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STP Nuclear Operating Company; South Texas Project Electric Generating Station, Units 1 and 2; Environmental Assessment and Finding of No Significant Impact

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Start Preamble

The U.S. Nuclear Regulatory Commission (NRC) is considering issuance of exemptions from certain regulations found in 10 CFR parts 21, 50, and 100 for Facility Operating License Nos. NPF-76 and NPF-80, issued to STP Nuclear Operating Company (STPNOC or the licensee) for operation of the South Texas Project Generating Station, Units 1 and 2, (STP) located in Matagorda County, Texas.

Environmental Assessment

Identification of the Proposed Action

The proposed action would grant the licensee relief from certain special treatment requirements found in Title 10 of the Code of Federal Regulations, Parts 21, 50, and 100 (10 CFR parts 21, 50 and 100) for certain structures, systems, and components (SSCs). The licensee has used a risk-informed process to categorize SSCs as low safety significant (LSS) or non-risk significant (NRS); and other SSCs as medium safety significant (MSS) or high safety significant (HSS). The purpose of this categorization process is to identify those SSCs for which the special treatment requirements may be relaxed. Currently, LSS and NRS SSCs, which are not as risk significant as MSS and HSS SSCs, are treated with the same level of protection. The licensee is seeking limited exemptions from the following regulations for just those SSCs that have been categorized as LSS or NRS:

1. Requirements for quality assurance (QA) found in:

a. 10 CFR part 50, Appendix B, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,” for QA requirements on SSCs that are safety-related (with the exception of the Criterion III, “Design Control,” Criterion XV, “Nonconforming Materials, Parts, or Components,” and Criterion XVI, “Corrective Action”);

b. 10 CFR part 50, Appendix A, General Design Criteria (GDC) 1, “Quality Standards and Records,” for SSCs important to safety that contains quality assurance program and record keeping requirements;

c. 10 CFR 50.34(b)(6)(ii) that requires the licensee to describe in the Final Safety Analysis Report how 10 CFR part 50, Appendix B, requirements are being satisfied;

d. 10 CFR 50.54(a)(3) regarding NRC review and approval of changes to the QA program that result in a reduction in commitments in the program description as accepted by the NRC for LSS and NRS SSC program descriptions; and,

e. 10 CFR 21.3 defining the term “basic component” that includes safety-related LSS and NRS SSCs and impose 10 CFR part 21 requirements for procurement, dedication, and reporting.

2. Requirements for environmental qualification (EQ) found in:

a. 10 CFR 50.49(b) that defines the scope of electric components important to safety subject to the EQ program requirements of 10 CFR 50.49;

b. 10 CFR part 50, Appendix A, GDC 2, “Design Bases for Protection Against Natural Phenomena,” for tests and inspections to demonstrate that SSCs important to safety are designed to withstand the effects of natural phenomena without loss of capability to perform their safety functions;

c. 10 CFR part 50, Appendix A, GDC 4, “Environmental and Dynamic Effects Design Bases,” for tests and inspections to demonstrate that SSCs important to safety are able to withstand environmental conditions of normal operation, maintenance, testing, and postulated accidents; and,

d. 10 CFR part 100, Appendix A, Sections VI(a)(1) and (a)(2) for testing and inspection to demonstrate that SSCs within the scope of these regulations[1] are designed to remain functional during a safe-shutdown earthquake and operating-basis earthquake, respectively, and 10 CFR 50.34(b)(10) and 10 CFR 50.34(b)(11) to the extent that they reference the 10 CFR part 100, Appendix A, criteria, discussed above.

3. Requirements for testing and inspection found in:

a. 10 CFR part 50, Appendix A, GDC 18, “Inspection and Testing of Electric Power Systems,” that requires SSCs important to safety be designed to permit inspection and testing; and

b. 10 CFR part 50, Appendix J, Option B, section III.B, “Type B and C Tests,” that requires Type C containment Start Printed Page 32398isolation valve leak rate tests for safety-related SSCs.

4. Requirements for monitoring the effectiveness of maintenance under 10 CFR 50.65 for safety-related SSCs and nonsafety-related SSCs that are relied upon to mitigate accidents or transients or are used in plant emergency operating procedures, or whose failure could prevent safety-related SSCs from fulfilling their safety-related function, or whose failure could cause a reactor scram or actuation of a safety-related system. The licensee is requesting an exemption to exclude the LSS and NRS SSCs from the scope of the maintenance rule but would still conduct monitoring at the plant, system, or train level. Failure of an LSS or NRS SSC would not count as a Maintenance Rule Functional Failure unless the failure caused a failure of a high or medium safety significant function.

