[Federal Register Volume 63, Number 88 (Thursday, May 7, 1998)]
[Notices]
[Pages 25242-25243]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-12183]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-282, 50-306]


Northern States Power Company (Prairie Island Nuclear Generating 
Plant, Units 1 and 2); Exemption

I

    Northern States Power Company (NSP, the licensee) is the holder of 
Facility Operating License Nos. DPR-42 and DPR-60, which authorize 
operation of Prairie Island Nuclear Generating Plant, Units 1 and 2, 
respectively. The licenses provide, among other things, that the 
licensee is subject to all rules, regulations, and orders of the 
Commission now or hereafter in effect.
    The facility consists of two pressurized-water reactors located at 
the licensee's site in Goodhue County, Minnesota.

II

    In its letter dated March 6, 1998, the licensee requested an 
exemption from specific requirements of Title 10 of the Code of Federal 
Regulations Part 50, Section 60, and Appendix G. Specifically, NSP 
proposed to use American Society of Mechanical Engineers (ASME) Code 
Case N-514 to permit setting the pressure setpoint of each unit's 
overpressure protection system (OPPS) so that the pressure-temperature 
(P-T) limits required by 10 CFR Part 50, Appendix G, could be exceeded 
by 10 percent during a low temperature pressure transient.
    The NRC has established requirements in 10 CFR Part 50 to protect 
the integrity of the reactor coolant system pressure boundary. As a 
part of these, 10 CFR Part 50, Appendix G, requires that P-T limits be 
established for reactor pressure vessels during normal operation, 
including anticipated operational occurrences and vessel hydrostatic 
testing and as stated in Appendix G, ``The appropriate requirements on 
* * * the pressure-temperature limits * * * must be met for all 
conditions.'' In order to ensure these P-T limit curves are not 
exceeded and provide pressure relief during low temperature 
overpressurization events, pressurized-water reactor licensees have 
installed protection systems (OPPS) as part of the reactor coolant 
system pressure boundary. NSP is required as part of the Prairie Island 
Units 1 and 2 Technical Specifications to develop, update, and submit 
reactor vessel P-T limits and OPPS setpoints for NRC review and 
approval.
    By letter dated March 6, 1998, NSP submitted an exemption request 
to enable the use of ASME Code Case N-514 as an alternative method for 
determining the OPPS pressure setpoint. NSP determined that the 
exemption request from the provisions of 10 CFR 50.60 and Appendix G 
was necessary since these regulations require, as noted above, that the 
reactor vessel conditions not exceed the P-T limits established by 
Appendix G. In referring to 10 CFR 50.12 on specific exemptions, NSP 
cited special circumstances as stated in 10 CFR 50.12(a)(2)(ii) on 
achieving the underlying purpose of the regulations as its basis for 
requesting this exemption.

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. 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, Appendix G, is to 
establish fracture toughness requirements for the RCS pressure boundary 
to provide adequate margins of safety during any condition of normal 
operation. NSP stated that the OPPS provides a physical means of 
protecting the vessel by not exceeding the limits. NSP proposed that 
establishing the OPPS pressure setpoint per the N-514 provisions such 
that the vessel pressure would not exceed 110 percent of the P-T limit 
allowables would still provide an acceptable level of safety and 
mitigate the potential for an inadvertent actuation of the OPPS. The 
finding of an ``acceptable level of safety'' while using N-514 was made 
based on the conservatisms that have been explicitly incorporated into 
the procedure for developing the P-T limit curves. This procedure, 
referenced from Appendix G to Section XI of the ASME Code, includes the 
following conservatisms: (1) A safety factor of 2 on the pressure 
stresses, (2) a margin factor applied to the determination of 
RT<INF>NDT</INF>

[[Page 25243]]

[reference temperature <INF>nil ductility temperature</INF>] (using 
Regulatory Guide 1.99 ``Radiation Embrittlement of Reactor Vessel 
Materials,'' Revision 2), and (3) a limiting material toughness curve 
based on bounding dynamic crack initiation and crack arrest data.
    In addition, NSP explained that plant operators must operate the 
plant between the minimum pressure required to preserve reactor coolant 
pump seals and a maximum pressure that does not challenge the power-
operated relief valve setpoint. Without the application of ASME Code 
Case N-514, Prairie Island would have an operating window that is too 
narrow to permit reasonable system makeup and pressure control. NSP 
continued by stating that further reduction of the OPPS pressure 
setpoint below 500 psig would increase the probability that the reactor 
coolant pump's no. 1 seal will fail as a result of OPPS operation, and 
that such a seal failure could produce a breach in the reactor coolant 
system boundary that could not be isolated. Therefore, inadvertent OPPS 
actuation could lead to a small break loss-of-coolant accident and the 
unnecessary release of reactor coolant inside containment.

IV

    For the foregoing reasons, the NRC staff has concluded that the 
licensee's proposed use of the alternate methodology in determining the 
acceptable setpoint for OPPS events will not present an undue risk to 
public health and safety and is consistent with the common defense and 
security. The NRC staff has determined that there are special 
circumstances present, as specified in 10 CFR 50.12(a)(2)(ii), in that 
the application of 10 CFR 50.60 is not necessary in order to achieve 
the underlying purpose of this regulation.
    The NRC staff agreed with NSP's determination that an exemption 
would be required to approve the use of Code Case N-514. The NRC staff 
examined NSP's rationale to support the exemption request and concluded 
that the use of Code Case N-514 would also meet the underlying intent 
of the regulations. Based upon a consideration of the conservatisms 
that are explicitly defined in the Appendix G methodology (as listed in 
Section III above), the staff concluded that permitting the OPPS 
setpoint to be established such that the vessel pressure would not 
exceed 110 percent of the limit defined by the P-T limit curves would 
provide an adequate margin of safety against brittle failure of the 
reactor vessel. This is also consistent with the determination that the 
staff has reached for other licensees under similar conditions based on 
the same considerations. Therefore, requesting the exemption under the 
special circumstances of 10 CFR 50.12(a)(2)(ii) was found to be 
appropriate. The staff also agrees that limiting the potential for 
inadvertent OPPS actuation (and limiting the potential for reactor 
coolant pump seal damage) may improve plant safety.
    Accordingly, the Commission has determined that, pursuant to 10 CFR 
50.12(a), an 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 an exemption from the 
requirements of 10 CFR 50.60 and Appendix G to allow NSP to apply the 
methods in ASME Code Case N-514 for the determination of the Prairie 
Island Nuclear Generating Plant Units 1 and 2 pressure setpoints.
    Pursuant to 10 CFR 51.32, the Commission has determined that the 
granting of this exemption will not have a significant effect on the 
quality of the human environment (63 FR 23477).
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 30th day of April 1998.

    For the Nuclear Regulatory Commission.
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 98-12183 Filed 5-6-98; 8:45 am]
BILLING CODE 7590-01-P