[Federal Register Volume 62, Number 151 (Wednesday, August 6, 1997)]
[Notices]
[Pages 42266-42267]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-20643]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-244]


In the Matter of Rochester Gas and Electric Corporation; R.E. 
Ginna Nuclear Power Plant; Exemption

I

    On December 10, 1984, the Nuclear Regulatory Commission issued 
Facility Operating License No. DPR-18 to Rochester Gas and Electric 
Corporation (RG&E or the Licensee) for the R.E. Ginna Nuclear Power 
Plant. The license stipulated, among other things, that the facility is 
subject to all rules, regulations, and orders of the Commission.

II

    In its letter dated June 12, 1997, the licensee requested an 
exemption from the Commission's regulations. Section 50.60 of Title 10 
of the Code of Federal Regulations, ``Acceptance Criteria for Fracture 
Prevention Measures for Lightwater Nuclear Power Reactors for Normal 
Operation,'' states that all lightwater nuclear power reactors must 
meet the fracture toughness and material surveillance program 
requirements for the reactor coolant pressure boundary as set forth in 
Appendices G and H to 10 CFR part 50. Appendix G to 10 CFR part 50 
defines pressure/temperature (P/T) limits during any condition of 
normal operation, including anticipated operational occurrences and 
system hydrostatic tests to which the pressure boundary may be 
subjected over its service lifetime. It also states that the American 
Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME 
Code) edition and addenda specified in 10 CFR 50.55a are applicable. It 
is specified in 10 CFR 50.60(b) that alternatives to the described 
requirements in Appendices G and H to 10 CFR Part 50 may be used when 
an exemption is granted by the Commission under 10 CFR 50.12.
    To prevent low-temperature overpressure transients that would 
produce pressure excursions exceeding the 10 CFR part 50, Appendix G, 
P/T limits while the reactor is operating at low temperatures, the 
licensee installed a low-temperature overpressure protection (LTOP) 
system. The system includes pressure-relieving devices called power-
operated relief valves (PORVs). The PORVs are set at a pressure low 
enough so that if an LTOP transient occurred, the mitigation system 
would prevent the pressure in the reactor vessel from exceeding the 10 
CFR part 50, Appendix G, P/T limits. To prevent the PORVs from lifting 
as a result of normal operating pressure surges (e.g., reactor coolant 
pump starting, and shifting operating charging pumps) with the reactor 
coolant system in a solid water condition, the operating pressure must 
be maintained below the PORV setpoint. Applying the LTOP instrument 
uncertainties required by the staff's approved methodology results in 
an LTOP setpoint that establishes an operating window that is too 
narrow to

[[Page 42267]]

permit reasonable system makeup and pressure control.
    To prevent these difficulties, the licensee has requested to use 
the ASME Code Case N-514, ``Low Temperature Overpressure Protection,'' 
which designates the allowable pressure as 110 percent of that 
specified by 10 CFR part 50, Appendix G. This would provide an 
increased band to permit system makeup and pressure control. ASME Code 
Case N-514 is consistent with guidelines developed by the ASME Working 
Group on Operating Plant Criteria to define pressure limits during LTOP 
events that avoid certain unnecessary operational restrictions, provide 
adequate margins against failure of the reactor pressure vessel, and 
reduce the potential for unnecessary activation of pressure-relieving 
devices used for LTOP. The content of this ASME Code Case has been 
incorporated into Appendix G of Section XI of the ASME Code and 
published in the 1993 Addenda to Section XI and has been incorporated 
into the latest draft of Regulatory Guide 1.147 (Draft Regulatory Guide 
DG-1050, Revision 12 of Regulatory Guide 1.147, Inservice Inspection 
Code Case Applicability ASME Section XI, dated May 1997). However, 10 
CFR 50.55a, ``Codes and Standards,'' only authorizes addenda through 
the 1988 Addenda.

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. According to 10 CFR 
50.12(a)(2)(ii), special circumstances are present whenever application 
of the regulation in question 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 ferritic materials of 
pressure-retaining components of the reactor coolant pressure boundary 
to provide adequate margins of safety during any condition of normal 
operation, including anticipated operational occurrences, to which the 
pressure boundary may be subjected over its service lifetime. Section 
IV.A.2 of Appendix G requires that the reactor vessel be operated with 
P/T limits at least as conservative as those obtained by following the 
methods of analysis and the required margins of safety of Appendix G of 
the ASME Code.
    Appendix G of the ASME Code requires that the P/T limits be 
calculated: (a) Using a safety factor of two on the principal membrane 
(pressure) stresses; (b) assuming a flaw at the surface with a depth of 
one-quarter (\1/4\) of the vessel wall thickness and a length of six 
(6) times its depth; and (c) using a conservative fracture toughness 
curve that is based on the lower bond of static, dynamic, and crack 
arrest fracture toughness tests on material similar to the Ginna 
reactor vessel material.
    In determining the setpoint for LTOP events, the licensee proposed 
to use safety margins based on an alternate methodology consistent with 
the ASME Code Case N-514 guidelines. The ASME Code Case N-514 allows 
determination of the setpoint for LTOP events such that the maximum 
pressure in the vessel would not exceed 110 percent of the P/T limits 
of the existing ASME Code Appendix G. This results in a safety factor 
of 1.8 on the principal membrane stresses. All other factors, including 
assumed flaw size and fracture toughness, remain the same. Although 
this methodology would reduce the safety factor on the principal 
membrane stress, the proposed criteria will provide adequate margins of 
safety on the reactor vessel during LTOP transients, and thus will 
satisfy the underlying purpose of 10 CFR 50.60 for fracture toughness 
requirements. Further, by relieving the operational restrictions, the 
potential for undesirable lifting of the PORV would be reduced, thereby 
improving plant safety.

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 LTOP 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), in that 
application of 10 CFR 50.60 is not necessary in order to achieve the 
underlying purpose of this regulation.
    Accordingly, the Commission has determined that, pursuant to 10 CFR 
50.12, this exemption is authorized by law, will not present an undue 
risk to the public health and safety, and is consistent with the common 
defense and security.
    Accordingly, the Commission hereby grants an exemption from 10 CFR 
50.60 such that in determining the setpoint for LTOP events, the 
Appendix G curves for P/T limits are not exceeded by more than 10 
percent. This exemption permits using the safety margins recommended in 
the AMSE Code Case N-514, in lieu of the safety margins required by 10 
CFR part 50, Appendix G. This exemption is applicable only to LTOP 
conditions during normal operation.
    Pursuant to 10 CFR 51.32, the Commission has determined that the 
granting of the exemption will have no significant impact on the 
quality of the human environment (62 FR 40554).
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 28th day of July, 1997.

    For the Nuclear Regulatory Commission.
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 97-20643 Filed 8-5-97; 8:45 am]
BILLING CODE 7590-01-P