5. Industry code standards found in:

a. 10 CFR 50.55a(f) and (g) that require repair and replacement, inservice testing (IST), and inservice inspection (ISI), under Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code; and,

b. 10 CFR 50.55a(h) that imposes the quality and qualification requirements of Sections 4.3 and 4.4 of Institute of Electrical and Electronics Engineers (IEEE) 279, “Criteria for Protection Systems for Nuclear Power Plant Generating Stations,” for electric SSCs important to safety.

6. 10 CFR 50.59 to the extent that this regulation requires a written evaluation and prior NRC review and approval of changes in special treatment requirements for LSS and NRS SSCs.

The proposed action is in accordance with the licensee's application for exemption dated July 13, 1999, as supplemented on October 14 and 22, 1999; January 26 and August 31, 2000; and January 15, 18, and 23, March 19, and May 8 and 21, 2001 (hereinafter, the submittal).

The Need for the Proposed Action

The exemptions are necessary to provide the licensee relief from regulatory requirements found in 10 CFR parts 21, 50, and 100 for LSS and NRS components currently within the scope of these regulations. In accordance with 10 CFR 21.7 and 10 CFR 50.12, the Commission may grant exemptions from the requirements of 10 CFR parts 21 and 50, respectively, under certain circumstances. Further, the NRC staff has determined that the requested exemptions from 10 CFR part 100, Appendix A, sections VI(a)(1) and (a)(2) may be granted in accordance with the requirements of 10 CFR 50.12. The NRC staff has approved a Graded Quality Assurance Program for STP. Exemptions from certain special treatment requirements are necessary to realize the full benefit of the Graded Quality Assurance Program.

The exemption is also necessary to reduce occupational radiation exposures and costs that would be expended in providing qualifications, quality assurance controls, maintenance, monitoring requirements, testing, and inspections for the LSS and NRS components that may not be necessary to maintain safety.

Environmental Impacts of the Proposed Action

The NRC staff has completed its evaluation of the proposed action and concludes that many of the exemption requests are not necessary for, or are not consistent with the objective of, maintaining design and functionality of an SSC and will not be granted. The NRC staff has determined that some of the exemption requests that would remove LSS and NRS SSCs from the scope of the regulations, if granted, would not present an undue risk to the public health and safety. The regulations for which exemptions are to be granted are listed below and are referred to as the proposed action in the following sections.

a. 10 CFR 21.3—definition of basic component;

b. 10 CFR 50.34(b)(10) and 10 CFR 50.34(b)(11), impose the requirements of 10 CFR part 100, Appendix A, section VI(a)(1) and (2);

c. 10 CFR 50.49(b), scope of electrical equipment subject to environmental qualification requirements [design aspects of 10 CFR 50.49(e)(1) through (7) continue to apply];

d. 10 CFR 50.55a(f)—IST requirements;

e. 10 CFR 50.55a(g)—repair and replacement, and ISI requirements;

f. 10 CFR 50.55a(h), quality and qualification requirements of sections 4.3 and 4.4 of IEEE 279;

g. 10 CFR 50.59—written evaluations and prior NRC review and approval for changes to special treatment requirements;

h. 10 CFR 50.65(b)—scope of maintenance rule [the requirements of 10 CFR 50.65(a)(4) continue to apply];

i. 10 CFR part 50, App. B—quality assurance requirements (the requirements of Criteria III, “Design Control,” XV, “Nonconforming Materials, Parts, or Components,” and XVI, “Corrective Action,” continue to apply);

j. 10 CFR part 50, Appendix J, Option B, section III.B, Type C containment isolation valve leak rate tests only;

k. 10 CFR part 100, Appendix A, sections VI(a)(1) and (2), seismic requirements for safe shutdown and operating basis earthquakes.

