exemption (see the NRC's SE dated August 3, 2001). The licensee shall incorporate this proposed FSAR submittal into the STP FSAR and shall implement the categorization, treatment, and oversight processes consistent with the STP FSAR descriptions.

2.The licensee shall implement a change control process that incorporates the following requirements:

a. Changes to FSAR Section 13.7.2, “Component Categorization Process,” dated May 21, 2001, and found acceptable by the NRC as described in the NRC's SE dated August 3, 2001, may be made without prior NRC approval, unless the change would decrease the effectiveness of the process in identifying high safety significant and medium safety significant components.

b. Changes to FSAR Section 13.7.3, “Treatment of Component Categories,” dated May 21, 2001, and found acceptable by the NRC as described in the NRC's SE dated August 3, 2001, may be made without prior NRC approval, unless the change would result in a reduction in the assurance of component functionality.

c. Changes to FSAR Section 13.7.4, “Continuing Evaluations and Assessments,” dated May 21, 2001, and found acceptable by the NRC as described in the NRC's SE dated August 3, 2001, may be made without prior NRC approval, unless the change would result in a decrease in effectiveness of the evaluations and assessments.

d. The licensee shall submit a report, as specified in 10 CFR 50.4, of changes made without prior NRC approval pursuant to these provisions. The report shall identify each change and describe the basis for the conclusion that the change does not involve a decrease in effectiveness or assurance as described above. The report shall be submitted within 60 days of the date of the change.

e. Changes to FSAR Sections 13.7.2, 13.7.3, and 13.7.4 that do not meet the criteria of a through c above shall be submitted to the NRC for prior review and approval.

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 32397). Accordingly, based upon the environmental assessment, the Commission has determined that the granting of this exemption will not have a significant effect on the quality of the human environment.

This exemption is effective upon submittal of a FSAR update pursuant to 10 CFR 50.71(e) incorporating the FSAR Sections described in the conditions above.

Start Signature

Dated at Rockville, Maryland, this 3rd day of August, 2001.

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. 01-19973 Filed 8-8-01; 8:45 am]

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