The regulations, listed above, apply to SSCs that are located entirely within the restricted area and, if the exemptions are granted, would not result in off-site impacts due to normal operation. The NRC staff evaluated the licensee's probabilistic risk analysis (PRA) sensitivity study that addressed the overall impact of reduced treatment for LSS and NRS SSCs on plant risk for those LSS SSCs that are modeled in the STP PRA.[2] Since the impact on failure rates for these SSCs resulting from a reduction in special treatment requirements is not known, a factor of 10 increase in the failure rates of all LSS SSCs modeled in the STP PRA was used. The results of the sensitivity analysis showed that the overall plant risk for a core damage event increased by 2.7 percent. The large early release frequency increased by about 1.2 percent. The NRC staff finds the sensitivity study to be an acceptable method of ensuring that the cumulative risk is only slightly impacted when predicting significant changes in the SSC failure rates, which may not occur. Therefore, the NRC staff finds that the postulated change in failure rates of the LSS SSCs that are modeled in the PRA would be expected to have a low overall impact on plant risk.

On the other hand, the proposed exemptions may have a beneficial impact on occupational exposure, since the additional requirements for QA, EQ, monitoring, testing, and inspection for certain LSS and NRS components would not be necessary. The magnitude of this benefit has not been quantified.

The proposed action will not significantly increase the probability or consequences of accidents, no changes are being made in the types of any effluents that may be released off site, and there is no significant increase in occupational or public radiation exposure. Therefore, there are no significant radiological environmental impacts associated with the proposed action.

With regard to potential nonradiological 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 Start Printed Page 32399are no significant nonradiological environmental impacts associated with the proposed action.

Accordingly, the NRC staff concludes that there are no significant environmental impacts associated with the proposed action.

Alternatives to the Proposed Action

As an alternative to the exemption requests listed above, the NRC staff considered denial of the proposed action (i.e., the “no-action” alternative). Denial of the application would result in no significant change in current environmental impacts.

Another alternative is to await applicable regulations that are the result of a future rulemaking under Option 2 of the Commission's alternatives to risk inform 10 CFR part 50 of the NRC's regulations discussed in SECY-98-300, “Options for Risk-Informed Revisions to 10 CFR part 50, “Domestic Licensing of Production and Utilization Facilities'.” The exemptions requested by the licensee are a proof-of-concept for this broader rulemaking effort. The Commission plans to use the STPNOC exemption request and other industry pilot programs to assist with the development of the revised risk-informed 10 CFR part 50. The only adverse environmental impact associated with this proposed action would be a slight increase in the risk of an accident, but this impact would not be significantly changed with the alternative of awaiting a rulemaking. Therefore, any relief granted under a subset of a larger set of risk-informed regulations under Option 2 in lieu of the exemption requests would not provide a significant benefit to public health or safety, or the environment. The environmental impacts associated with granting the exemptions found to be acceptable by the NRC staff and the alternatives listed above are similar.

Alternative Use of Resources

This action does not involve the use of any resources not previously considered in the Final Environmental Statement (NUREG-1171) for the South Texas Project, Units 1 and 2, dated August 1986.

Agencies and Persons Consulted

In accordance with its stated policy, on June 1, 2001, the NRC staff consulted with the Texas State official, Arthur C. Tate, of the Division of Compliance and Inspection, Bureau of Radiation Control, Texas Department of Health, regarding the environmental impact of the proposed action. The State official had no comments.

Finding of No Significant Impact

On the basis of the environmental assessment, the NRC staff concludes that the proposed action will not have a significant effect on the quality of the human environment. Accordingly, the NRC staff has determined not to prepare an environmental impact statement for the proposed action.

For further details with respect to the proposed action, see the licensee's letter dated July 13, 1999, as supplemented on October 14 and 22, 1999; January 26 and August 31, 2000; and January 15, 18, and 23, March 19, and May 8 and 21, 2001. 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 of the NRC web site http://www.nrc.gov (Electronic Reading Room).

Start Signature

Dated at Rockville, Maryland, this 8th day of June, 2001.

For the Nuclear Regulatory Commission.

Cynthia A. Carpenter,

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

End Signature End Preamble

Footnotes

1.  The scope of (a)(1) are those SSCs necessary to assure (i) the integrity of the reactor coolant pressure boundary, (ii) the capability to shut down the reactor and maintain it in a safe condition, or continued (iii) the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures of part 100. The scope of (a)(2) are those SSCs necessary for continued operation without undue risk to the health and safety of the public.

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2.  There are no NRS SSCs and limited LSS SSCs modeled in the plant's PRA due to a negligible impact on risk or due to implicit modeling.

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[FR Doc. 01-14976 Filed 6-13-01; 8:45 am]

BILLING CODE 7590-01-P